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	description	This application is based on and claims the benefit of priority from prior Japanese Patent Application No. 2006-250316, filed on Sep. 15, 2006 the entire contents of which are incorporated herein by reference. 1. Field of the Invention The present invention relates to a charged particle beam lithography system and a method for evaluating the same. 2. Description of the Related Art As semiconductor integrated circuits are highly integrated, LSI patterns are becoming much smaller and complicated. The light exposure is thus desired to use a much smaller exposure mask and reticle used for the LSI patterns. Another proposed method for dealing with such smaller LSI patterns directly draws the LSI pattern by the charged particle beam including an electron beam and an ion beam. This charged particle beam lithography technology is generally not a batch exposure such as the light exposure using the LSI mask and the like, but a process that scans a substrate with a charged particle beam to draw a pattern thereon. The charged particle beam lithography technology thus takes a time to draw a pattern, providing less throughput. Some methods are proposed for improving the throughput of the charged particle beam lithography system. Specifically, one method is a stage control method referred to as a step & repeat scheme. This scheme stops the stage during drawing and moves the stage to the next area when a pattern is drawn to a drawable area. The scheme may advantageously use a large charged particle beam deflection area, i.e., a drawable area. The scheme requires, however, time for actually drawing a pattern and additional time for moving to the next drawing area (a non-exposure time). Such a non-exposure time added to the total drawing time reduces the throughput. Another proposed scheme continuously moves a stage to increase the throughput. This scheme divides the pattern to be drawn into strip-like areas referred to as “frames”. Each frame is drawn while the stage is continuously moved. When the beam reaches the edge of the frame, the stage stepwise moves in a direction perpendicular to the direction of the continuous motion, and a pattern is drawn again while the stage is continuously moved in the opposite direction. This scheme requires a smaller charged-particle-beam deflection area than the step & repeat scheme because a pattern is drawn during the stage motion. This scheme may, however, increase the throughput because the stage stops only at the frame edge. The charged particle beam may be deflected in a scheme combined with a vector scan scheme (two-dimensional scanning scheme), for further increasing the throughput. The vector scan scheme further divides the frame into areas referred to as “subfields”. Only the necessary portion within the subfield is drawn by the deflected charged particle beam. Unlike the one-dimensional scanning scheme, the vector scan scheme does not scan a non-drawn portion with the turned-off charged particle beam, thus increasing the drawing speed. The vector scan scheme uses a main deflector and an auxiliary deflector. The main deflector positions the charged particle beam to the subfield. The auxiliary deflector draws the subfield by the charged particle beam. JP 10-284392A describes a drawing method that increases drawing accuracy in the charged particle beam lithography system. A charged particle beam lithography system according to an aspect of the present invention is a charged particle beam lithography system in which a pattern area to be drawn is divided into a plurality of frames, a main deflection positions a charged particle beam to a subfield within the frame, and an auxiliary deflection draws a pattern in units of the subfield, the lithography system comprising: a beam optical system including a deflector deflecting the beam; a driver driving the deflector; and a deflection control portion controlling the driver according to drawing data indicating a pattern to be drawn, the deflection control portion controlling the driver according to a settling time that is determined so that an offset of an irradiation position of the charged particle beam has a certain value irrespective of any changes in deflection amount of the auxiliary deflection in the subfield. A charged particle beam lithography system according to another aspect of the present invention is a charged particle beam lithography system in which a pattern area to be drawn is divided into a plurality of frames, a main deflection positions a charged particle beam to a subfield within the frame, and an auxiliary deflection draws a pattern in units of the subfield, the lithography system comprising: a beam optical system including a deflector deflecting the beam; a driver driving the deflector; a deflection control portion controlling the driver according to drawing data indicating a pattern to be drawn; a sub-field deflection calculation portion computing a deflection amount of the auxiliary deflection in the subfield; a settling time setup unit determining, according to the deflection amount of the auxiliary deflection, a settling time of the charged particle beam in the auxiliary deflection so that an offset of an irradiation position of the charged particle beam is a certain value irrespective of any changes in the deflection amount of the auxiliary deflection in the subfield. A method for evaluating a charged particle beam lithography system according to an aspect of the present invention is a method for evaluating a charged particle beam lithography system in which a pattern area to be drawn is divided into a plurality of frames, a main deflection positions a charged particle beam to a subfield within the frame, and an auxiliary deflection draws a pattern in units of the subfield, an irradiation pattern of the charged particle beam used for the evaluation being arranged on a first line and a second line including an intersection therebetween, the first and second lines being substantially perpendicular to each other, the method comprising: a first step of irradiating the charged particle beam to a first irradiation position at an intersection of the first and second lines, moving the charged particle beam from the first irradiation position to a second irradiation position in the first line, and irradiating the beam thereto; a second step of moving, after the first irradiation step, the charged particle beam from the second irradiation position to a third irradiation position in the second line, and irradiating the beam thereto; a third step of repeating, after the second irradiation step, the motion of the irradiation position along the first or second line and the irradiation of the charged particle beam, and the motion of the irradiation position between the first and second lines and the irradiation of the charged particle beam; and a fourth step of measuring an error between the irradiation pattern and an actual irradiation position of the charged particle beam to measure an offset of the deflection position in the subfield. A method for evaluating a charged particle beam lithography system according to another aspect of the present invention is a method for evaluating a charged particle beam lithography system in which a pattern area to be drawn is divided into a plurality of frames, a main deflection positions a charged particle beam to a subfield within the frame, and an auxiliary deflection draws a pattern in units of the subfield, a plurality of shots of the charged particle beams being irradiated to the pattern, the shots including odd shots and even shots of the charged particle beams, the odd shot beams having less deflection than the even shot beams or the even shot beams having less deflection than the odd shot beams. An embodiment of the present invention is described below. FIG. 1 is a configuration diagram of an electron beam lithography system according to a first embodiment of the present invention. The electron beam lithography system includes a sample chamber 1 that contains a stage 3 holding a sample 2 thereon. The sample 2 is, for example, a mask to which a pattern is drawn by the electron beam. The stage 3 is driven by a stage driver 4 in the X direction (the horizontal direction in the plane of FIG. 1) and in the Y direction (the vertical direction in the plane of FIG. 1). The position of the stage 3 is measured by a position circuit 5 including a laser distance meter or the like. An electron beam optical system 10 resides above the sample chamber 1. The optical system 10 includes an electron gun 6, various lenses 7, 8, 9, 11, and 12, a blanking deflector 13, a beam-dimension adjusting deflector 14, a beam-scanning main deflector 15, a beam-scanning auxiliary deflector 16, and two beam-shaping apertures 17 and 18. Referring to FIG. 2, the system draws a pattern on the sample 2 by the electron beam as follows. A pattern 51 to be drawn on the sample 2 is divided into strip-like frame areas 52. Each frame area 52 is drawn, while the stage 3 is continuously moved in the X direction. The frame area 52 is further divided into subfield areas 53. Only the necessary portion within the sub-field area 53 is drawn by a deflected variable shaped beam such as an electron beam 54. The optical system 10 includes two deflectors: a main deflector 15 and an auxiliary deflector 16. The main deflector 15 positions the electron beam to the sub-field area 53. The auxiliary deflector 16 draws the sub-field area 53 by the electron beam. The electron beam lithography system in this embodiment may draw a pattern on the sample as follows. The main deflector 15 positions the electron beam to a certain sub-field area 53. The auxiliary deflector 16 determines a pattern-drawing position within the certain sub-field area 53. The beam-dimension adjusting deflector 14 and the beam-shaping apertures 17 and 18 control the beam shape. The sub-field area 53 is drawn while the stage 3 is continuously moved in one direction. After one sub-field area 53 is thus drawn, the next sub-field area 53 is drawn. The frame area 52 is a strip-like drawing area defined by the deflection width of the main deflector 15. The sub-field area 53 is a unit drawing area defined by the deflection width of the auxiliary deflector 16. After the frame area 52 including a plurality of subfield areas 53 is drawn, the stage 3 is moved stepwise in a direction perpendicular to the direction in which the stage 3 is continuously moved, and then the above processes are repeated. The frame areas 52 are thus sequentially patterned. The controller 20 stores in its storage medium such as a magnetic disk 21 drawing data of a mask. After read from the magnetic disk 21, the drawing data is temporarily stored in the pattern memory 22 for each frame area 52. The pattern data for each frame area 52 stored in the pattern memory 22, i.e., frame information including drawing position and drawing figure data and the like is sent to a pattern data decoder 23 and a drawing data decoder 24 as a data analysis portion. Output from the pattern data decoder 23 connects to a blanking circuit 25 and a beam shape driver 26. Specifically, the pattern data decoder 23 uses the pattern data to create blanking data. The blanking data is sent to the blanking circuit 25. The pattern data decoder 23 also creates desired beam-dimension data. The beam-dimension data is sent to the beam shape driver 26. The beam shape driver 26 sends a certain deflection signal to the beam-dimension adjusting deflector 14 in the electron optical system 10. The deflection signal controls a dimension of the electron beam. The controller 20 also has a deflection controller 32 connected thereto. The deflection controller 32 connects to a settling time setup unit 31. The settling time setup unit 31 connects to a sub-field deflection calculator 30. The sub-field deflection calculator 30 connects to the pattern data decoder 23. The deflection controller 32 also connects to the blanking circuit 25, the beam shape driver 26, a main deflection driver 27, and an auxiliary deflection driver 28. The drawing data decoder 24 sends its output to the main deflection driver 27 and the auxiliary deflection driver 28. The main deflection driver 27 sends a certain deflection signal to the main deflector 15 in the electron optical system 10. The deflection signal deflects the electron beam to a certain main deflection position. The auxiliary deflection driver 28 sends a certain auxiliary deflection signal to the auxiliary deflector 16. The auxiliary deflection signal causes the electron beam to draw a pattern in the sub-field area 53. [Electron Beam Lithography Method] A drawing method of an electron beam lithography system in this embodiment is described below. The electron beam lithography system usually moves the electron beam, i.e., deflects the beam, to a certain position by applying a certain voltage to the deflector. The certain voltage cannot be applied quickly. It takes a time to reach the certain voltage. A settling time is thus set up. The settling time is a time necessary to stabilize the electric circuit such as a deflection amplifier in the deflector. The settling time corresponds to an interval between the electron-beam shots during drawing. It is assumed that the settling time has a correlation with an amount of deflection (moving distance) of the electron beam. The settling time is thus usually set up to have a linear relationship between the settling time in the sub-field area 53 and the amount of the deflection of the electron beam in the sub field. In contrast, in this embodiment, the settling time is set up according to the amount of the deflection of the electron beam and the actual time for stabilizing the deflection amplifier. It is thus possible to irradiate the electron beam to a position more accurately and improve the throughput. FIG. 3 shows a relationship between the amount of deflection of the electron beam and the settling time in the sub-field area 53. FIG. 3 relatively shows an error of a settling time with respect to an amount of the deflection of the electron beam (moving distance) in the sub-field area 53. Contours shown in FIG. 3 indicate relative values of the error. The contours sequentially show areas of, from the highest area, 1 or more, 0.7 or more, 0.5 or more, 0.2 or more, and 0.1 or more. As shown, in the vicinity of the maximum sub-field deflection amount of the electron beam in the sub-field area 53, a shorter settling time causes a larger error. The electron beam is thus irradiated to a position farther from the certain irradiation position. In the sub-field area 53, the maximum error has a relative value (maximum value) of about 1.2. It is empirically known that an error equal to or less than 1/12 of the maximum value may provide drawing with high accuracy. It is also empirically known that the throughput may benefit from an error equal to or more than 1/24 of the maximum value, which is half of 1/12 of the maximum value. For the high drawing accuracy and the high throughput, therefore, the settling time is preferably set up to provide an error between 1/12 and 1/24 of the maximum value, and more preferably, an error close to 1/12 of the maximum value. This embodiment therefore computes, according to the line 101 in FIG. 3 corresponding to the relative value of an error of 0.1 ( 1/12 of the maximum value), the settling time in the subfield from the subfield moving distance. This embodiment then uses the settling time to control the deflection in the subfield of the electron beam lithography system. It is thus possible to provide an electron beam lithography system having the high drawing accuracy and the high throughput. In addition, when it is desired that a high drawing accuracy is sacrificed and instead a drawing speed is made high, it is possible to calculate a settling time based on lines 100 or 99, not the line 101, to conduct a deflection control in a subfield. The drawing method of the electron beam lithography system thus configured is described below. First, the sample 2 is disposed on the stage 3 in the sample chamber 1. The position of the stage 3 is then detected by the position circuit 5. Then, according to the signal from the controller 20, the stage driver 4 moves the stage 3 to a position that allows a pattern to be drawn on the sample 2. The electron gun 6 then emits the electron beam. The electron beam is focused by a condenser lens 7. The blanking deflector 13 transmits the electron beam to the sample 2 or cut off the electron beam. The electron beam is then incident on the aperture 17. The beam passes through the opening of the aperture 17. The beam is then deflected by the beam-dimension adjusting deflector 14 that is controlled by the beam shape driver 26. The beam then passes through the opening of the beam-shaping aperture 18. The electron beam is thus finally shaped to a desired beam shape such as a spot pattern. The spot pattern is a drawing unit of the electron beam irradiated to the sample 2. A plurality of spot patterns form one drawing pattern. After the electron beam is shaped to the spot pattern, it is reduced by the reduction lens 11. A pattern is drawn on the sample 2 by irradiating the electron beam to the sample 2. The irradiation position on the sample 2 is controlled by the main deflector 15 and the auxiliary deflector 16. The deflection driver 27 controls the main deflector 15. The auxiliary deflection driver 28 controls the auxiliary deflector 16. The main deflector 15 positions the electron beam to a certain auxiliary deflection area (sub-field area 53) on the sample 2. The auxiliary deflector 16 controls the position where a pattern is to be drawn within the sub-field area 53. A pattern is drawn on the sample 2 by the electron beam as follows. The stage 3 is moved in one direction, while the sample 2 is scanned by the electron beam, thus drawing a pattern in the sub-field area 53. While the stage 3 is moved in one direction, a pattern is drawn in each sub-field area 53. The frame area 52 includes a plurality of sub-field areas 53. After patterns are drawn in the frame area 52, the stage 3 is moved to the new frame area 52. Similarly, patterns are drawn in the new frame area 52. After patterns are completely drawn in the entire area of the sample 2 by the electron beam in this way, the sample 2 is replaced with a new sample. Patterns are drawn again on the new sample by the electron beam in a similar way. A description is given of how the controller 20 controls the drawing. The controller 20 reads the drawing data of the mask recorded in storage medium such as the magnetic disk 21. The controller 20 then temporarily stores the drawing data in the pattern memory 22 for each frame area 52. The drawing data for each frame area 52 stored in the pattern memory 22 is the frame information including the drawing position and the drawing figure data and the like. The pattern memory 22 sends the drawing data to the pattern data decoder 23 and the drawing data decoder 24 as the data analysis portion. The decoders 23 and 24 then send the data to the sub-field deflection calculator 30, the blanking circuit 25, the beam shape driver 26, the main deflection driver 27, and the auxiliary deflection driver 28. The pattern data decoder 23 creates, according to the drawing data, the blanking data. The decoder 23 then sends the blanking data to the blanking circuit 25. The decoder 23 also creates, according to the drawing data, the desired beam-shape data. The decoder 23 then sends the beam-shape data to the sub-field deflection calculator 30 and the beam shape driver 26. The sub-field deflection calculator 30 computes, according to the beam-shape data created by the pattern data decoder 23, the amount of deflection (moving distance) of the electron beam for each shot in the sub-field area 53. The calculator 30 then sends the computed information to the settling time setup unit 31. The settling time setup unit 31 includes a table 311 indicating a relationship between the subfield moving distance and the settling time as shown in line 101 in FIG. 3. The settling time setup unit 31 refers to the table 311 and uses the line 101 in FIG. 3 to determine the settling time corresponding to the subfield moving distance. The settling time setup unit 31 sends the determined settling time to the deflection controller 32 controlled by the controller 20. The controller 32 sends, synchronously with the pattern drawing, the settling time to one of the blanking circuit 25, the beam shape driver 26, the main deflection driver 27, and the auxiliary deflection driver 28. The beam shape driver 26 applies a certain deflection signals to the beam-dimension adjusting deflector 14 in the optical system 10. The deflection signal controls the dimension of the electron beam. The drawing data decoder 24 creates, according to the drawing data, subfield positioning data. The decoder 24 then sends the subfield positioning data to the main deflection driver 27. The main deflection driver 27 applies a certain deflection signal to the main deflector 15. The deflection signal deflects the electron beam to a certain subfield position. The drawing data decoder 24 generates, according to the drawing data, a control signal for scanning of the auxiliary deflector 16. The decoder 24 then sends the control signal to the auxiliary deflection driver 28. The auxiliary deflection driver 28 applies a certain auxiliary deflection signal to the auxiliary deflector 16. The auxiliary deflection signal allows the electron beam to be repeatedly irradiated onto the sub-field area 53 after the set settling time has elapsed. A pattern is thus drawn in the area 53. Thus, this embodiment allows the sub-field deflection calculator 30 and the settling time setup unit 31 to compute the moving distance of the electron beam in the drawing data in the sub-field area 53. Depending on the moving distance of the electron beam, this embodiment determines, referring to the table 311, the settling time so that the offset of the irradiation position of the electron beam has a certain value irrespective of any changes in the deflection amount of the auxiliary deflector. The throughput may thus be increased while maintaining the high drawing accuracy. Specifically, this embodiment may provide a shorter total settling time than when the settling time is set up corresponding to the subfield moving distance according to the line 102 in FIG. 3, i.e., when the subfield moving distance and the settling time are set up in a linear relationship. This embodiment may thus increase the throughput of the electron beam lithography system. A second embodiment of the present invention includes a method for evaluating a position accuracy in the electron beam lithography system. Specifically, the second embodiment includes a method for evaluating an irradiation position of the electron beam by the auxiliary deflector, i.e., a method for evaluating an irradiation position in the subfield. Referring to FIG. 4, this embodiment is described below. Referring to FIG. 4, the irradiation position of the electron beam by the auxiliary deflector is evaluated by irradiating the electron beam to an irradiation position in an evaluation pattern including two perpendicular lines A and B. As necessary, a surface of the sample to be evaluated may be applied with a photosensitive material in advance that is sensitive to the electron beam and changes its characteristics when exposed to the beam. Specifically, the electron beam is first irradiated to an irradiation position 101 at the intersection of the two lines A and B. The electron beam is then moved to an irradiation position 102 in the line A nearest to the position 101, and the electron beam is irradiated thereto (the first irradiation process). A constant settling time is set up to irradiate the electron beam. After the electron beam is irradiated to the irradiation position 102, the electron beam is moved to the next irradiation position 103 in the line B, and is irradiated thereto (the second irradiation process). The moving distance from the irradiation position 102 to the irradiation position 103 is usually longer than the distance from the position 101 to the position 102 in the first irradiation process. After the electron beam is irradiated to the irradiation position 103, the electron beam is moved to the next irradiation position 104 in the line B nearest to the position 103, and the electron beam is irradiated thereto (the third irradiation process). The moving distance from the irradiation position 103 to the irradiation position 104 substantially equals the distance from the position 101 to the position 102 in the first irradiation process. After the electron beam is irradiated to the irradiation position 104, the electron beam is moved to the next irradiation position 105 in the line A and is irradiated thereto (the fourth irradiation process). Similar operations are then repeated to irradiate the electron beam. Specifically, the electron beam is moved to the next irradiation position 106 in the same line A as the position 105 and is irradiated to the position 106 (the first irradiation process). The electron beam is then moved to the next irradiation position 107 in the line B and is irradiated thereto (the second irradiation process). The electron beam is then moved to the next irradiation position 108 in the same line B as the position 107 and is irradiated to the position 108 (the third irradiation process). The electron beam is then moved to the next irradiation position 109 in the line A and is irradiated thereto (the fourth irradiation process). After the above operations are performed on the entire evaluation pattern, the offset of the actual irradiation position of the electron-beam from the desired irradiation position is measured. The offset is used to evaluate an error. Specifically, after the sample is developed, the remaining or removed pattern of the photosensitive material is measured, comparing the pattern with the desired irradiation pattern, thus measuring the offset of the irradiation position of the electron beam. The offset from the desired irradiation pattern may thus be measured to determine the error. This embodiment alternately provides an irradiation position after a small deflection and an irradiation position after a large deflection, thus providing more accurate evaluation. The electron beam is irradiated after it is deflected for a certain settling time. If the previous irradiation is performed after insufficient settling time, the electron beam is irradiated to a position shifted from the correct irradiation position. In this case, the subsequent electron beam is irradiated after it is deflected relative to the previous irradiation position, thus gradually increasing the offset of the irradiation position. The settling time and the offset (error) of the irradiation position may thus not be determined accurately, thereby making it difficult to easily optimize the settling time. To describe specifically, a comparison is made by describing a method for evaluating an irradiation position by alternately irradiating the electron beam as in FIG. 5 in the evaluation pattern including the two perpendicular lines A and B as in FIG. 4. In this method, the electron beam is first irradiated to an irradiation position 201 at the intersection of the two lines. The electron beam is then moved to an irradiation position 202 in the line A adjacent to the position 201 and is irradiated thereto. The electron beam is then moved to an irradiation position 203 in the line B and is irradiated thereto. The electron beam is then moved to an irradiation position 204 in the line A and is irradiated thereto. The electron beam is then moved to an irradiation position 205 in the line B and is irradiated thereto. The electron beam is then moved to an irradiation position 206 in the line A and is irradiated thereto. The electron beam is then moved to an irradiation position 207 in the line B and is irradiated thereto. The electron beam is then moved to an irradiation position 208 in the line A and is irradiated thereto. The electron beam is then moved to an irradiation position 209 in the line B and is irradiated thereto. In this way, the electron-beam irradiation pattern is formed. This method alternately irradiates the beam to the irradiation positions in the lines A and B. The electron beam deflection thus gradually increases, as shown in FIG. 5, every time the electron beam is irradiated. With the constant settling time, therefore, the greater is the deflection, the greater is the offset of the irradiation position of the electron beam. The offset of the irradiation position includes the offset of the previous irradiation position. It is thus difficult to accurately determine the relationship between the settling time and the offset of the irradiation position. Specifically, it is difficult for this method to detect only the offset that depends on the settling time. Referring to FIG. 4, the evaluation method in this embodiment irradiates alternately and repeatedly the electron beam after a large deflection and the electron beam after a small deflection. The small-deflection electron beam is irradiated with little offset and a constant deflection, thereby giving no effect to the subsequent irradiation of the large-deflection electron beam. The relationship between the settling time and the error may thus be directly determined. FIG. 6 shows another example of this embodiment. The irradiation pattern corresponds to the irradiation pattern in FIG. 4 that is rotated by 45 degree. An irradiation pattern including the substantially perpendicular lines C and D as shown in FIG. 6 is irradiated with the electron beam as in FIG. 4. Specifically, after the electron beam is irradiated to an irradiation position 301 at the intersection, the electron beam is moved to an irradiation position 302 in the line C and is irradiated thereto (the first irradiation process). The electron beam is then moved to an irradiation position 303 in the line D and is irradiated thereto (the second irradiation process). The electron beam is then moved to an irradiation position 304 in the same line D as the position 303 and is irradiated thereto (the third irradiation process). The electron beam is then moved to an irradiation position 305 in the line C and is irradiated thereto (the fourth irradiation process). The electron beam is then moved to an irradiation position 306 in the same line C as the position 305 and is irradiated thereto (the first irradiation process). The electron beam is then moved to an irradiation position 307 in the line D and is irradiated thereto (the second irradiation process). The electron beam is then moved to an irradiation position 308 in the same line D as the position 307 and is irradiated thereto (the third irradiation process). The electron beam is then moved to an irradiation position 309 in the line C and is irradiated thereto (the fourth irradiation process). In this way, the electron-beam irradiation pattern is formed. As in FIG. 4, the following electron-beam motions are along the same line, and have short moving distances and small deflections: the electron beam motion from the irradiation position 301 to the irradiation position 302, the electron beam motion from the irradiation position 303 to the irradiation position 304, the electron beam motion from the irradiation position 305 to the irradiation position 306, and the electron beam motion from the irradiation position 307 to the irradiation position 308. The moving distances are substantially constant. A certain settling time may thus be set up to move the electron beam to the substantially accurate irradiation positions. The subsequent electron-beam deflections may thus give no effect to the following electron beam motions: the electron beam motion from the irradiation position 302 to the irradiation position 303, the electron beam motion from the irradiation position 304 to the irradiation position 305, the electron beam motion from the irradiation position 306 to the irradiation position 307, and the electron beam motion from the irradiation position 308 to the irradiation position 309. The directly correlation between the electron beam deflection and the error may thus be determined easily. In contrast, in FIG. 7, the irradiation patterns on the substantially perpendicular lines C and D as in FIG. 6 are irradiated with the electron beam as in FIG. 5. Specifically, after the electron beam is irradiated to an irradiation position 401 at the intersection, the electron beam is irradiated to the following positions in the following order, forming the electron-beam irradiation pattern: an irradiation position 402, an irradiation position 403, an irradiation position 404, an irradiation position 405, an irradiation position 406, an irradiation position 407, an irradiation position 408, and an irradiation position 409. The electron beam is thus alternately irradiated to the irradiation positions in the lines C and D. Referring to FIG. 7, the electron beam travels a different distance in each motion, and the electron beam deflection increases every time the electron beam is irradiated. With the constant settling time, therefore, the irradiation position of the electron beam gradually shifts from the original desired irradiation position. The subsequent electron beam is irradiated after it is deflected relative to the previous position. The error of the irradiation position of the electron beam thus includes the offset of the irradiation position of the previous electron beam. It is thus difficult to directly understand the relationship between the settling time and the error. The electron beam may thus be irradiated for evaluation using the methods in this embodiment shown in FIGS. 4 and 6, thereby easily determining the direct relationship between the settling time and the error. Because the direct relationship between the settling time and the error is obtained, it is possible to easily optimize the settling time. Thus, an electron beam lithography system and a method for evaluating the same according to preferred embodiments of the present invention have been described in detail, but the present invention is not limited to the disclosed embodiments, and other embodiments may also be possible.
052992411	summary	BACKGROUND OF THE INVENTION The present invention relates to a technology for transmuting transuranium elements and more particularly to a transuranium transmuting reactor core for transmuting the transuranium elements at a fast reactor and also to a transuranium elements transmuting fuel pin and fuel assembly charged into a reactor core of a fast reactor. A spent fuel discharged from a thermal reactor such as boiling water reactor or the like includes transuranium elements (hereinafter called TRU elements) such as neptinium-237 (.sup.237 Np), americium-241 (.sup.241 Am), americium-243 (.sup.243 Am), curium-242 (.sup.242 Cm), curium-244 (.sup.244 Cm) and others which are high-level radioactive wastes, and in minor actinides (hereinafter called MA elements) present after eliminating plutonium (Pu) from the TRU elements, there exists elements such as .sup.237 Np, .sup.241 Am, .sup.243 Am or the like having an extremely long half life such as 2.14 million Years, 432 years, 7,380 years, which cannot be quenched within a short period of time. Thus, it is desired that the MA elements are transformed into elements with a short half life through a nuclear transmutation in a short period of time. A prior art includes technique for transmuting the TRU element which comprises using a fast reactor extremely high in a neutron energy as compared with a thermal reactor and subjecting the TRU elements charged into a fuel charged in a core of the fast reactor to a nuclear transmutation ((1) "Conceptional Design Study on Actinide burning Fast Reactor", T. Osugi et al., JAER1-M 83-217, issued by Japan Atomic Energy Research Institute in December 1983; (2) "Transmutation of Transuranics in FBR", A. Sasahara, T. Matsumura, F7, Fall Meeting Reports, Atomic Energy Society of Japan, 1988). The prior art TRU elements transmuting comprises transmuting the aforementioned MA elements by causing a transmutation shown in FIGS. 9A to 9C to the typical MA elements of .sup.237 Np, .sup.241 Am and .sup.243 Am which are main objects of transmuting at a fast reactor core. In FIGS. 9A to 9C, F.P. denotes fission products, and elements given in a square border around indicates that of being easy to cause a fission against a neutron energy in the fast reactor, namely, that its energy averaged fission cross-sections are about 1 burn or over. The prior art TRU elements transmuting process utilizes a feature of the fast reactor core effectively, and the feature comes in: (1) Since a neutron energy of the fast reactor core is high, a neutron capture is hard to occur in .sup.237 Np, .sup.241 Am and .sup.243 Am and the like, and thus an evil influence of the fast reactor on a neutron economy according to the charging of the TRU elements into the reactor core is relatively small (a neutron capture cross-section getting small according as the neutron energy becomes high as shown in FIG. 18). (2) The fast reactor is generally high by about 1 digit in a neutron flux level as compared with the thermal reactor, therefore the TRU elements can be subjected to a nuclear transformation even if a fission and neutron capture cross section on an energy average is small, and thus a high transmuting efficiency of the TRU elements is ensured. In the prior art transmuting of the TRU elements, nothing has been taken particularly into consideration for charging amount of the TRU elements charged into a fast reactor core and its distribution in core when carrying out a transmuting of the TRU elements. Still, however, only a self-evident technical care on charging the core with the TRU elements as much as possible has been considered for enhancing a transmuting efficiency of the TRU elements. However, if the fast reactor core is charged with the TRU elements as much as possible, then the following problems are capable of resulting therefrom. (1) If the MA elements to be transmuted is added to uranium-plutonium mixed fuel, a melting point of the mixed fuel lowers. Then the melting point drop is capable of causing a fuel melting, thus a measure such as lowering a reactor power or the like will be necessary for avoiding the fuel melting, which may deteriorate the transmuting efficiency of the MA elements. (2) As will be apparent from FIGS. 9A to 9D, the typical MA elements to be transmuted is generally hard to bring about a fission, and hence is transformed into fissionable elements by a neutron capture. Accordingly, if the fast reactor core is charged with the TRU element excessively much, then, as shown in FIG. 19, an amount of fissionable elements produced newly by the neutron capture of the MA element according to a neutron irradiation comes to exceed fissionable elements transmuted by fission, thus an excess reactivity of the fast reactor increasing. Consequently, if the charging amount of the TRU elements and its distribution are not specified properly, an excessive change or distortion may arise on a reactor power distribution and a neutron flux distribution, thus leading to problems on safety and characteristics of the reactor. (3) The TRU elements to be transmuted are easy to cause an alpha-decay in most cases, and an alpha ray energy emitted at the time of the alpha-decay is relatively high at 4 to 6 MeV generally. Accordingly, if the MA elements are added much to a fuel, a calorific value and a source intensity of gamma ray, neutron and others become excessive from the state of a fresh fuel before loading into the fast reactor core. Further, at the time of assembling, storage and transportation of new fuel assemblies in which the MA elements are enclosed, a heat removing of the alpha ray energy becomes difficult and the fuel overheats to lead to a failure in a worst case. (4) When charging a fast reactor uniformly with the TRU elements to be transmuted at the core with a core for which a plutonium enrichment is one kind as a base, a radial distribution of the power density, namely a radial power distribution during operation of the reactor becomes small according as it comes outside, as shown in FIG. 20, therefore a transmuting efficiency of the TRU element and a plant power generation efficiency being unsatisfying. (5) When charging the reactor uniformly with the TRU elements at the core with the fast reactor core for which a plutonium enrichment is two or more than two kinds as a base, a radial power distribution of the core is improved as compared with FIG. 20 by an adjustment of the plutonium enrichment, a flatting requirement can thus be satisfied, however, as shown in FIG. 21, for example, there arises a portion where the power distribution largely fluctuates according to burn-up. On the other hand, a flow rate of a coolant flowed for cooling down the fast reactor core is constant through the lifetime of a reactor plant. The flow rate of the coolant to fuel assemblies is set adaptively to the time when the power is maximized. Thus, when the output distribution fluctuates largely according to the burn-up of the fuel, a heat removing efficiency deteriorates, a heating efficiency gets lowered furthermore, which is not preferable from the viewpoint of an economical operation of the reactor plant. SUMMARY OF THE INVENTION An object of the present invention is to substantially eliminate defects or drawbacks encountered in the prior art described above and to provide a transuranium elements transmuting reactor core capable of transmuting the TRU elements efficiently without causing a failure of the fuel assemblies, increase of excess reactivity, deterioration of thermal efficiency and others. Another object of the present invention is to provide a transuranium element transmuting fuel and fuel assembly capable of preventing the lowering of the power density of a fast reactor and the distortion of the power distribution of the fast reactor and effectively transmuting the TRU elements. These and other objects can be achieved according to the present invention by providing, in one aspect, a transuranium element transmuting reactor core in which a reactor is charged with a plurality of fuel assemblies at a core and an amount of a transuranium element to be added is controlled so as to prevent a fuel element contained in the fuel assemblies from melting, and in the improvement, the amount of the transuranium elements to be added to the fuel assemblies is controlled so as to keep an excess reactivity of the reactor substantially zero through an operation of the reactor. In another aspect, there is provided a transuranium element transmuting reactor core in which a reactor is charged with a plurality of fuel assemblies at a core and an amount of a transuranium element to be added is controlled so as to prevent a fuel element contained in the fuel assemblies from melting, and in the improvement, charging amounts of .sup.242 Cm, .sup.244 Cm and .sup.241 Am are set so as to satisfy an equation EQU 1.2.times.10.sup.2 .times.M.sub.242 +2.8.times.M.sub.244 +1.1.times.10.sup.-1 .times.M.sub.241 <Q.sub.1 where an upper bound of heating rates of the single fuel assembly outside the reactor is.sub.1 Q from the view point of the fuel assembly integrity, charging amounts of .sup.242 Cm, .sup.244 Cm and .sup.241 Am and also satisfy an equation EQU 1.2.times.10.sup.2 .times.M.sub.242.sup.L +2.8.times.M.sub.244.sup.L +1.1.times.10.sup.-1 .times.M.sub.241.sup.L 21 Q.sub.2 where an upper bound of the heating rates, per unit length of the fuel pellet contained in the fuel pins is Q.sub.2 from the view point of the fuel element integrity, charging amounts of .sup.242 Cm, .sup.244 Cm and .sup.241 Am per the unit length are M.sub.242.sup.L, M.sub.244.sup.L and M.sub.241.sup.L. In a further aspect, there is provided a transuranium element transmuting reactor core in which a reactor is charged with a plurality of fuel assemblies at a core and an amount of transuranium elements to be added is controlled so as to prevent a fuel element contained in the fuel assemblies from melting and in the improvement, a charging density of minor actinides is set to lessen outwards of a core central portion in a core area where a plutonium content is made even. In a still further aspect, there is provided a transuranium element transmuting reactor core in which a reactor is charged with a plurality of fuel assemblies at a core and an amount of transuranium elements to be added is controlled so as to prevent a fuel element contained in the fuel assemblies from melting and in the improvement, a charging density of minor actinides is set high accordingly in an area where a plutonium is enriched high at the core of a plutonium enriched area where a plutonium content varies. In a still further aspect, there is provided a transuranium element transmuting fuel pin wherein a transuranium fuel pin is formed by charging a transuranium fuel material in a fuel clad and the transuranium fuel material includes at least one of fuel materials consisting of an enriched uranium and a uranium-plutonium mixed fuel and a fertile material consisting of a degraded uranium, a natural uranium and a depleted uranium contain transuranium elements such as Np, Am and Cm. In a still further aspect, there is provided a transuranium element transmuting fuel assembly including a wrapper tube and a plurality of fuel pins enclosed in the wrapper tube, each of the fuel pins including a fuel clad, wherein at least one part of the fuel pins are formed by charging a transuranium fuel material in the fuel clad with a transuranium fuel material inside. In a preferred embodiment, the fuel pins enclosed in the wrapper tube comprises transuranium fuel pins charged with the transuranium fuel material and fuel material pins charged with a fuel material consisting of an enriched uranium and a uranium-plutonium mixture fuel, and a radioactive fission product such as Sr or alkaline metals is contained in the transuranium fuel material. In the transuranium element transmuting reactor core according to the present invention, since an amount of transuranium elements to be added to a fuel pin of the fuel assemblies is controlled so as to keep an excess reactivity of the reactor substantially zero through an operation of the reactor, a decrease of effective multiplication factor according to the lapse of time for operation will be prevented, an excessive deterioration or turbulence of the reactor power distribution can be prevented, and as looking for improvement of a power plant capacity factor from enhancing a reliability of the plant, transuranium elements (TRU elements) can be transmuted efficiently. Further, from setting loading amounts of .sup.242 Cm, .sup.244 Cm and .sup.241 Am so as to realize: EQU 1.2.times.10.sup.2 .times.M.sub.242 +2.8.times.M.sub.244 +1.1.times.10.sup.-1 .times.M.sub.241 <Q.sub.1 where an upper bound of the single fuel assembly power assembly outside the reactor is.sub.1 Q from the view point of the fuel assembly integrity, loading amounts of .sup.242 Cm, .sup.244 Cm and .sup.241 Am which can be loaded into the single fuel assembly are M.sub.242, M.sub.244 and M.sub.241, and also to realize: EQU 1.2.times.10.sup.2 .times.M.sub.242.sup.L +2.8.times.M.sub.244.sup.L +1.1.times.10.sup.-1 .times.M.sub.241.sup.L <Q.sub.2 where an upper bound of the heating per unit length of the fuel element contained in the fuel assemblies is Q.sub.2 from the view point of the fuel element integrity, charging amounts of .sup.242 Cm, .sup.244 Cm and .sup.241 Am per the unit length are M.sub.242.sup.L, M.sub.244.sup.L and M.sub.241.sup.L, a melting of the fuel element during operation of the reactor and an overheating or failure of the fuel assemblies outside the reactor can effectively be prevented, and an accident of a control rod and a neutron absorbing material of the control rod can be reduced by a neutron absorption effect of .sup.242 Cm, .sup.244 Cm and .sup.241 Am, an enhancement of heat removing efficiency of the core can thus be realized, an economical operativity is also improved, and a safety and reliability of the core and the fuel assemblies are ensured as well, thus transmuting the TRU elements efficiently. Further, by setting a charging density of minor actinides to lessen outwards of a core center in a core area where a plutonium content is even, and also by setting a charging density of minor actinides high accordingly in an area where Pu is enriched high at the core of a Pu-enriched area where a plutonium content varies, a flatting requirement of a radial distribution of the reactor power can be satisfied, an enhancement of safety and reliability of the core and the fuel assemblies will be realized without causing the excessive deterioration and turbulence of the reactor power distribution, thus transmuting the TRU elements efficiently. In a further aspect, according to the transuranium element transmuting fuel assembly of the characters described above, even if the transuranium fuel material is charged in the transuranium fuel pin, the degradation of the core power density and the distortion of the core axial power distribution can be effectively prevented, thus improving the core cooling efficiency and effectively transmuting the transuranium element.
047524324	claims	1. A sytem for the production of nitrogen-13 atoms, said system including: means for producing a proton beam which travels along a preselected path; target means containing a slurry target material of powdered carbon-13 in a liquid, said target means being positioned in said path of said proton beam whereby subjection of said target material to said proton beam produces nitrogen-13 atoms in a desired chemical form; means for dissipating heat generated during said production of said nitrogen-13 atoms; means for retaining said powdered carbon-13 in said target means; means for separation of said nitrogen-13 atoms in said chemical form for collection of desired purified product containing said nitrogen-13 atoms; and means for transporting said nitrogen-13 atoms in said chemical form from said target means to said means for separation of said desired purified product containing nitrogen-13 atoms. a delivery water inlet tube for supplying an inflow of water from a water source to said target material; a recovery water outlet tube, providing an outflow of water from said target material, whereby said nitrogen-13 ammonium ions are transported from said target material to said means for separation; and frit means for preventing said powdered carbon-13 from passing into said inlet tube and said outlet tube. means for producing a proton beam which travels along a preselected path, said means for producing said proton beam comprising a cyclotron; target means containing a slurry target material, said target means being positioned in said path of said proton beam whereby subjection of said target material to said proton beam produces nitrogen-13 ammonium ions in an aqueous solution, said target material constituting an aqueous slurry of carbon-13 powder and water, means for retaining said carbon-13 powder in said target means; means for dissipating heat generated during said production of said nitrogen-13 ammonium ions; means for separating said nitrogen-13 ammonium ions for collection of purified product containing nitrogen-13 ammonium ions; and means for transporting said nitrogen-13 ammonium ions in said aqueous solution from said target means to said means for separation of said nitrogen-13 ammonium ions. a delivery water inlet tube for supplying an inflow of water from a water source to said target material; a recovery water outlet tube for providing an outflow of water from said target material, whereby said nitrogen-13 ammonium ions are transported from said target material to said means for separation; and frit means for preventing said carbon-13 powder from passing into said inlet tube and said outlet tube. positioning a target material within a target holder, said target material consisting essentially of a slurry of carbon-13 particles in water, said water containing oxygen-16; irradiating said target material within said target holder with a beam of protons to produce nitrogen-13 atoms by a reaction of said protons with said carbon-13 and with said oxygen-16; removing heat from said target holder during said irradiation of said target material with said protons; passing water through said target material, without removing said carbon particles, to remove said nitrogen-13 atoms from said target material as nitrogen-13 ammonium ions and nitrogen-13 oxides in said water passing through said target material; and collecting said nitrogen-13 ammonium ions contained in said aqueous solution. positioning a target material in a target holder, said target material consisting essentially of a water slurry of carbon particles, said carbon particles enriched in an isotope of carbon selected from carbon-12 and carbon-13; irradiating said target material within said target holder with a beam of energetic particles selected from deuterons for a slurry of carbon-12 and protons for a slurry of carbon-13 to product nitrogen-13 atoms through the reaction of deuterons with carbon-12 and protons with carbon-13, respectively; removing heat from said transfer holder during said irradiation of said target material with said energetic particles; passing water through said target material, without removing said carbon particles, to remove said nitrogen-13 atoms from said target material as nitrogen-13 ammonium ions and nitrogen-13 oxides in an aqueous solution; separating said nitrogen-13 ammonium ions from said nitrogen-13 oxides; and collecting said nitrogen-13 ammonium ions. 2. The system of claim 1 wherein said means for producing said proton beam along a preselected path comprises a cyclotron. 3. The system of claim 1 wherein said liquid is natural water and subjection of said target material to said proton beam produces nitrogen-13 ammonium ions in aqueous solution. 4. The system of claim 1 wherein said target means positioned in path of said proton beam is a target material constituting carbon-13 powder and water in aqueous slurry whereby said subjection of said target material to said proton beam produces nitrogen-13 ammonium ion in aqueous solution. 5. The system of claim 3 wherein said means for dissipating heat generated during production of said nitrogen-13 ammonium ions comprises a cooling water supply and a circulating system for cooling said target means. 6. The system of claim 3 wherein said means for transporting said nitrogen-13 ammonium ions in aqueous solution from said location of said target means comprises: 7. A system for the production of nitrogen-13 ammonium ions, said system including: 8. The system of claim 7 wherein said means for dissipating heat generated during production of said nitrogen-13 ammonium ions comprises a cooling water supply and a circulating system for removing heat from said target means. 9. The system of claim 7 wherein said means for transporting said nitrogen-13 ammonium ions in aqueous solution from said location of said target means comprises: 10. A process for the production of nitrogen-13 ammonium ions in an aqueous solution, which comprises: 11. The process of claim 10 further comprising separating said nitrogen-13 oxides from said nitrogen-13 ammonium ions after removing said nitrogen-13 atoms from said target material and prior to said collecting of said nitrogen-13 ammonium ions. 12. The process of claim 10 further comprising maintaining a pressure upon said target material during said irradiation and during said passing of water through said target material to maintain a dense slurry and thereby enhance efficiency of said irradiation of said slurry by said protons. 13. A process for the production of nitrogen-13 ammonium ions in an aqueous solution, which comprises:
	abstract	A nuclear reactor includes a pressure vessel, and a control rod assembly (CRA) including at least one movable control rod, a control rod drive mechanism (CRDM) for controlling movement of the at least one control rod, and a coupling operatively connecting the at least one control rod and the CRDM. The coupling includes a connecting rod engaged with the CRDM and a terminal element connected with a lower end of the connecting rod and further connected with the at least one control rod. In some embodiments the terminal element includes a first portion comprising a first material having a first density and a second portion comprising a second material having a second density that is greater than the first density. In some embodiments the terminal element has a largest dimension parallel with the connecting rod that is greater than or equal to a largest dimension transverse to the connecting rod.
	claims	1. An assembly intended to be inserted into a nuclear reactor, comprising:an assembly hollow body, of elongate shape along a longitudinal axis X, a wall of the hollow body comprising at least one through-opening that extends through the wall;an assembly element inserted at least in part into the hollow body, the assembly element comprising at least one flexible blade of which a free end is shaped into a clip-fastening hook collaborating in clip-fastening with the through-opening from inside the hollow body, so as to connect the assembly element to the hollow body; andat least one removable means for locking the at least one flexible blade clip-fastened into the through-opening, each of the at least one removable means for locking including a locking screw making it possible to prevent the at least one flexible blade from flexing and thus lock the connection between the assembly element and the hollow body. 2. The assembly according to claim 1, wherein the clip-fastening hook is produced by a thickening of the free end of the at least one flexible blade. 3. The assembly according to claim 1, wherein the hollow body is of hexagonal cross section and each face of the hexagon comprises one through-opening, and the at least one flexible-blade includes a flexible blade clip-fastened into each of the through-openings. 4. The assembly according to claim 1, wherein the locking screw of each of the at least one removable means for locking, when in a position in which the locking screw is screwed into the at least one flexible blade, makes it possible to prevent the at least one flexible blade from flexing and from becoming unclipped. 5. The assembly according to claim 1, wherein each of the at least one flexible blade is produced by cutting into a thickness of the assembly element. 6. The assembly according to claim 5, wherein the wall of the assembly hollow body comprises at least one open-ended bore designed to allow the locking screw to pass from the outside of the hollow body and a screw head to be housed. 7. The assembly according to claim 1, wherein each of the at least one flexible blade is attached and fixed to the assembly element by a fixing screw. 8. The assembly according to claim 7, wherein each fixing screw is welded to the assembly element and/or to the at least one flexible blade in its screwed-in position. 9. The assembly according to claim 8, wherein the assembly element comprises at least one cavity in which the clip-fastening hook can become lodged when the at least one flexible blade is in the flexed position, the locking screw being screwed through the clip-fastening hook and housed in the cavity when the at least one flexible blade is clipped in, so as to prevent the at least one flexible blade from flexing. 10. The assembly according to claim 1, wherein the assembly is a nuclear fuel assembly, the hollow body being a central portion forming a casing configured to clad fuel pins, the assembly element being an upper neutron shield (UNS) device or the assembly element forming an upper portion forming a gripper head of the assembly. 11. The assembly according to claim 1, wherein the nuclear reactor is a liquid sodium-cooled fast neutron reactor. 12. The assembly according to claim 1, wherein the assembly element includes a shoulder and, when the assembly element is inserted into the assembly hollow body, a top of the assembly hollow body abuts the shoulder of the assembly element. 13. A method for assembling the assembly according to claim 1, wherein the assembly comprises at least two of the flexible blades, the method comprising:a/ inserting the assembly element into the assembly hollow body by a translational movement, so as to achieve simultaneous flexing of the at least two flexible blades toward the inside of the hollow body, lowering of the assembly element down inside the hollow body until the at least two flexible blades return to their position with their hooks clip-fastened into the corresponding through-openings of the hollow body so as to connect the hollow body to the assembly element; andb/ locking each of the at least two flexible blades clip-fastened into the through-opening using the at least one removable means for locking. 14. A method for dismantling the assembly according to claim 1, wherein the assembly comprises at least two of the flexible blades, the method comprising:a1/ unlocking each of the at least two flexible blades, the hook of each of which is clip-fastened into the corresponding through-opening, by removing the at least one removable means for locking;b1/ applying a radial force to the hook of each of the at least two flexible blades from the outside of the through-opening, so as to cause the at least two flexible blades to flex simultaneously toward the inside of the hollow body, unclipping the hooks; andc1/ extracting the assembly element from inside the assembly hollow body via a translational movement. 15. The dismantling method according to claim 14,wherein the assembly element comprises a plurality of the flexible blades and the hook of each of the flexible bladesis individually clip-fastened into the corresponding through-opening of the hollow bodyand locked,step b1/ being performed by simultaneous actuation of actuators, and arranged individually facing one of the through-openings.
	description	This application claims priority to U.S. Provisional Patent Application No. 61/555,375, filed on Nov. 3, 2011, the entire contents of which are hereby incorporated by reference. The present invention relates to fuel channels for nuclear reactors. More particularly, the invention relates to annulus spacers for use in fuel channels having concentric tubes (e.g., a calandria tube with an internal pressure tube) of a nuclear reactor, an example of which is the CANDU (“CANada Deuterium Uranium”) reactor. By way of example only, the CANDU reactor is a heavy water, or light water cooled and heavy-water moderated fission reactor capable of using fuels composed of natural uranium, other low-enrichment uranium, recycled uranium, mixed oxides, fissile and fertile actinides, and combinations thereof. Annulus spacers (AS), often provided as garter springs (GS), are used in CANDU reactors to maintain an annular gap between two tubes of a fuel channel assembly, such as an inner pressure tube (PT) and an outer calandria tube (CT) as mentioned above. The PTs are located inside the reactor CTs that insulate the PTs from the heavy water moderator in the calandria. The annular gap between the PTs and CTs is typically filled with an annulus gas. In some cases, four annulus spacers are used per fuel channel assembly, each at a specified axial position along the length of the fuel channel. It is important that the spacers are in their correct positions, as incorrect positioning may lead to contact between the hot PT and cooler CT. Such contact is unacceptable. In some embodiments, a fuel channel assembly consists of a PT, two end fittings and associated hardware, wherein the PT is connected to the two end fittings by a mechanical roll-expanded joint. A bellows assembly rolled into the fuelling machine side tubesheet and welded to the bellows attachment ring can be used to seal the annulus at both ends. Therefore, in some fuel channel assembly embodiments, there is no direct access to the annular space between the PTs and CTs. One known type of annulus spacer is a close-coiled helical spring. For example, such a spring can have a 4.83 mm (0.190 inch) outside diameter, can be formed into a torus using a Zircaloy-2 girdle wire, and can be formed from Zirconium alloy wire of square cross-section (e.g., 1.02×1.02 mm (0.040×0.040 inch)). The spacers can prevent direct contact between the PTs and CTs, which would be undesirable because of the increased susceptibility to blister formation as local hydrogen concentration increases due to deuterium ingress. In some embodiments, there are four spacers in each channel assembly spaced approximately 1.02 m (40 inches) apart and located in a manner offset towards the outlet end of the fuel channel assembly. The position of each of the annulus spacers is important to ensure that they meet a variety of functional, performance, safety, environmental and inter-facing system requirements. Some annulus spacers are loose-fitting spacers provided with a garter spring and a girdle wire held within an annular cavity formed by the coiled wire of the garter spring. The girdle wire can enable a position of the annulus spacer along the fuel channel to be detected using eddy current testing (ECT) techniques (i.e., based upon the fact that the girdle wire can be made from a material that forms a loop of continuous conductivity). Other annulus spacers are tight fitting, and can have a spring tension which draws them tight onto the outside surface of the PT. In many cases, eddy current technology cannot positively locate tight fitting spacers because the design of tight fitting spacers does not include a welded girdle wire. Other challenges with loose- and tight-fitting spacers exist, as do challenges to identifying the locations of such spacers along the axial length of fuel channel assemblies. In one aspect, the invention provides an apparatus for detecting the location of at least one annulus spacer between concentric interior and exterior tubes when a temperature gradient is present therebetween. A probe head assembly is movable within the interior tube. At least one temperature sensor is coupled to the probe head assembly and configured to detect a temperature of an interior surface of the interior tube. A drive assembly is operable to move the probe head assembly relative to the interior tube. A data acquisition system is coupled to the at least one temperature sensor and configured to receive a plurality of temperature measurements in order to identify at least one position along the interior surface having a temperature abnormality corresponding to a reduced temperature gradient. In another aspect, the invention provides a method for detecting the location of at least one annulus spacer between concentric interior and exterior tubes having a temperature gradient therebetween. A probe head assembly including at least one temperature sensor is inserted into the interior tube. The temperature of an interior surface of the interior tube is detected at a plurality of locations along the interior surface. At least one position is identified along the interior surface having a temperature abnormality corresponding to a reduced temperature gradient. Other aspects of the present invention will become apparent by consideration of the detailed description and accompanying drawings. Before any embodiments of the invention are explained in detail, it is to be understood that the invention is not limited in its application to the details of construction and the arrangement of components set forth in the following description or illustrated in the accompanying drawings. The invention is capable of other embodiments and of being practiced or of being carried out in various ways. For example, the devices and methods disclosed herein are introduced and described in the context of application in a nuclear fuel channel, and to detect garter springs or other annulus spacers. However, the devices and methods disclosed herein can be utilized in other applications and environments, and to detect other objects besides annulus spacers. By way of example only, the devices and methods can be used in downhole drilling applications to detect objects adjacent or around drilling equipment, in pipelines to detect the locations of adjacent equipment, structural supports, other objects, and in other applications as desired. FIG. 1 is a perspective of a reactor core of a CANDU-type reactor, known as a CANDU 6, by way of example only. A generally cylindrical vessel, known as a calandria 10, contains a heavy-water moderator. The calandria 10 has an annular shell 14 and a tube sheet 18 at a first end 22 and second end 24. A number of fuel channel assemblies 28 pass through the calandria 10 from the first end 22 to the second end 24. As illustrated in FIG. 2, in each fuel channel assembly 28 a calandria tube 32 forms a first boundary between the heavy water moderator of the calandria 10 and structure within the CT 32. A pressure tube (PT) 36 forms an inner wall of the fuel channel assembly 28. The PT 36 provides a conduit for reactor coolant and fuel assemblies 40. The CT 32 and the PT 36 form a concentric tube assembly, and an annulus space 44 is defined by a gap between the PT 36 and the CT 32. The CT 32 constitutes an exterior tube of the pair of concentric tubes and the PT 36 constitutes an interior tube of the pair of concentric tubes. The annulus space 44 is normally filled with a circulating gas, such as dry carbon dioxide, nitrogen, air or mixtures thereof. The annulus space 44 and gas can be part of an annulus gas system. The annulus gas system has two primary functions. First, a gas boundary between the CT 32 and PT 36 provides thermal insulation between hot reactor coolant and fuel within the PTs 36 and the relatively cool CTs 32. Second, the annulus gas system provides indication of leaking calandria tubes, pressure tubes 36, or their connections via the presence of moisture in the annulus gas. A series of annulus spacers 48 is disposed between the CT 32 and PT 36 (i.e., between an exterior surface 36E of the PT 36 and an interior surface 32I of the CT 32). One such spacer is shown in FIG. 3. Functionally, the annulus spacers 48 maintain the prescribed gap between the PT 36 and its corresponding CT 32, while allowing the passage of annulus gas through and around the annulus spacer 48. Other functions of the annulus spacer 48 include accommodating relative axial movement between the PT 36 and CT 32 while limiting wear, scratches, deformation, or damage to the PTs 36 and CTs 32. Although annulus spacers 48 are usually designed to generally limit heat transfer from the PT 36 to the heavy-water moderator during normal operating conditions of the reactor, the annulus spacers 48 provide the only points of potential conduction directly between the PT 36 and the CT 32 between the two ends of a fuel channel assembly 28. As the annulus spacer 48 is located in the fuel channel annulus space 44, its temperature can be influenced by either the hot PT 36 (approximately 300° C. during normal reactor operation in some embodiments) or the cooler CT 32 (approximately 80° C. during normal reactor operation in some embodiments), depending on which component the annulus spacer 48 is contacting. If the annulus spacer 48 is in contact with both tubes, it will likely experience a temperature gradient between the temperatures of the PT 36 and the CT 32. With continued reference to FIG. 3, the annulus spacer 48 is received within the annulus space 44 between the PT 36 and CT 32 as discussed above. Although the annulus spacer 48 is shown in FIG. 3 as a loose fitting spacer including a garter spring 52 and a girdle wire 56, the annulus spacer 48 may be a tight fitting spacer having no girdle wire, or can take a number of other forms suitable for maintaining the desired space between the PT 36 and CT 32. In some embodiments, the garter spring 52 is formed from a length of coiled wire 60. Two ends 64, 68 of the coiled wire 60 can be connected so that the garter spring 52 forms a toroid 72. If provided, the girdle wire 56 can be held within an annular cavity 80 formed by the coiled wire 60 of the garter spring 52. The garter spring 52 can be dimensioned to fit tightly around the PT 36, and in some embodiments is resilient so that it may be expanded to a dimension greater than an outside diameter of the PT 36 during installation, yet fit tightly and securely once positioned. The garter spring 52 may be formed from a nickel-chromium based alloy such as INCONEL X-750. In other embodiments, the garter spring 52 may be formed of other alloys, including a zirconium-based alloy such as ZIRCALOY or a zirconium-niobium-copper alloy. In still other embodiments, the garter spring 52 may be formed of an alloy including, but not limited to, a combination of zirconium, niobium, and copper. Detecting the position of each annulus spacer 48 is necessary in many applications in order to verify the location of the annulus spacer 48, and to thereby ensure that the annulus spacer 48 meets a variety of functional, performance, safety, environmental, and inter-facing system requirements. Embodiments of the present invention provide an apparatus and a method of locating annulus spacers 48 via thermal analysis or profiling of one or both of the CT 32 and the PT 36 by detecting a response of each annulus spacer 48 to a temperature gradient present between the CT 32 and the PT 36. In some embodiments, the location of a tight fitting garter spring installed on the PT 36 may be discerned within a tolerance of about 15 mm or better without the use of eddy current testing. FIGS. 4-25 illustrate a lab-based detection apparatus 100 and method, according to certain embodiments of the present invention, for identifying the position of annulus spacers 48 between a CT 32 and a PT 36 outside of a calandria 10. The CT 32 is suspended between the centre lattice sites of two 3×3 lattice site stands 104 (FIG. 4). A first end of the apparatus 100 includes an end fitting 108 complete with a clearance fit pressure tube hub. This end fitting does not feature a liner tube. A liner tube protective sleeve can be used in the first end fitting to retain the pressure tube and to provide a uniform tool delivery surface. Four garter springs 52 or other annulus spacers 48 are provided between the PT 36 and the CT 32. The garter springs 52 can be tight-fitting on the PT 36 in some embodiments. It should also be noted that more or less than four annulus spacers 48 can be provided, including zero, and the number need not be known prior to carrying out the detection method, which can effectively determine both the number and position of annulus spacers 48. Again referencing the apparatus 100 of FIG. 4, the exterior surface 32E of the CT 32 is tightly wrapped with a heater 112, such as a BriskHeat FE heating cable (FIG. 5), and can be wrapped in a helical manner along the CT 32. The pitch spacing of the heating cable wrap can be approximately 2.5″, for example. A heater control box 116 (FIG. 6) can be mounted at the first end of the CT 32, or in any other suitable location. The installed heating cable 112 can be covered with insulation 120, such as 6-inch diameter (Iron Pipe Size), 2-inch thick fiberglass pipe insulation with ASJ (All Service Jacket) as shown in FIG. 7. The insulation 120 may be provided in 3-foot long sections. A thermostat bulb 124 can be installed at a location (e.g., the midpoint along the CT 32) on the exterior surface 32E of the CT 32, through the insulation 120, as shown in FIG. 8. The thermostat bulb 124 can be coupled with the heater control box 116 of FIG. 6 to maintain the exterior surface 32E of the CT 32 at a predetermined set point temperature above ambient. A digital thermometer 128 can also be used to enable an operator to visually monitor the temperature of the exterior surface 32E of the CT 32 as shown in FIG. 9. Regardless of its construction, the thermometer 128 confirms that the exterior surface 32E of the CT 32 is heated to a predetermined temperature above ambient. This may be carried out as an initial step in an annulus spacer locating method in a laboratory setting, and furthermore, can simulate heating of the exterior surface 32E of the CT 32 (e.g., from moderator liquid in a functional reactor). An indicator light 132 can be installed on the heater control box 116 as shown in FIG. 10 to provide a visual indication of when the heater 112 is turned on and off. A probe head assembly 136 according to an embodiment of the present invention is illustrated in FIG. 11. As shown in FIGS. 12A and 12B (detail views of FIG. 11), some embodiments of the probe head assembly 136 include inboard and outboard centering assemblies 140, although a single centering assembly can be used in other embodiments. Any or all of the centering assemblies 140 of the probe head assembly 136 can be provided with one, two, or more wheels 142 that contact and roll along the interior surface 36I of the PT 36 to help in smooth travel and guidance of the probe head assembly 136 as it travels axially within the PT 36. Examples of such wheels 142 are shown in FIGS. 12A and 12B. In one configuration, referred to as the circumferential (absolute) configuration of the probe head assembly 136, at least one radial temperature sensor array 144 is provided. A plurality of spring-loaded (e.g., biased radially outward) guide blocks 146 are provided around the circumference of the probe head assembly 136 at multiple axial positions (e.g., 2 positions in the illustrated construction). Each of the guide blocks 146 can carry one or more temperature probes 150. In the illustrated construction, each of the radial array of guide blocks 146 is provided with a single temperature probe 150 so that, in total, the guide blocks 146 at each axial position provide a radial array of temperature probes 150. The temperature probes 150 may be thermistor or thermocouple probes in any configuration (e.g., thermistor probes on one assembly, thermocouple probes on another, a mix of thermistor and thermocouple probes on either or both assemblies, and the like. In some embodiments by way of example, each temperature probe 150 is set at 0.002 inch past the outer surface of the corresponding guide block 146 to ensure contact of the temperature probe 150 with the interior surface 36I of the PT 36 (see FIGS. 13 and 14). The outer surface of the guide blocks 146 can be generally flat as shown in FIG. 13, or curved about the central axis, as shown in FIG. 14. The guide blocks 146 include tapered outward end surfaces 146A to facilitate easier insertion into to the PT 36. In the illustrated apparatus, the probe head assembly 136 is provided with first and second pluralities of spring-loaded guide blocks 146 each associated with a centering assembly 140. In this setup, the two sets of guide blocks 146 (and corresponding probes 150) are axially spaced from each other. As indicated above, in other constructions, a single array of temperature probes 150 may be provided in other embodiments of the present invention. As described in further detail below, the radial array(s) of temperature probes 150 in some embodiments may be conveyed or driven axially to additional section(s) of the PT 36 to obtain data on the temperature profile the additional section(s), or the entire interior surface 36I, of the PT 36. FIGS. 15 and 16 illustrate a probe head assembly 236 of an alternate construction of the present invention, referred to as the axial probe head assembly. In this construction, multiple temperature probes 250 are provided in an axial array, with each temperature probe 250 positioned at a unique axial position. In the illustrated construction, all of the temperature probes 250 are provided in alignment at a common radial position in a row. All of the temperature probes 250 in the illustrated embodiment of FIGS. 15 and 16 are mounted to a carrier bar 252 (e.g., a plexiglass bar), which is coupled in a spring-loaded manner (e.g., biased radially outward) to a perforated tubular body 254 of the probe head assembly 236. In the illustrated construction, the carrier bar 252 is securely coupled between a pair of guide blocks 246, which are spring-biased with respect to the tubular body 254. The guide blocks 246, like the guide blocks 146 of FIGS. 13 and 14, include tapered outward end surfaces 246A. The temperature probes 250 can be secured to the carrier bar 252 individually with corresponding fasteners 256 (e.g., set screws). Each temperature probe 250 can be positioned to make contact with the interior surface 36I of the PT 36 when the probe head assembly 236 is inserted into the PT 36. The row of temperature probes 250 can contain thermocouples, thermistors, or a combination of both in any desired arrangement. As described in further detail below, the row of temperature probes 250 may be rotated about the axis of the PT 36 to obtain data on the temperature profile of the interior surface 36I of the PT 36, or at least a lengthwise section thereof. The probe head assembly 250 can then be conveyed or driven to additional section(s) of the PT 36 to obtain data on the temperature profile of the additional section(s), or the entire interior surface 36I, of the PT 36. Alternatively or in addition, the system and method may utilize one or more non-contact temperature sensors 800 or “thermal imaging devices” to collect data representative of the temperature of various locations on the interior surface 36I of the PT 36. For example, one or more infrared cameras 800 may be provided on the probe head assembly 136, 236 and inserted into the PT 36 and operated to detect the temperature profile thereof, and to output a corresponding electrical signal representative thereof. The probe head assembly 236 of FIG. 19 is illustrated with one non-contact temperature sensor 800 for illustrative purposes. FIGS. 17 and 18 illustrate a water jacket 300 for use with either of the probe head assemblies of FIGS. 11-16. The illustrated water jacket 300 includes a central support spool 302 and a coil 304 of water tubing for conveying cooling water between an inlet 306 and an outlet 308 of the water jacket 300. The water jacket 300 can have an outer diameter about the same size as, or slightly larger than, the interior diameter of the PT 36 so that the coil 304 of water tubing is pressed upon the interior surface 36I of the PT 36. The water jacket 300 can be supplied with chilled water (e.g., supplied axially down the center of the pressure tube) to produce a localized cooling effect at or adjacent the probe head assembly 136, 236. Thus, the water jacket 300 can act as an active heat sink to increase the temperature gradient between the interior surface 36I of the PT 36 and the exterior surface 32E of the CT 32 (which may be heated) as described above immediately prior to or during the temperature detection of the PT 36. For example, if the water jacket 300 is attached to a distal end of the probe head assembly 136, chilled water running through the coil 304 of water tubing can cool the interior surface 36I of the PT 36 at the location of the water jacket 300, after which time the probe head assembly 136 can be axially moved along the PT 36 to bring the temperature probes 150 of one of the arrays into an axial position adjacent the cooled section of the PT 36 in order to perform the temperature measurements of the interior surface 36I of the PT 36 in the cooled section. Alternatively, the water jacket 300 can be located sufficiently close to (e.g., immediately adjacent one or more temperature probes 150) to generate a similar cooling effect of the PT 36 without subsequent movement of the probe head assembly 136 to take the above-described temperature measurements. Of course, the water jacket 300 can be used with similar effect with the probe head assembly 236 of FIGS. 15 and 16. It should be understood that the water jacket 300 may convey water or any other suitable heat exchange fluid. FIG. 19 illustrates an alternate apparatus to actively cool the interior surface 36I of the PT 36. A cold air generator 400 (e.g., a “ColdGun”, available from EXAIR Corporation) may be used to provide a supply of air chilled to a predetermined temperature below ambient. The chilled air can be blown into the inside of the PT 36 around and/or through the probe head assembly 136, 236. For example, FIG. 20 illustrates a fitting 404 for inserting an outlet tube 408 of the cold air generator 400 into the interior of the probe head assembly 236 of FIGS. 15 and 16. The tubular body 254 of the probe head assembly 236 is perforated through the sides, via apertures 259, and includes an open distal end 254A to allow passage of the chilled air into contact with the interior surface 36I of the PT 36. As shown in FIGS. 21 and 22, the tubular body 254 of the probe head assembly 236 can be configured for different axial lengths corresponding to different axial-length temperature probe constructions. The temperature probes 250 or non-contact temperature sensor(s) 800 are not shown here for simplicity. When provided in a short-length configuration, the end of the tubular body 254 can be closed (e.g., an end cap 258 is provided over the distal end 254A) so that localized cooling is provided via the apertures 259. When provided in a long-length configuration, the distal end 254A of the tubular body 254 can be open (e.g., the end cap 258 is removed) for large area pressure tube cooling. The chilled air can be provided through the distal end 254A instead of or in addition to the apertures 259. As shown in FIG. 23, a push rod 504 for guiding the probe head assembly 136, 236 through the PT 36 can include sections of tubing (e.g., four 7-foot lengths of ABS tubing, for example). A drive system 500 can also be provided to drive the probe head assembly 136, 236 through the PT 36 (via the push rod 504). In one construction, the drive system 500 can include two threaded (e.g., ⅜-16) driving rods 508 together with an associated frame 512 and clamps 516 (FIG. 24). In a basic configuration, the probe head assembly 136, 236 can be driven manually (by hand via cranks 520 coupled to the driving rods 508) or by a hand held tool such as a cordless drill, and the probe head assembly 136, 236 can be driven a relatively steady speed through the PT 36 between temperature measurements or as the temperature probes 150, 250 are measuring the temperature of the interior surface 36I of the PT 36. However, in other embodiments, the probe head assembly 136, 236 can be driven with a computer-controlled drive system 500′ (including one or more electric motors 524) to provide predictable, programmable drive speed. In this way, position data can be provided to a data acquisition system (DAS) along with temperature data from the temperature probes 150, 250 to generate a coordinated output representing the temperature profile of the interior surface 36I of the PT 36. As shown in FIG. 25, signals representative of measured temperatures can be routed via temperature probe leads 528 through a GEC junction box 532 to a computer (e.g., a laptop computer 536), used to capture, store, process and display the acquired data. Signals representative of position data (e.g., from a motor encoder) can be routed back to the computer 536 directly or indirectly via motor leads 540. The temperature and/or position data may be displayed in real time in some configurations. By creating a temperature gradient between the exterior surface 32E of the CT and the interior surface 36I of the PT 36 and obtaining a temperature profile of the interior surface 36I of the PT 36, the position of each annulus spacer 48 can be detected. This can be accomplished in the above-described embodiments of the present invention by observing localized temperature variations in the area of each annulus spacer 48. The above-described test setup sets forth several specific structures and methods for determining the positions of annulus spacers 48 in a particular laboratory setting, but the invention is not limited to such constructions, and the invention may be carried out in a variety of ways with a variety of equipment. Temperature profiling can be used to detect the position of an annulus spacer 48 in a fuel channel assembly, which may be adapted to other specific tooling assemblies, test setups, and/or processes for either a laboratory setting (e.g., prior to assembly of a fuel channel assembly into a reactor) a functional reactor, a shut down reactor or a laid up reactor. In working on a functional reactor or shutdown reactor, the temperature gradient between the exterior surface 32E of the CT 32 and the interior surface 36I of the PT 36 can be generated by heated moderator liquid outside the CT 32, and/or by optionally cooling the interior surface 36I of the PT 36 (e.g., in any of the above-described manner, or in another manner) when fuel 40 has been removed from the PT 36. In some embodiments, the thermal gradient detected by the apparatus and methods described and illustrated herein can be generated by moderator liquid that is colder than the interior surface 36I of the PT 36—a state that can be accomplished by permitting the moderator to cool sufficiently (or even actively cooling the moderator liquid) and/or by heating the interior surface 36I of the PT 36 in a similar manner as described above in connection with cooling devices. It should be clear that the present invention contemplates both a “passive” location method, which does not require the implementation of an artificial heat source or heat sink to enhance the thermal gradient between the PT 36 and the CT 32, and an “active” location method, which takes advantage of a separate heat source and/or a heat sink to generate or enhance a temperature gradient and therefore to enhance the ability to detect the locations of the annulus spacers 48 (i.e., solely for the purpose of locating the annulus spacers 48). In the passive method, the temperature reading (recording)/mapping instruments map the temperature variation (changes) on the interior surface 36I of the PT 36 to identify, through the localized temperature anomaly or abnormality (where the temperature of the interior surface 36I of the PT 36 is closest to the temperature of the exterior surface 32E of the CT 32, in other words, a reduced temperature gradient), the effect of the “heat leak” between various parts of the fuel channel assembly. Active methods may include the implementation of the continuous or specifically pulsed “heat source” or continuous or specifically pulsed “heat sink” inside the PT 36, before performing the temperature recoding/mapping. As in the passive method, the temperature reading (recording)/mapping instruments map the temperature variation (changes) on the interior surface 36I of the PT 36 to identify, through the localized temperature, the effect of the “heat leak” between various parts of the fuel channel assembly. However, in an active method, the “heat leak” is enhanced by localized pre-heating or pre-cooling of the interior surface 36I of the PT 36. A location of contact or near-contact between the PT 36 and the CT 32 through an annulus spacer 48 will be presented as the localized temperature gradient detected by one or more of the temperature probes 150, 250 (or non-contact temperature sensors) described above, such as at different circumferential positions (e.g., at the 6 o'clock or nearby position), or continuously around the circumference of the PT 36. It should also be understood that the concept of thermal mapping to locate annulus spacers 48 can be effective regardless of which one of the inside and the outside is the “hot” side. In other words, it is possible to obtain good results by heating the interior surface 36I of the PT 36 to a temperature in excess of the temperature of the exterior surface 32E of the CT 32. Furthermore, any method of thermal mapping disclosed herein can be supplemented by monitoring of the ultrasonic test (UT) (shear and/or longitudinal) velocity change in the “heat leak” area, which (ultrasonic velocity change) will be an effect of the localized temperature difference corresponding to the location of contact between the annulus spacer 48 and the PT 36 Although the description above is in connection with the detection of annulus spacers between concentric tubes of a fuel channel assembly in a nuclear reactor, it will be appreciated that the present invention can be used in other applications to detect the location of an object outside of a tube (and capable of thermal detection by a thermal gradient relative to the tube). Such objects need not necessarily encircle the tube or otherwise be annular in shape, and need not necessarily be located in a space between concentric tubes. Accordingly, aspects of the present invention find application in pipeline repair and maintenance, drilling systems, and the like.
061954047	description	DESCRIPTION OF THE PREFERRED EMBODIMENT FIG. 1 shows the apparatus of the invention. It shows a container for radioactive material seen in cross-section and the detail of another cross-section view of two adjacent chambers according to the invention. (1) is the thick metal barrel, usually made of steel, that constitutes the body of the container in cavity (2) in which the radioactive material is stored. The container is handled using trunnions (3) that are fastened onto barrel (1). A plurality of chambers (4) according to the invention cover the outer surface of barrel (1). Each chamber mainly comprises a metal section (5) that is hollow and closed along its length and has 4 long surfaces. Cooling fins (6) are located on the outer surface of said sections. Internal surface (7) is molded to the outer surface of barrel (1) such that it ensures perfect thermal contact. Inside each chamber a lead plate (8) is installed in contact with internal surface (7). The plate is covered with a protective steel sheet (9). The assembly consisting of section (5) and plates (8, 9) is fastened to barrel (1) using bolts (11). The remaining space inside section (5) is filled with a resin (10) that constitutes a neutron shield. Contained within (12) a passage tube may be seen that enables fastening from the outer surface of section (5) through the resin when said resin has been poured in section (5) before said section is fastened.
055436153	summary	BACKGROUND OF THE INVENTION 1. Field of the Invention This invention relates to a beam charge exchanging apparatus for converting a fast charged particle beam such as an ion beam into a neutral particle beam, the apparatus being employed in the manufacture of semiconductors or the like. 2. Prior Art Conventionally, an apparatus shown in FIG. 7, which is known as a gas cell, has been widely used as a charge exchanging apparatus. The gas cell includes a gas container 6, which is placed in a vacuum container and includes holes 16 through which a charged particle beam is passed. The charged particle beam 1 which is passed through the gas container 6 collides with a gas 3, which is introduced in the gas container through a gas introducing pipe (not shown), whereby charges are exchanged between the charged particle beam and the gas. There is another conventional apparatus called a metallic vapor cell which employs an alkaline metallic vapor such as Li as a gas for the cell. In this apparatus, the charged particle beam 1 is passed through the vapor so as to cause the charged particle to collide with the metallic vapor, whereby charges are exchanged therebetween. In the conventional gas cell, however, since gas 3 is introduced into the gas container 6 located in the vacuum container, the holes 16 of the gas container 6 should be made small to prevent the vacuum from deteriorating due to the diffusion of gas from the gas cell. For this reason, it has been impossible to obtain many gas particles or charged particles in the gas cell. The metallic vapor cell has a similar drawback in that the quantity of beams introduced in the container is limited if the vacuum level is to be maintained has another drawback in that the container body is damaged by the chemically active alkaline metal. According to the charge neutralizing method based on a molecular beam as disclosed in Japanese Patent Laid-Open Appln. No. 5-129096 which has recently been published, the charge exchanging efficiency is low because the number of molecules is limited and, therefore, it is difficult to accomplish a charge exchange for a large quantity of beams. Furthermore, using the aforesaid method to change a large quantity of ion beams into neutral beams requires a large apparatus. Hence, it has been difficult to obtain high-energy neutral particle beams effectively, and the highest possible energy value which could be obtained using this method is limited to a few KeV. Furthermore, in the conventional charge exchanging apparatus, there was no means for immediately measuring the quantity of neutral particles after charge exchange. SUMMARY OF THE INVENTION Therefore, this invention has been made to solve the above problems, and it is an object of the present invention to provide an apparatus which is capable of exchanging the charges of a large quantity of high-energy beams with another media with high efficiency. It is an another object of the present invention to provide an apparatus which is capable of exchanging the charges of a large quantity of high-energy beams with another media with high efficiency and of measuring the quantity of obtained neutral particles. In order to accomplish the first object, the present invention provides a beam charge exchanging apparatus including a gas fluid container disposed in a vacuum and having holes for passing allowing a particle beam to pass through the container; a source of gas; and means for introducing the gas/fluid into the container at a high speed so that the fast particle beam will collide with the high speed gas/fluid in the container and a charge exchange will occur therebetween that converts the particle beam into a neutral particle beam. In order to accomplish the second object, the apparatus further includes means for detecting the quantity of neutral particles resulting from the exchange of charges by detecting the quantity of ionized fluid generated as an electric current. According to the present invention as described above, a charged particle beam is passed through a high speed gas/fluid having directivity in the container so as to convert the charged particle beam into a neutral particle beam. Therefore, the problem in which the high speed gas/fluid diffuses out of the container through the holes can be prevented. Hence, it is possible to make the holes of the container larger to increase the area of contact between the gas/fluid and the charged particle beam within the container, whereby a highly efficient charge exchange can be achieved, leading to the production of higher energy beams. As the charged particle beam is passed through the high speed gas/fluid, the high speed gas/fluid obtains surplus charges resulting from the charge exchange. Accordingly, the high speed gas/fluid is charged, i.e. is ionized. The ionized gas/fluid can be collected on an ion capturing electrode and the quantity of ions in the gas/fluid can be measured by measuring the current flowing through the electrode. Thus, the quantity of the neutral particles in the neutral beam can be measured. The above and other objects, features and advantages of the present invention will become more apparent from the following description when taken in conjunction with the accompanying drawings in which preferred embodiments of the present invention are described by way of illustrative examples.
	summary
046831088	abstract	An apparatus and method for under water remote replacement of the screws which secure together first and second structures in the internal region of a nuclear reactor core, wherein the screws are received and counterbored screw bores in the first structure. First, one or more lateral recesses are machined in the side wall of the counterbore portion. Then an external hex head screw with a shoulder flange is inserted in the screw bore and is threadedly engaged with the second structure with the flange seated in the counterbore portion. A locking cup with a central hex opening is fitted over the hex head of the bolt in the counterbore portion. Then a bifurcated die is seated against the first structure with die fingers extending into the cup, and a drive member is driven through the die coaxially with the screw to move the die fingers against the side wall to deform it into the recesses, thereby locking the assembly in place.
052572986	description	DESCRIPTION OF THE PREFERRED EMBODIMENTS Hereinafter, manufacturing processes according to one embodiment of the present invention will be described with reference to FIG. 1. First, a sintering agent was manufactured in the following manner. Specifically, Al.sub.2 O.sub.3 of about 40 wt % and SiO.sub.2 of about 60 wt % were coarsely mixed. The mixture thereof was exposed to a mixed gas flow of 8%-H.sub.2 /N.sub.2, and was heated up to about 2100.degree. C., and then melted. Thereafter, the melt was cooled, and homogeneous aluminum silicate was obtained. The thus obtained aluminum silicate was then ground. As a result, uniform powder was obtained as a sintering agent. By use of this sintering agent, nuclear fuel pellets were manufactured in accordance with processes shown in FIG. 1. Specifically, the sintering agent was added to UO.sub.2 powder, and then Gd.sub.2 O.sub.3 powder was mixed. Further, a lubricant (stearic acid, polyethylene glycol and the like) was added, as shown in FIG. 1. This mixture was compacted by an uniaxial press and then green pellets were obtained. The adding amount of the sintering agent was about 30 through about 500 ppm, and the adding amount of the Gd.sub.2 O.sub.3 powder was about 10 wt %, both with respect to the total amount of UO.sub.2, Gd.sub.2 O.sub.3 and the above-described sintering agent. Next, the thus obtained green pellets were processed in a degreasing process. Thereafter, the green pellets were sintered in a humid hydrogen atmosphere at about 1760.degree. C. for about 5.6 hours. In some cases, the lubricant-mixing process and the degreasing process may be omitted. The sintering density and average grain diameter of the thus manufactured nuclear fuel pellets were compared with those of the nuclear fuel pellets obtained by use of the sintering agent of adding amount of 0.25 wt %. The results of comparison are as follows: ______________________________________ sintered average grain adding amount of density diameter sintering agent (g/cm.sup.3) (.mu.m) ______________________________________ 30 ppm 10.40 27.2 70 ppm 10.34 25.6 130 ppm 10.39 25.7 250 ppm 10.37 32.7 500 ppm 10.33 33.3 0.25 wt % 10.20 24.7 ______________________________________ As can be seen from the comparison results, the sintered densities of the pellets manufactured in accordance with this embodiment are significantly higher than that of the pellet manufactured with addition of the sintering agent by 0.25 wt %. Further, the grain diameters of the pellets of this invention are also increased. For the sake of comparison, FIG. 2 and FIG. 3 respectively show the microstructure of each nuclear fuel pellet after polished and chemically etched, when observed by a microscope. Specifically, FIG. 2 shows a UO.sub.2 -10 wt % Gd.sub.2 O.sub.3 nuclear fuel pellet including the sintering agent of 30 ppm according to this embodiment. FIG. 3 shows conventional a UO.sub.2 -10 wt % Gd.sub.2 O.sub.3 without a sintering agent. As can be seen from FIG. 2 and FIG. 3, in the pellet of this embodiment (FIG. 2), the portions of free UO.sub.2 phase, which are indicated by the shaded portions (blue portions in the actual microphotograph), are much smaller (2 vol % at a maximum) than those shown in FIG. 3. In this embodiment according to the present invention, a humid hydrogen gas was used as a sintering atmosphere. However, a mixed gas of carbon monoxide and carbon dioxide may also be used as a sintering atmosphere. Further, in this embodiment, aluminum oxide and silicon oxide were mixed and melted, and then aluminum silicate was obtained as a sintering agent. However, besides this, the mere mixed powder of aluminum oxide and silicon oxide may also be used as a sintering agent. Further, the mixture of alkoxides such as aluminum isopropoxide, tetraethyl orthosilicate, etc. and precursors such as aluminum hydroxide, aluminum stearate and the like may also be used as a sintering agent. As described above, when nuclear fuel pellets are manufactured in accordance with the manufacturing method of this invention, a series of phenomena occur in the following manner. Specifically, a sintering agent becomes a single liquid phase during the sintering, and through this liquid phase, Gd.sub.2 O.sub.3 is dispersed into the entire pellets. Thus, the effective inter-diffusion distances between UO.sub.2 and Gd.sub.2 O.sub.3 become smaller, so that the generation of solid-solution phase is promoted. Further, a liquid phase-sintering mechanism promoters reaction between particles, so that the growth of grains is promoted. This growth of grains increases the diffusion distance between FP gas and grain boundaries. Thus, a FP gas release rate from the pellets decreases. In the present invention, the mixing proportion of a sintering agent (consisting of aluminum oxide and silicon oxide) is determined to be about 10 ppm through about 500 ppm with respect to the total amount of nuclear fuel pellets. This is based on that the following facts have been confirmed. Specifically, the adding amount of the sintering agent must be 10 ppm at a minimum in order to improve the solid-solution state (i.e., to obtain homogeneous pellets) while a free UO.sub.2 phase is maintained to be 5% at a maximum. Further, the grain-growth-promoting effect reaches a maximum when the adding amount of the sintering agent is about 250 ppm. Further, when the adding amount of the sintering agent exceeds 500 ppm, this is not only insignificant but also decreases the pellet density. Further, if the sintering agent includes Al.sub.2 O.sub.3 of more than 60 wt %, the grain-growth-promoting effect decreases. As described above, according to the present invention, the solid-solution state (i.e., homogeneous state) of nuclear fuel pellets can be improved. Further, the creep characteristics of nuclear fuel pellets can also be improved by grain boundaries softened by glassy phase in spite of large grain diameter. Therefore the nuclear fuel pellets of this invention can decrease a FP gas release rate, and can also improve the PCI resistance, whereby burnup extensin toward higher levels of the nuclear fuel can be achieved. Obviously, numerous additional modifications and variations of the present invention are possible in light of the above teachings. It is therefore to be understood that within the scope of the appended claims, the invention may be practiced otherwise than as specifically described herein.
	description	The present disclosure relates to nuclear reactors and, more specifically, fuel assemblies and fuel supports for supporting the fuel assemblies in a reactor core. The statements in this section merely provide background information related to the present disclosure and may not constitute prior art. A nuclear reactor pressure vessel (RPV) has a generally cylindrical shape and is closed at both ends, e.g., by a bottom head and a removable top head. A top guide is spaced above a core plate within the RPV. A core shroud, or shroud, surrounds the core plate and is supported by a shroud support structure. Particularly, the shroud has a generally cylindrical shape and surrounds both the core plate and the top guide. The top guide includes several openings, and fuel assemblies are inserted through the openings and are supported by the core plate. The core plate includes a flat plate supported by a plurality of beams. A nuclear reactor core includes a plurality of individual fuel assemblies that have different characteristics that affect the strategy for operation of the core. For example, a nuclear reactor core typically has several hundred individual fuel assemblies that have different characteristics, each fuel bundle having a plurality of fuel rods. The fuel assemblies are arranged within the reactor core so that the interaction between the fuel assemblies satisfies regulatory and reactor design guidelines and constraints. In addition, the core arrangement determines the cycle energy, which is the amount of energy that the reactor core generates before the core needs to be refreshed with new fuel elements. The core loading arrangement preferably optimizes the core cycle energy. A core cycle is determined from one periodic reactor core refueling to a second reactor core refueling. During the course of the cycle of operation, the excess reactivity, which defines the energy capability of the core, is controlled in two ways. Specifically, a burnable poison, e.g., gadolinia, is incorporated in the fresh fuel. The quantity of initial burnable poison is determined by design constraints typically set by the utility and by the NRC. The burnable poison controls most, but not all, of the excess reactivity. A second way is through the manipulation of control rods within the core. Control rods control the excess reactivity. Specifically, the reactor core contains control rods, which assure safe shutdown and provide the primary mechanism for controlling the maximum power peaking factor. The total number of control rods available varies with core size and geometry, and is typically between 50 and 269 in a reactor core. The position of the control rods, i.e., fully inserted, fully withdrawn, or somewhere between, is based on the need to control the excess reactivity and to meet other operational constraints, such as the maximum core power peaking factor. Coolant is introduced in the core to cool the core and to be transitioned into steam as a working fluid for energy generation. Normal coolant flow enters the fuel assemblies as a single phased flow with slightly sub-cooled coolant. The flow approaches the fuel support vertically upward and then turns horizontally as the flow enters the inlet to a fuel support supporting a fuel assembly. The flow then passes through an orifice of the fuel support to provide a pressure drop to assist coolant distribution to the fuel assemblies. The flow then turns vertical and enters the lower tie plate of the fuel assembly and is distributed around the individual fuel rods of the fuel assembly. It is known that reactor core design can be varied by design and layout of the control rods within the fuel assembly lattice. Often, the fuel assembly lattice is configured with differently configured fuel rods such that the fuel assembly has a defined orientation. The core is designed with a plurality of oriented fuel assemblies to improve the performance and operation of the reactor. However, fuel assemblies typically have a round shaped tie plate configured for mating with a round hole or orifice defined by the fuel support. The tie plate's rod shaped end includes a lumen for receiving fluid flow from the fuel support. Current tie plates, fuel assemblies, and fuel supports do not provide any capabilities to ensure that the fuel assemblies are installed onto the fuel supports in the orientation within the core as designed and specified. Orientation of the fuel assemblies are the responsibility of the fuel assembly installation personnel based on a visual inspection. As such, a typical problem encountered with reactor design implementation is errors due to fuel assemblies being installed having an incorrect orientation which is commonly referred to as rotated bundle error. As identified by the inventors hereof, an improved assembly and method for eliminating or at least minimizing rotated bundle error would be desirable. The inventors hereof have succeeded at designing assemblies and methods for reduction, and possibly completely eliminating, rotated bundle error due to the improper orientation of fuel assemblies within the core during fuel assembly installation. The assemblies and their use include a mating configuration between fuel assembly and particularly the tie plate of the fuel assembly, and the fuel support to ensure an orientation of the fuel assembly relative to the fuel support. Such improved fuel assemblies and fuel supports can provide for the reduction and potentially elimination of rotated bundle error. According to one aspect, an assembly for mounting a fuel assembly in a nuclear reactor includes a fuel assembly including fuel rods within the fuel assembly and having a lower tie plate with a fuel assembly mating fixture and a fuel support including a fuel support mating fixture. The fuel support mating fixture is constructed to selectively engage the fuel assembly mating fixture during installation of the fuel assembly onto the fuel support for providing a predefined orientation of the fuel assembly to the fuel support. According to another aspect, a fuel assembly in a nuclear reactor includes an elongated body defining a top end portion defining an upper aperture, a bottom end portion defining a lower aperture, and a cavity. A plurality of elongated fuel rods are positioned within the cavity in a predetermined pattern having a predetermined orientation and positioned between the top end portion and the bottom end portion. A lower tie plate is attached to the bottom end portion of the body including an orifice for receiving fluid flow into the cavity through the lower aperture. The lower tie plate includes a mating fixture structured to define an orientation of the lower tie plate relative to the fuel rods within the cavity. According to yet another aspect, a fuel support for a nuclear reactor includes a body having an upper end and a lower end portion and an aperture positioned about the upper body end and structured to selectively receive a fuel assembly in a predetermined orientation relative to the body. According to still another aspect, a method of mounting a fuel assembly in a nuclear reactor includes inserting a plurality of fuel rods in a pattern within a fuel assembly having a predetermined orientation relative to an orientation of a lower tie plate of the fuel assembly, and installing the lower tie plate of the fuel assembly into an aperture of a fuel support. The fuel assembly includes a predetermined orientation for receiving the lower tie plate. The method also includes installing by selectively aligning the orientation of the fuel assembly with the orientation of the fuel support. Further aspects of the present disclosure will be in part apparent and in part pointed out below. It should be understood that various aspects of the disclosure may be implemented individually or in combination with one another. It should also be understood that the detailed description and drawings, while indicating certain exemplary embodiments, are intended for purposes of illustration only and should not be construed as limiting the scope of the disclosure. It should be understood that throughout the drawings, corresponding reference numerals indicate like or corresponding parts and features. The following description is merely exemplary in nature and is not intended to limit the present disclosure or the disclosure's applications or uses. In some embodiments, an assembly for mounting a fuel assembly in a nuclear reactor includes a fuel assembly including fuel rods within the fuel assembly. The fuel assembly includes a lower tie plate with a fuel assembly mating fixture and a fuel support including a fuel support mating fixture. The fuel support mating fixture is constructed to selectively engage the fuel assembly mating fixture during installation of the fuel assembly onto the fuel support for providing a predefined orientation of the fuel assembly to the fuel support. In other embodiments, an assembly for mounting a fuel assembly in a nuclear reactor includes a fuel assembly including fuel rods within the fuel assembly. A lower tie plate with a fuel assembly mating fixture and a fuel support with a fuel support mating fixture are provided. The fuel support mating fixture is constructed to selectively engage the fuel assembly mating fixture during installation of the fuel assembly onto the fuel support for providing a predefined orientation of the fuel assembly to the fuel support. These and other embodiments can be better understood with reference to the figures. As seen by way of the exemplary operating environment of FIG. 1, a conventional boiling water reactor (BWR) has a reactor pressure vessel 10 and a core shroud 12 arranged concentrically in the reactor pressure vessel 10 with an annular region, namely, the downcomer annulus 14, therebetween. It should be understood that while a BWR is depicted here, the present disclosure applies to other types of nuclear reactors as well. In the BWR, the core shroud 12 is a stainless steel cylinder surrounding the nuclear fuel core 13. In particular, the core shroud 12 comprises a shroud head flange 12a for supporting the shroud head (not shown); a circular cylindrical upper shroud wall 12b having a top end portion welded to shroud head flange 12a; an annular top guide support ring 12c welded to the bottom end portion of upper shroud wall 12b; a circular cylindrical middle shroud wall welded assembly 12d welded to the top guide support ring 12c; and an annular core plate support ring 12e welded to the bottom of the middle shroud wall 12d and to the top of a lower shroud wall 12f. As seen in FIG. 1, the shroud 12 is vertically supported by a plurality of shroud support legs 16, each of the latter being welded to the bottom head 17 of the reactor pressure vessel 10. The core shroud 12 is laterally supported by an annular shroud support plate 18, which is welded at its inner diameter to the core shroud 12 and at its outer diameter to the reactor pressure vessel 10. The shroud support plate 18 has a plurality of circular apertures 20 in flow communication with diffusers of a plurality of jet pump assemblies (not shown), The fuel core 13 of a BWR consists of a multiplicity of upright and parallel fuel bundle assemblies 22 arranged in arrays. Each fuel assembly 22 includes an array of fuel rods inside a fuel channel made of zirconium-based alloy. Each array of fuel bundle assemblies 22 is supported at the top by a top guide 24 and at the bottom by a core plate 26 and its underlying support structure 27. The core plate 26 subdivides the reactor into the fuel core 13 and a lower plenum 15. The core top guide 24 provides lateral support for the top of the fuel assemblies 22. The core plate 26 provides lateral support for the bottom of the fuel assemblies 22. This lateral support maintains the correct fuel channel spacing in each array to permit vertical travel of a control rod 28 including the control rod blades 29 between the fuel assemblies 22. The power level of the reactor is maintained or adjusted by positioning the control rods 28 up and down within the core 13 while the fuel bundle assemblies 22 are held stationary. Each control rod 28 has a cruciform cross section consisting of four wings or control rod blades 29 at right angles. Each blade 29 consists of a multiplicity of parallel tubes welded in a row with each tube containing stacked capsules filled with neutron-absorbing material. Each control rod 28 is raised or lowered with the support of a control rod guide tube 30 by an associated control rod drive 33 which can be releasably coupled by a spud at its top to a socket in the bottom of the control rod 28. The control rod drives 33 are used to position control rods 28 in a BWR to control the fission rate and fission density, and to provide adequate excess negative reactivity to shutdown the reactor from any normal operating or accident condition at the most reactive time in core life. Each control rod drive 33 is mounted vertically in a control rod drive housing 32 that is welded to a stub tube 34, which in turn, is welded to the bottom head 17 of the reactor pressure vessel 10. The control rod drive 33 is a double-acting, mechanically latched hydraulic cylinder. The control rod drive 33 is capable of inserting or withdrawing a control rod 28 at a slow controlled rate for normal reactor operation and of providing rapid control rod 28 insertion (scram) in the event of an emergency requiring rapid shutdown of the reactor. The control rod drive housing 32 has an upper flange that bolts to a lower flange of the control rod guide tube 30. Each control rod guide tube 30 sits on top of and is vertically supported by its associated control rod drive housing 32. The uppermost portion of the control rod guide tube 30 penetrates a corresponding circular core plate aperture 35 in the core plate 26. There can be more than one hundred and forty (140) control rod guide tubes 30 penetrating an equal number of circular core plate apertures 35 (of the core plate 26), each core plate aperture 35 typically has a diameter slightly greater than the outer diameter of the control rod guide tube 30. The control rod drive housings 32 and control rod guide tubes 30 have two functions: (1) to house the control rod drive 33 mechanisms and the control rods 28, respectively, and (2) to partially support the weight of the fuel in the fuel assemblies 22. The fuel weight is reacted at an orifice of a fuel support 36 that is positioned on the core plate 26 and underlying support structure 27 as well as on top of the control rod guide tube 30. The control rod guide tubes 30 and rod drive housings 32 act as columns carrying the weight of the fuel. During operation of the reactor, water in the lower plenum 15 enters an inlet of the fuel support 36. The water is channeled within the fuel support 36 to a lumen of a lower tie plate of the fuel assembly 22. The water continues to rise in the fuel assembly 22 and in the fuel core 13, with a substantial amount turning to steam that may be used in the production of electrical energy. Referring now to FIG. 2, one exemplary embodiment illustrates a fuel support 36 mounting in a core plate aperture 35 of core plate 26. Two fuel assemblies 22 (denoted as 22A and 22B) are positioned for mounting in one of the fuel support apertures 40 of the fuel support 36. As known to those skilled in the art, a fuel support 36 typically supports four fuel assemblies 22 and therefore has four fuel support apertures 40. However, only two are shown in FIG. 2 for simplicity and ease of depiction. The fuel support 36 includes a control rod chamber 42 for receiving and retaining a control rod 28 and its blades 29 (a single blade is shown in FIG. 2 by way of example). The fuel support 36 is typically coupled to the control rod guide tube 30 for channeling the control rod. The fuel support 36 receives coolant from the lower plenum either directly or indirectly such as through a portion of the control rod guide tube 30. A lumen 46 is attached to each aperture for channeling fluid (denoted by C and the arrows) through the fuel support 36 and into a lower tie plate 48 of each fuel assembly 22. After entering the tie plate 48, the fluid travels through the fuel assembly 22 and about fuel rods 50 formed in a lattice 51 contained therein. The lattice 51 can be designed to have a plurality of different fuel rods 50 for providing a desired performance including having a single design orientation. Generally, the fuel assembly 22 has an elongated body that defines a top end portion 60 with an upper aperture 62, a bottom end portion 64 defining a lower aperture 66, and a cavity 68. The bottom end portion 64 and the lower aperture 66 can be defined by the lower tie plate 48 of the fuel assembly 22. The lower tie plate 48 can include one or more fuel assembly mating fixture 52 for coupling to and engaging with a fuel support mating fixture 54 within the fuel support aperture 40. The fuel assembly mating fixture 52 can be positioned about a radial from an axis defined by the fuel assembly 22 or the lower tie plate 48 for providing a particular orientation with the fuel rods 50 and the lattice 51 within the fuel assembly 22. The mating fixtures 52 and 54 are each designed for mating to each other and provide a predefined and limited orientation relative to each other. In such a manner, the fuel assembly 22 can only be fully installed in the fuel support aperture 40 of the fuel support 36 in the predefined orientation. The fuel assembly mating fixture 52 and fuel support mating fixture 54 are shown in FIG. 2 as being a slot and a hole, respectively. However, it should be understood that each could be any shape capable of mating to the other in order to achieve a predefined orientation relative to each other. In some embodiments, this can include multiple predefined orientations, and in others, the mating provides for only mating to a single predefined orientation. Some exemplary embodiment of shapes for the lower tie plate 48 and the fuel support aperture 40 that can also provide for mating at a predefined orientation are provided in FIGS. 4 and 5. In some embodiments, as shown in FIG. 2, the lower tie plate 48 can include a tie plate beveled edge 56 for aiding in the installation into the fuel support aperture 40. Additionally, in some embodiments, the fuel support aperture 40 can include one or more aperture beveled or chamfered edges 58 which can assist in the seating of the lower tie plate 48 into the fuel support aperture 40. In some embodiments, the aperture chamfered edge 58 can also assist in the proper alignment of the fuel assembly 22 with the fuel support 36 for mating of the fuel assembly mating fixture 52 with the fuel support mating fixture 54. Generally, as noted above, the fuel assembly 22 cannot be fully engaged with the fuel support 36 unless they are properly aligned and oriented to allow the mating of the fuel assembly mating fixture 52 with the fuel support mating fixture 54. If not properly aligned, the fuel assembly 22 will not seat into the fuel support aperture 40 and will be elevated relative to other properly mated fuel assemblies 22 within the core 13. Such non-mating and non-alignment is visually detectable by operating personnel during reactor core refueling. Referring now to FIG. 3, one embodiment of a fuel support aperture 40 having a pentagon shape with each edge of the pentagon having a chamfered edge 58. The pentagon shape can be such that each side has the same length, or it can be an irregular pentagon wherein one of the arms of the pentagon is elongated for providing a single orientation within three hundred and sixty (360) degrees, as shown in FIG. 4B by way of example. FIGS. 4A and 4B illustrate an example of another set of shapes according to another embodiment. FIG. 4A illustrates an irregular shape of a fuel assembly mating fixture 52 that is formed from the shape of the lower tie plate 48 configured for mating. FIG. 4B illustrates a fuel support mating fixture 54 also defined by the irregular shape of the fuel support aperture 40. The fuel support 36 defines the fuel support aperture 40 that is coupled to the lumen 46 from which coolant flows into the lower tie plate 48. The fuel support aperture 40 includes in this exemplary embodiment, a chamfered edge 58 that provides for alignment between the lower tie plate 48 with the fuel support aperture 40. FIGS. 5A, 5B, 5C, and 5D provide for additionally exemplary shapes that can be applied to both the lower tie plate 48 and the fuel support aperture 40 for providing the fuel assembly mating fixture 52 and the fuel support mating fixture 54. FIG. 5A provides for mating to a predefined orientation due to two trimmed corners of a square shape. While shown by example here as a square, this could be a square or any parallelogram, triangle, or any shape have a plurality of defined edges meeting to form an angular shape. Additionally, the alignment could be with a single trimmed corner or two or more, but less than all, trimmed corners. FIG. 5B provides a different mating shape for providing the predefined orientation wherein the mating fixture is an inward slot versus an outward slot. FIG. 5C is a circle having a single mating fixture along one radius. FIG. 5D is an Isosceles triangle. Other triangular shapes such as a Scalene can also used for proper orientation. As known to those skilled in the art after reviewing this disclosure, other applicable shapes can include, but is not limited to, a quadrilateral, an octagon, a heptagon, a hexagon, nonagon, decagon, and a star. It should be understood that the fuel support mating fixture 54 can be either a male member or a female member and the fuel assembly mating fixture 52 can be either, but is generally the opposite for providing proper mating. For example, in some embodiments, a fuel support 36 supports four fuel assemblies 22, each fuel assembly 22 has a fuel assembly mating fixture 52 and the fuel support has four apertures 40 each with a fuel support mating fixture 54 for selectively receiving a different one of the fuel assemblies 22. Each fuel assembly 22 has a predefined orientation to the fuel support 36. Additionally, it is also possible for the fuel support 22 to be keyed with the circular core plate apertures 35 in which they are mounted. Referring now to FIG. 6, in operation and according to one embodiment, a method of mounting a fuel assembly in a nuclear reactor includes inserting the fuel rods in a pattern within a fuel assembly having a predetermined orientation relative to an orientation of a lower tie plate of the fuel assembly as in process 70. Process 72 then provides for installing the lower tie plate of the fuel assembly into an aperture of a fuel support wherein the fuel assembly has a predetermined orientation for receiving the lower tie plate. The installing can include selectively aligning the orientation of the fuel assembly with the orientation of the fuel support. In some embodiments, process 74 can provide for installing the fuel support within a core of the reactor including aligning the orientation of the fuel support with an orientation within the core. In other embodiments, the lower tie plate includes a lower tie plate mating fixture that provides the predetermined fuel assembly orientation and the fuel support includes a fuel support mating fixture providing the predetermined orientation for receiving the lower tie plate as provided by process 76. The installing can include selectively engaging the lower tie plate mating fixture with the fuel support mating fixture. In yet another embodiment as shown in process 78, the fuel rods are inserted in the fuel assemblies. Each fuel assembly has a different pattern of fuel rods and a predetermined orientation relative to an orientation of the respective lower tie plates. Process 80 provides for installing the fuel assemblies in a core of the reactor with each fuel assembly having a lower tie plate selectively installed in a fuel support having a predetermined orientation for selectively orienting the fuel assembly within the core. Process 82 provides for installing a plurality of fuel supports in the core with each fuel support being installed to have a predetermined orientation within the core. Process 84 provides for installing the lower tie plate of the fuel assembly into the aperture of a fuel support and thereby decreasing the elevation of the fuel rod from a preinstalled height to an installed and selectively aligned height. When describing elements or features and/or embodiments thereof, the articles “a”, “an”, “the”, and “said” are intended to mean that there are one or more of the elements or features. The terms “comprising”, “including”, and “having” are intended to be inclusive and mean that there may be additional elements or features beyond those specifically described. Those skilled in the art will recognize that various changes can be made to the exemplary embodiments and implementations described above without departing from the scope of the disclosure. Accordingly, all matter contained in the above description or shown in the accompanying drawings should be interpreted as illustrative and not in a limiting sense. It is further to be understood that the processes or steps described herein are not to be construed as necessarily requiring their performance in the particular order discussed or illustrated. It is also to be understood that additional or alternative processes or steps may be employed.
051981858	claims	1. In a nuclear reactor having fuel in the form of a core and a plenum above said core, said core having a center and a periphery, said plenum receiving coolant, apparatus comprising: a plurality of hollow sleeves for housing said fuel, said sleeves having an upper portion and a lower portion, said upper portion in said plenum, said lower portion below said plenum, said fuel positioned in said lower portion, said sleeves having holes formed in said upper portion to admit said coolant to the interior of said sleeves, said holes admitting varying amounts of coolant from sleeve to sleeve, said amount being greater in sleeves located toward said center of said core than toward said periphery of said core. boring at least two different sets of holes in said upper portion of said sleeves, said different sets capable of admitting different amounts of coolant therethrough; distributing said sleeves about said core so that said sleeves capable of admitting the highest amount of coolant are placed toward said center of said core and those capable of admitting the least amount of coolant are placed toward said periphery of said core. 2. The apparatus as recited in claim 1, wherein said holes have the same diameter and the number of holes in sleeves located toward said center of said core is greater than the number of holes located toward said periphery of said core. 3. The apparatus as recited in claim 1, wherein said holes are arranged in rows and the number of said rows is varied from sleeve to sleeve, said number of rows being greater in sleeves toward said center of said core than toward said periphery of said core so that the amount of coolant entering the interior of said sleeves toward said center of said core is greater than the amount of water entering the interior of said sleeves toward said periphery of said core. 4. The apparatus as recited in claim 1, wherein said holes are of equal diameter and arranged in rows and the number of said rows is varied from sleeve to sleeve, said number of rows being greater in sleeves toward said center of said core than toward said periphery of said core so that the amount of coolant entering the interior of said sleeves toward said center of said core is greater than the amount of coolant entering the interior of said sleeves toward said periphery of said core. 5. The apparatus as recited in claim 1, wherein said upper portion has a top and a bottom and said holes are arranged in rows beginning at said bottom of said upper portion, whereby said sleeves near said center of said core have more rows, said rows extending more toward said top from said bottom of said upper portion than sleeves located near said periphery of said core. 6. The apparatus as recited in claim 1, wherein said upper portion has a top and a bottom and said holes are of equal diameter and are arranged in rows beginning at said bottom of said upper portion, whereby said sleeves near said center of said core have more rows of holes, said rows extending more toward said top from said bottom of said upper portion than sleeves located near said periphery of said core. 7. In a nuclear reactor having a plurality of hollow sleeves forming a core, said core having a center and a periphery, said sleeves having an upper portion and a lower portion, and a plenum disposed about said upper portion of said sleeves, said plenum receiving coolant, a method for cooling said core comprising the steps of: 8. The method as recited in claim 7, wherein said holes have the same diameter and said at least two different sets each has a different number of holes. 9. The method as recited in claim 7, wherein said upper portion has a top and a bottom and said holes are bored in rows beginning at said bottom, and each set of said at least two sets has a different number of rows of holes. 10. The method as recited in claim 7, wherein said holes have the same diameter, said upper portion has a top and a bottom and said holes are bored in rows beginning at said bottom, and each set of said at least two sets has a different number of rows of holes.
	abstract	Inspection systems employing radiation filters with different attenuation characteristics to determine specimen irregularities, and related methods are disclosed. An inspection system includes a radiation emitter configured to emit a radiation beam along a radiation trajectory. Some of the radiation may be reflected by the specimen as backscatter and received by at least one radiation detector of the inspection system along the radiation trajectory. Irregularities and various materials of the specimen may produce backscatter radiation at different energies and/or scatter angles which may be identified by employing radiation filters having different attenuation characteristics. By employing these filters in communication with the radiation emitter and the radiation detector, the backscatter radiation passed through the filters may be measured and integrated at different positions of the radiation beam to produce a composite image of the specimen. In this manner, irregularities and associated materials within the specimen may be more easily identified.
052157040	summary	FIELD OF THE INVENTION This invention relates to a method and apparatus for testing the heat transfer rate of a heat exchanger in situ and, in particular, to a method and apparatus for testing the heat transfer rate of a heat exchanger located in a nuclear or other power plant. BACKGROUND OF THE INVENTION One type of heat exchanger consists of a number of tubes through which a service fluid (normally a coolant) circulates and on the outside of which a process fluid (the fluid being cooled) flows. When both the service fluid and the process fluid are water, the heat exchanger is referred to as a water-to-water heat exchanger. In another common type of heat exchanger, the process fluid flows through a number of tubes, and a gas (frequently air) is circulated around the pipes, which often have fins attached to them to improve their heat transfer capabilities. When the process fluid is water, this type of heat exchanger is referred to as an air-to-water heat exchanger. Normally the process fluid is cooled in a heat exchanger, but there is no reason in principle that a heat exchanger cannot be used to heat a process fluid. When a liquid such as water is used as the service fluid or the process fluid, the surfaces of the tubes that are in contact with the liquid may become fouled and the heat transfer efficiency of the device will therefore be impaired. (Fouling contamination is not usually a problem where only air contacts the tubes.) Fouling can take several forms: (i) particulate matter in the liquid may settle on or otherwise become attached to the surface of the tubes; (ii) substances dissolved in the fluid (e.g., calcium carbonate dissolved in water) may come out of solution and precipitate onto the heat transfer surfaces; (iii) the fluid may react with the heat transfer surface, forming a layer (e.g., corrosion on carbon steel) which acts as a barrier to the flow of heat; (iv) macroorganisms (e.g., Asiatic clams) or microorganisms (e.g., bacteria) may become attached to the tubes and thereby impede the heat flow between the process fluid and the service fluid. Microorganic fouling is a particular problem where the ultimate heat sink is an open body of water (an ocean, river or pond), and it is often more difficult to predict than the other kinds of barriers described above. Moreover, a layer of microorganisms may send out a layer of hairs or other projections to feed on nutrients in the water. These projections can impede the flow of the service fluid, producing a layer of relatively still water which acts as a further barrier to heat flow. A nuclear power plant contains a number of heat exchangers which are designed to remove heat that may be generated during an emergency. Unless the plant actually experiences an emergency, these heat exchangers remain unused, and whether their heat removal capabilities have become impaired as a result of fouling is unknown. Recognizing the risks of this situation, the U.S. Nuclear Regulatory Commission on Jul. 18, 1989 issued Generic Letter 89-13, which requires that operators of nuclear power plants adopt a program to verify the heat transfer capability of all safety-related heat exchangers cooled by service water. Because of the large volume of service and process water which flows through the heat exchangers used in nuclear power plants, testing the efficiency of such a heat exchanger presents problems. Whichever fluid (i.e., service or process) is to be heated (or chilled) to conduct the test, a temperature differential of several degrees (for example, 2 or 3 degrees F.) at most can be obtained. This is far less than the temperature differential that would occur during an actual emergency, and thus the behavior of the heat exchanger during an emergency can be predicted only by extrapolating the results of the test to a much larger temperature differential. Therefore, extremely accurate (and hence expensive) instruments must be used to avoid any errors of measurement that would be unduly magnified in the extrapolation process. The efficiency of a heat exchanger can also be gauged relatively inexpensively by measuring the pressure drop between the inlet and outlet of the service water. The pressure drop is related to flow restriction which in turn reflects the amount of fouling, and for a particular exchanger and type of fouling this information can be used to estimate the heat transferability of the exchanger. However, this test is not very useful unless the operator develops a correlation between the pressure drop and the heat transfer rate of the particular exchanger involved. This in turn requires an accurate means of directly determining the heat transfer rate of the exchanger. The difficulty of measuring the heat transfer performance of their heat exchangers has led some operators to clean them periodically, whether or not their performance is known to be impaired. While this is one solution to the problem, unnecessary cleaning may shorten the service life of a heat exchanger. Ideally, a heat exchanger should be cleaned only as often as is necessary to assure that its heat transfer capabilities are satisfactory. SUMMARY OF THE INVENTION Using the method of this invention, the performance of a heat exchanger is determined by measuring the heat transfer capabilities of an individual tube. A relatively small reservoir of service fluid is connected to the inlet and outlet ports of a tube. The reservoir is provided with a heater or chiller and the service fluid is circulated through the tube. When a steady state is reached, the heat transfer characteristics of the tube (including the fouling resistance) are measured using known mathematical relationships. By testing an individual tube, a relatively large differential between the temperature of the service fluid at the inlet and outlet of the tube (for example, 10 to 15 degrees F.) can be obtained. The results of this test can be extrapolated to the temperature differentials that would be encountered in an emergency more easily and without the need for unduly expensive instruments. Apparatus for conducting such a test is also described.
061608653	abstract	A synchrotron radiation measuring system includes an X-ray detector movable in a direction of intensity distribution of synchrotron radiation to follow shift of the synchrotron radiation, and a computing device for reserving therein one of (i) a relation between a signal of the X-ray detector and the intensity of synchrotron radiation and (ii) a relation among a signal of the X-ray detector, the level of vacuum at a synchrotron ring and the intensity of synchrotron radiation, wherein the intensity of synchrotron radiation is measured through the computing device on the basis of an output signal of the X-ray detector.
	abstract	A storage phosphor panel can include an extruded inorganic storage phosphor layer including a thermoplastic polyolefin and an inorganic storage phosphor material, where the extruded inorganic storage phosphor panel has a DQE comparable to that of a traditional extruded inorganic solvent coated inorganic storage phosphor screen. Also disclosed is an inorganic storage phosphor detection system including an extruded inorganic storage phosphor panel that can include an extruded inorganic storage phosphor layer including a thermoplastic olefin and an inorganic storage phosphor material; and photodetector(s) coupled to the extruded inorganic storage phosphor panel to detect photons generated from the extruded inorganic storage phosphor panel. Further disclosed is a method of making an extruded inorganic storage phosphor panel that can include providing thermoplastic particles including at least one thermoplastic polyolefin and an inorganic storage phosphor material; and melt extruding the thermoplastic particles to form an extruded inorganic storage phosphor layer.
046577297	description	DESCRIPTION OF THE PREFERRED EMBODIMENTS Solid materials can be inserted into reactor elements before they are sealed that generate detectable gases on exposure to the neutron flux within a nuclear reactor to signal the occurrence of a leak or failure in the particular element. By using different solid tag materials and mixtures of different materials, particular reactor elements, such as specific tube assemblies, can be identified as being the source of a leak or failure. Elements such as the halogens, Group VIIA of the periodic table, generate radioisotopes when exposed to a neutron flux. These isotopes undergo beta decay to become noble gases. Examples of preferred embodiments of the process using halogens as solid tags are: EQU .sup.79 Br(n,g).sup.80 Br .beta..fwdarw..sup.80 Kr A. EQU .sup.81 Br(n,g).sup.82 Br .beta..fwdarw..sup.82 Kr B. EQU .sup.127 I(n,g).sup.128 I .beta..fwdarw..sup.128 Xe C. EQU .sup.37 Cl(n,g).sup.38 Cl .beta..fwdarw..sup.38 Ar D. EQU .sup.19 F(n.g).sup.20 F .beta..fwdarw..sup.20 Ne E. Although any solid material which produces a detectable gas may be used, the preferred embodiments of the invention are limited to those which result in detectable gases that are not produced to any significant degree during reactor operation. That is, materials which result in fission products cannot easily be distinguished from the background material found in the cover gas after a fuel element ruptures. For example, in a sodium cooled reactor, the use of fluorine may not be useful because sodium is also converted to neon. Similarly, .sup.127 I and .sup.81 Br may not be useful in every case as they may also be fission products. It is possible, however, in another embodiment of the invention, that a tag could be used which is similar to naturally occurring background constituents found in the cover gas as fission products if the background levels are recorded through constant or periodic monitoring. In such an embodiment, a significant departure from a recently measured or anticipated background level would indicate a component failure or leak. A procedure dependent on departures from known background levels used to indicate problems would be particularly useful in situations where there has been some build-up of gaseous isotopes from earlier tagged reactor component failures or other accountable sources. Gaseous tags are generated from solid materials according to the equation: EQU Atoms of tag gas=.phi..sigma.N where .phi.= ##EQU1## .sigma.=cross section (cm.sup.2 /atom N=atoms of target material For example, in the reaction of Br to Kr, neglecting bromine depletion, one gram of Br will yield somewhat more than 1/3 cc (STP) of Kr tag gas in 10 days in a typical liquid metal fast breeder reactor environment. EQU N=approximately 0.008.times.10.sup.24 atoms of Br/g EQU .phi.=approximately 10.sup.15 neutrons/cm.sup.2 .times.sec (typical fast breeder reactor) EQU .sigma.=approximately 10.sup.-24 cm.sup.2 /atom (typical fast breeder reactor) Approximately 10.sup.18 atoms of tag gas are generated each day. At STP one cc of Kr=2.76.times.10.sup.19 atoms. Therefore, approximately 0.37 cc of Kr at STP are generated from 1 g. Br in 10 days in this example. Although the above example is based on characteristics of fast breeder reactors, the generation rate of isotopic tag gases in a typical thermal reactor would be on the same order of magnitude. Although the neutron flux may be 10 to 50 fold smaller in a thermal reactor, its cross section would be larger by the same factor. Hence, the present invention may be used equally well in a thermal reactor. The production of detectable gaseous isotopes is linear with time, which is evident from the above equation. Starting with, for example, one gram Br, approximately 0.037 cc (STP) Br will be generated on the first day, and on every day thereafter. A leak in a reactor component will be detected within a fraction of the first day based on the use of one gram amounts of a tag material. Early detection can be assured, of course, by using slightly greater quantities of the tag material. The preferred method for testing the cover gas for tag gas is with a mass spectrometer. Conventionally, the cover gas is monitored by continuously passing a sample through a device for detecting fission products, i.e., radioactivity. When fission products are detected, a sample is assayed with a mass spectrometer to identify the isotope and determine the isotopic ratio. A mass spectrometer identifies identically charged ions having different masses by measuring differences in their deflection in electrical and magnetic fields. The quantity of gas generated from one gram of solid tag material in 10 days is easily detected. Detection of lesser concentrations can be enhanced using cryogenic techniques and concentrating the tag gas in an activated charcoal filter. Using a mass spectrometer with these methods, very small amounts of tag gases, present in amounts as low as 10.sup.-11, can be detected, identified, and accurately measured; thus enabling the early detection of a leak or component failure and identifying the source of the problem. In a typical reactor, a one gram amount of solid tag materials should generate sufficient isotopic tag gases to be detectable in the cover gas within a few hours. Requirements of a tag material, in addition to being detectable and measurable in contrast to the background, include the characteristics of not affecting the neutron flux adversely, not being further changed from the identifiable product on continued exposure to radiation, and not being significantly soluble in the reactor coolant. The solid tag materials of the present invention may be used in conjunction with the gas tags used in the prior art, if necessary, to obtain a greater variety of isotopic ratios. Moreover, failures which occur at the time a fuel assembly or other tagged reactor component is placed into the reactor would be signalled by the immediate detection of the gaseous tag. The solid tag alone requires a finite time to generate sufficient gaseous isotopes to be detected. The solid tag materials may be introduced as a salt of the desired element. For example, the sodium, potassium, calcium, iron or nickel salts of a halogen isotope tags. The salts may be introduced in the form of pellets or as powder or granules in containers which are fairly non-reactive in the fuel rod environment. The tag salts selected must consist of predetermined ratios of particular isotopes. For example, salts consisting of known ratios of .sup.79 Br.sup.- and .sup.81 Br.sup.- may be used to generate the tag gases comprising known ratios of .sup.80 Kr and .sup.82 Kr. Even though the element is in the form of an ion in the salt, it will be transmuted in the neutron flux to the noble gas. Similarly, salts having particular proportions of .sup.35 Cl and .sup.37 Cl will generate .sup.36 Ar and .sup.38 Ar in unique ratios quite different from the ratio of these isotopes found in natural argon. Mixtures of different elements may also be employed to establish unique tags. The same isotopic ratios of chloride could be used with or without iodine salts to double the number of different tags available. Any cation component may be used in the solid salt used for tagging as it is irrelevant to generating the isotopic gas. The commonly available sodium, potassium, or calcium salts may be used. Also, ion salts and similar salts of metals which are fairly unreactive in their elemental form may be preferred in some applications. The invention provides a method for tagging reactor components with solid salts that are inexpensive to use and are easily handled. It is believed that the cost of tagging reactor components will be reduced one hundred fold using these salts. Moreover, as the isotopic salts are stable, the solid tags can be produced on a large scale and inventoried at a central facility, then shipped to reactor sites as needed. In addition, as physically small amounts of solid salts quickly generate measurable quantities of isotopic gas, units of different salts can be combined at the reactor site as required to obtain a variety of different gaseous isotope ratios. The above embodiments are presented as examples to illustrate the invention without intending to limit the invention thereby. It will be understood that the present description is susceptible to various modifications, changes, and adaptations within the invention defined by the following claims.
060977795	abstract	Fuel bundle and control rod assemblies for a nuclear reactor are described. In one embodiment of the apparatus, the nuclear reactor includes several conventional size fuel bundles and at least one large control rod. The large control rod includes four fuel bundle receiving channels, and each such channel is sized to receive four conventional size fuel bundles.
	description	This is a division of application Ser. No. 09/446,144, filed Mar. 2, 2000 now U.S. Pat. No. 7,796,720. The present invention proposes a method of element transmutation by efficient neutron capture Ei(A, Z)+n→ES(A+1,Z) of an initial “father” isotope, embedded in a diffusing medium which is highly transparent to neutrons and which has the appropriate physical properties as to enhance the occurrence of the capture process. The produced “daughter” nucleus, depending on the application, can either be used directly, or in turn, allowed for instance to beta-decay, E S ⁡ ( A + 1 , Z ) ⁢ → β - - ⁢ decay ⁢ E f * ⁡ ( A + 1 , Z + 1 ) ,or more generally, to undergo an adequate: spontaneous nuclear transformation into another radio-active isotope. Accordingly, the basis of the present transmutation scheme is a method of exposing a material to a neutron flux, wherein said material is distributed in a neutron-diffusing medium surrounding a neutron source, the diffusing medium being substantially transparent to neutrons and so arranged that neutron scattering within the diffusing medium substantially enhances the neutron flux, originating from the source, to which the material is exposed. The device employed to achieve the efficient neutron capture according to the invention is referred to herein as a “Transmuter”. The term “transmutation” is understood herein to generally designate the transformation of a nuclear species into another nuclear species, having the same or a different atomic number Z. The Transmuter is driven by an internal neutron source, which, depending on the application, can be of a large range of intensities and appropriate energy spectrum. It may be, for instance, a beam from a particle accelerator striking an appropriate neutron generating and/or multiplying target or, if a more modest level of activation is required, even a neutron-emitting radioactive source. The source is surrounded by a diffusing medium in which neutrons propagate, with a geometry and composition specifically designed to enhance the capture process. The material to be exposed to the neutron flux is located in a dispersed form inside the diffusing medium. The Transmuter presently describe&relies on a vastly increased neutron capture efficiency. Neutron capture efficiency is defined as the capture probability in the sample for one initial neutron and unit mass of father element. It is designated by the symbol η, typically in units of g−1. In the case of a gas, the mass is replaced with the unit volume at normal pressure and temperature conditions (n.p.t., i.e. atmospheric pressure and 21° C.), and the capture efficiency is indicated with ηv for which we use typical units of litre−1. According to the invention, the increased neutron capture efficiency is achieved with the help of the nature and of the geometry of the medium surrounding the source, in which a small amount of the element to be transmitted is introduced in a diffused way: (1) The medium is highly transparent, but highly diffusive. Transparency is meant as the property of a medium in which neutrons undergo mostly elastic scattering. The succession of many, closely occurring elastic scattering events (generally about isotropic) gives a random walk nature to the neutron propagation. The flux is enhanced because of long resulting, tortuous, random, paths that neutrons follow before either being captured or exiting the large volume of the transparent medium. Using an optical analogy, the target-moderator sphere is chosen to be diffusive, but highly transparent to neutrons. Doping it with a small amount of additional material makes it “cloudy”. As a consequence, most of the neutrons are captured by the absorbing impurities. (2) In addition, the large peak values of the capture cross-section of the sample which correspond to the nuclear resonances may be exploited using a diffusing medium having the above feature (1), but of large atomic mass A. In such medium, he neutron energy is slightly reduced at each (elastic) scattering, thus “scanning” in very tiny energy steps through the resonance spectrum of the sample during the smooth, otherwise unperturbed, energy slow-down of the initially high energy (MeV) neutrons of the source. The choice of the diffusing medium depends on the most appropriate energy at which neutron captures must occur. If neutrons are to be thermalised, i.e. captures have to occur at thermal energies (≈0.025 eV, only the previously mentioned feature (1) is used and a low A (atomic mass number) medium but very transparent to neutrons is to be used, like for instance reactor purity grade graphite or D2O (deuterated water). If, instead, neutron capture has to be performed with father elements having large values of capture cross-section in correspondence with resonances, both features (1) and (2) are used and the best elements for the diffusing medium are Lead and Bismuth (or a mixture thereof), which have simultaneously an anomalously small neutron capture cross-section and a very small “lethargy”, ξ=9.54×10−3. According to the Shell Nuclear model, built in analogy to atomic electrons, “magic” numbers occur in correspondence of “closed” neutron or proton shells. Atomic number Z=82 is magic, so is the number of neutrons in correspondence of 208Pb. Magic number elements in the nuclear sense have a behaviour similar to the one of Noble Elements in the atomic scale. Therefore, the neutron transparency is the consequence of a specific nuclear property, similar to the one for electrons in noble gases. Lethargy (ξ) is defined as the fractional average energy loss at each neutron elastic collision. While 209Bi is a single isotope, natural Lead is made of 204Pb (1.4%), 206Pb (24.1%), 207Pb (22.1%) and 208Pb (52.4%), which have 10 quite different cross-sections. Isotopic enrichment of isotope 208Pb could be beneficial. However, the use of natural Pb will be more specifically considered herein, for its excellent neutron properties, low activation and its low cost. The domain of applications of the present method of enhancement of neutron captures is very vast. A first applicative aspect of the invention relates to a method of producing a useful isotope, which comprises transforming a first isotope by exposing a material containing said first isotope to a neutron flux as set forth hereabove, and the further step of recovering said useful isotope from the exposed material. A second applicative aspect of the invention relates to a method of transmuting at least one long-lived isotope of a radioactive waste, by exposing a material containing said long-lived isotope to a neutron flux as set forth hereabove, wherein at least the portion of the diffusing medium where the exposed material is distributed is made of heavy elements, so that multiple elastic neutron collisions result in a slowly decreasing energy of the neutrons originating from the source. (1) Activation of (Short-Lived) Isotopes for Industrial and Medical Applications. In this case, the Transmuter will be denominated as the Activator. Radio-nuclides are extensively used for medical diagnosis applications and more generally in Industry and Research. As well known, these nuclides are used as “tracing” elements, i.e. they are directly detectable within the patient or material under study because of their spontaneous radioactive decays. In order to minimise the integrated radio-toxicity, the half-life of the chosen tracing isotope should be short, ideally not much longer than the examination time. As a consequence, its utilisation is limited to a period of a few half-lives from activation, since the radioactivity of the isotope is decaying exponentially from the moment of production. Another application of growing interest for Radio-nuclides is the one of (cancer) Therapy, for which doses significantly larger than in the case of diagnosis are required. Most of these isotopes must have a relatively short half-life, since they are generally injected or implanted in the body of the patient. The main supplies for these isotopes are today from Nuclear Reactors and from particle accelerators in which a suitable target is irradiated with a charged particle beam. The simplicity of the device proposed and its relatively modest cost and dimensions are intended to promote “local” production of short-lived radio-isotopes, thus eliminating costly, swift transportation and the consequent need of larger initial inventories and thus extending their practical utilisation. This is made possible by the high neutron capture efficiency as the result of the present method, which permits to produce the required amount of the radio-isotope with a relatively modest neutron generator. The present method of neutron activation is intended to be a competitive alternative to Reactor-driven, neutron capture activation. In addition, several isotopes which are difficult to produce by activation with the (usually thermal) neutrons of an ordinary Reactor, can be produced using the broad energy spectrum of the neutrons in the Activator, extending to high energies and especially designed to make use of the large values of the cross-section in correspondence of resonances. This is the case for instance in the production of 99mTc (99Mo), widely used in medicine and which is nowadays generally chemically extracted from the Fission Fragments of spent Nuclear Fuel. According to the present method, this popular radio-isotope can be obtained, instead, by direct neutron resonant activation of a Molybdenum target, with the help of a much simpler and less costly Activator driven by small particle Accelerator. Incidentally, the total amount of additional, useless radioactive substances which have to be produced and handled in association with a given amount of this wanted radio-nuclide is also greatly reduced.(2) Transmutation into Table Species of Offending, Long-Lived Radio-Isotopes, as an Alternatives to Geologic Storage. In this case, the Transmuter will be denominated as the Waste Transmuter. Since the totality of the sample should be ideally transmuted, a much stronger neutron source is required. Even for the strongest sources, the highest efficiency of neutron capture is crucial to the complete elimination. The present method of enhanced captures makes practical this technique of elimination. Ordinary Nuclear Reactors. (Light Water Reactors, LWR) produce a considerable amount of radioactive waste. The radiotoxicity of such waste persists over very long periods of time, and it represents a major drawback of the Nuclear Technology. Fortunately, only a very small fraction of the waste resulting from a Reactor is responsible for the bulk of the long lasting radiotoxicity, and it is easily separable chemically. In order of importance, the by far largest contribution comes from the Actinides other than Uranium (Trans-uranic elements, or TRU's), which represent about 1% of the waste by weight. These elements are fissionable under fast neutrons. Therefore, they may be eliminated with considerable extra recovered energy, for instance with the help of an Energy Amplifier (EA) as disclosed in International Patent Publication WO 95/12203(See C. Rubbia, “A High Gain Energy Amplifier Operated with Fast Neutrons”, AIP Conference Proceedings 346, International Conference on Accelerator-Driven Transmutation Technologies and Applications, Las Vegas, July 1994). Next in importance for elimination are the Fission Fragments (FF), which are about 4% of the waste mass, and which divide into (1) stable elements (2) short-lived radio-nuclides and (3) long-lived radio-nuclides. The separation between short- and long-lived elements is naturally suggested by the 30 years half-life of 90Sr and 137Cs, which are dominating the FF activity at medium times (<500 years) after an initial cool-down of the fuel of a few years. Finally, there are some activated materials, like the cladding of the fuel, which represent a much smaller problem, and which can be disposed without problems. Whilst the elimination of the TRU's is performed best by burning them in a fast neutron-driven EA, the present method of element transmutation can be used to transform the long-lived FF's into harmless, stable nuclear species (it is assumed that elements with half-life of less than 30 years may be left to decay naturally). The simultaneous elimination of the TRU's and of the long-lived FF's suggests the use of the core of the EA (in which TRU's are burnt) as the neutron source for the Transmuter, dedicated to the long-lived FF's. In this case, the Transmuter will surround the EA, using neutrons escaping from it. The combination of the EA operated with TRU's and of the Transmuter as long-lived FF's Waste Transmuter is both environmentally very beneficial and economically advantageous, since (1) considerable additional energy is produced by the EA (>40% of the LWR) and (2) the simultaneous elimination of the FF's can be performed “parasitically”, with the help of the extra neutrons available. However, as already pointed out, in order to eliminate completely the unwanted FF's with these extra neutrons, a very high neutron capture efficiency is required, as made possible with the present method. The method is first elucidated in some of the applications as Activator for medical and industrial applications. The procedures to be followed in order to prepare the radioactive sample are better illustrated by the following practical examples: (1) A first procedure, suitable for medical examinations (e.g., for thyroid), consists of activating directly inside the Transmuter an already prepared, pharmacological Iodine compound. The element is initially available in the most appropriate chemical compound, such as Sodium, Iodide (NaI), made with natural Iodine (stable isotope 127I). Shortly before administration, the compound is introduced in the Activator driven by a small proton accelerator (23 MeV and 1 mA) and activated—for instance during a time of the order of one 128I half-life (25 minutes=25 m) or correspondingly less for smaller activation strengths—with the help of the reaction 127I+n→128I+γ, which transforms natural Iodine into the tracing element 128I which undergoes β−-decay with a prominent γ-line at 443 keV. There is no chemical “preparation” between activation and examination. This very simple procedure is becoming practical with the present method because of the higher neutron capture efficiency, which produces the required source strength (≦1 GBq, with 1 GBq=109 disintegrations/s=27.0 mCie (milli-Curie). 1 Cie=1 Curie=3.7×1010 dis/s), starting from a tiny, initial amount of natural Iodine (≦1 gram), and using a conventional accelerator of the scale already in wide use within hospitals for other applications such as Positron-Electron-Tomography (PET). The present method makes practical the use of 128I as a tracing element for thyroid diagnosis, with a much shorter half-life (25 m) than the one of currently used Iodine isotopes (131I and 123I), and the corresponding important advantage of a much smaller, dose to patients. The current methods of Iodine examinations are based on 131I, which has a relatively long half-life of 8 days, and which causes large, intake doses for the patients (roughly in the ratio of half-lives (461/1), and 123I which has the shorter half-life of 13.2 hours (31.8 times the one of 128I), but which is of difficult, costly production since it is normally produced by 30 MeV protons and (p,2n) reaction on isotopically-separated 124Te (natural abundance 4.79%). In order to use natural Xe, the reaction is (p, 5n) and the energy must be at least 60 MeV. The presently proposed method has therefore both a very simple applicability and leads to much smaller doses to the patient for a given disintegration rate during the examination. It is noted that the larger doses of the current methods generally hamper extensive applicability in the cases of young subjects and of pregnant women. (2) A second example illustrates the case in which some (simple) chemical transformation is needed between (i) the activation and (ii) the use of the radioactive compound. We visualise this procedure in the case of a 99mTc medical examination, of which many millions are done annually world-wide (see for instance Table 9) In this case, the small sample to be irradiated consists of Molybdenum, for instance in the form of MoO3. The isotopic content of 98Mo in natural Molybdenum is 24%. Isotopic enrichment will be convenient, though not mandatory. The appropriate sample of 99Mo (τ1/2=65 hours=65 h) is produced with the help of an Accelerator-driven Activator and the capture reaction 99Mo+n→99Mo+γ. The activated Molybdenum sample is then handled according to a generally used procedure: transformed, for instance, in the form of an appropriate salt, it is captured in an Alumina absorber. The production of 99mTc proceeds inside the absorber through the subsequent decay reaction The 99mTc (which has a relatively short τ1/2=6.01 h) is extracted in the form of the ion Tc4+, for instance by passing through the Mo sample in the Alumina (which remains insoluble) a solution of water with a small amount of NaCl. Since only a very small fraction of the compound is activated at each exposure, the Molybdenum “father” can be recycled, which is of economical importance if the Molybdenum is isotopically enriched, by flushing it from the Alumina absorber and repetitively re-introducing it in the Activator. (3) Many radio-isotopes used in medicine and in industry are extracted from fragments of Uranium fission. The group of these elements is referred to herein under the generic name of “Fissium”. The increased capture efficiency offered by the method works as well in the case of neutron captures leading to fission. Fissium can be produced in the Activator introducing a small target of Uranium, possibly enriched, which, as in the previous examples, is strongly activated by primarily resonance-driven captures. The system is not critical and a small amount of fissile target material is sufficient to obtain relatively large amounts of Fissium. In the case of activation of short-lived elements, the target must be frequently extracted and reprocessed. This is made extremely easy in the geometry and otherwise general conditions of the operation of an Activator, when compared with a nuclear Reactor. The amount of Plutonium produced by the captures in 238U is negligibly small and it represents no proliferation concern. (4) The present method may further be employed in order to dope pure Silicon crystals with Phosphorus, for use in the semiconductor industry. Neutron-driven doping is a very uniform doping which can be performed in the bulk of a large crystal. Natural Silicon is made of three isotopes 28Si (92.23%) 29.Si(4.67%). and 30Si (3.1%). Neutron captures transform the isotopes into the A+1 Silicon elements. 31Si is unstable, (τ1/2=157 m),. and it decays into the stable 31P, which is the only stable isotope of Phosphorus. This method offers a simple way of doping the inside of relatively large crystals. A reasonable exposure can lead to an implantation of several parts per billion (p.p.b.=10−9) of Phosphorus atoms inside a very pure crystal. The exact amount of the implantation can be precisely controlled by the parameters of the exposure. These cases are examples of the potentialities of the Transmuter operated in the Activator mode. Obviously, a variety of scenarios are possible, depending on the type of radio-isotope and of the specific application. More generally, and as described in more detail later on, one can achieve capture efficiencies η which are of the order of η=1.74×10−6 g−1 of all produced neutrons for Mo activation (99mTc production), and of the order of η=2.61×10−5 g−1 for activating 128I in a pharmaceutical Iodine sample. If neutrons are produced by the source at constant rate S0=dn/dt for the period T, the number of activated daughter nuclei Nd(T) of decay constant τ (the decay constant τ is defined as the time for 1/e reduction of the sample. It is related to the half-life τ1/2 of the element by the relation τ=τ1/2/ln(2)=1.4436×τ1/2) and from a mass m0 of the father element, builds up as: N d ⁡ ( T ) = m 0 ⁢ η ⁢ ⅆ n ⅆ t ⁢ τ ⁡ ( 1 - ⅇ - T / t ) ; ⅆ β ⅆ t ⁢ ( T ) = N d ⁡ ( T ) τ = m 0 ⁢ η ⁢ ⅆ n ⅆ t ⁢ ( 1 - ⅇ - T / τ ) [ 1 ] We have indicated with dβ/dt the corresponding decay rate. An equilibrium sets between production and decay of the daughter element for T>>τ, in which decay dβ/dt and neutron capture rates m0 η dn/dt become equal. To produce, for instance, 0.1 GBq (dβ/dt=108 sec−1) of activation in each gram of sample material (m0=1 gram) at equilibrium, the neutron production rates required are then 108/(1.738×10−6)=5.75×1013 n/sec and 108/(2.61×10−5)=3.8×1012 n/sec in the above examples for 99mTc and 128I, respectively. In the case of element activation through Fissium, let us indicate with ηf the efficiency for Fissium production (fission), and with λ the atomic fraction of the element in the Fissium. After an exposure time texp, and a reprocessing time trep of a fissionable mass m0, the activity of the extracted compound is given by: ⅆ β ⅆ t = ⅇ - t rep / τ ⁡ ( 1 - ⅇ - t exp / τ ) ⁢ ⅆ n ⅆ t ⁢ m 0 ⁢ λ ⁢ ⁢ η f [ 2 ] The method is elucidated in the case of the transmutation of the long-lived FF's of the waste (spent fuel) from a typical Light Water Nuclear Reactor (LWR) Chemical reprocessing of the spent Fuel can separate: (1) the unburned Uranium (874.49 ton), which can be recycled, provided of sufficient purity; (2) the TRU's (10.178 ton) which are destined to be incinerated in a Fast Breeder or in an Energy Amplifier (EA) . The actual breakdown of the TRU's, taken after a 15-year cool-down, is as follows: Np, 545.6 kg , Pu, 8761.8 kg.; Am, 852.37 kg ; and Cm, 18.92 kg. (3) the FF (38.051 ton), which will be further considered, in view of selective transmutation. Figures within parenthesis refer to standard LWR (≈1 GWattelectric) and 40 years of calendar operation. Burn-up conditions and initial Fuel composition refer to the specific case of Spain after 15 years of preliminary cool-down (we express our thanks to the company ENRESA for kindly supplying all relevant information in this respect) FF's are neutron-rich isotopes, since they are the product of fission. It is a fortunate circumstance that all truly long-lived element in the waste are such that adding another neutron is, in general, sufficient to transform them into unstable elements of much shorter life, ending up quickly into stable elements. If elimination is simultaneously performed both for the TRU's and the selected FF's, the surplus of neutrons produced by fission can be exploited to transmute the latter as well, of course provided that the transmutation method makes an efficient use of the surplus neutron flux. The simultaneous combination of TRU incineration and of selective FF transmutation is environmentally highly beneficial, since then only those products which are either stable or with acceptable half-life (<30 years) will remain. Contrary to chemical waste, which is generally permanent, natural decay of these elements makes them “degradable”. It is noted, for instance, that the elimination time of fluoro-carbons and of CO2 in the atmosphere of the order of several centuries. In the case of an EA, the proposed method is directly applicable on the site of the Reactor, provided that a suitable (pyro-electric) reprocessing technique is used. Therefore, the combination closes the Nuclear Cycle, producing at the end of a reasonable period only Low Level Waste (LLW) which can be stored on a surface, presumably on the site of the Reactor. The list of the major long-lived FF's from the discharge of nuclear fuel is given in the first column of Table 1, for a standard LWR (≈1 GWattelectric) and 40 years of calendar operation. The initial mass mi of each isotope and of the other isotopes of the same element are listed, as well as their half-lives τ1/2, expressed in years. Further separation of individual elements obviously requires isotopic separation technologies, which are not considered for the moment. Under irradiation, as will be shown later on, the rate of transmutation is, in a first approximation, proportional to the resonance integral, defined as Ires=∫σn,γ(E)dE/E and measured in barns (1 barn=1 b=10−24 cm2), σn,γ(E) being the cross-section of the (n,γ)-capture process for a neutron of energy E. As shown in Table 1, the daughter element (column “next”) is normally either stable, hence harmless, or short-lived, quickly decaying into a stable species (column “last”) The total activity ζ, in Cie, accumulated after the 40 years of operation is also shown. Since the lifetime of these elements is very long, unless they are transmuted, they must be safely stored without human surveillance. TABLE 1Stockpile of the most offending, long-lived FF's producedby a “standard” LWR after 40 years of operation.Othermiτ1/2IresζνminIsot.isot.(kg)(y)(b)nextτ1/2last(Cie)(m3)99Tc8432.11E5 310.100Tc15.0s100Ru14455.48181.All:843129I196.021.57E7 26.5130I12.36h130Xe34.74327.127I59.4stable149.128I25.0m128XeAll:255.4293Zr810.41.53E6 15.294Zrstable94Zr2040.1583.90Zr257.8stable0.1791Zrstable91Zr91Zr670.4stable6.892Zrstable92Zr92Zr724.6stable0.6893Zr1.53E6y94Zr94Zr838.4stable15.495Zr64.02d95Mo96Zr896.8stable5.897Zr16.9h97MoAll:4198.4135Cs442.22.30E6 60.2136Cs13.16d136Ba510.1510.133Cs1228.4stable393.134Cs2.06y134Ba137Cs832.230.1 0.616138Cs32.2m138BaAll:2502.8126Sn29.481.0E5 0.139127Sn2.10h127I838.1239.116Sn7.79stable12.4117Snstable117Sn117Sn8.67stable17.8118Snstable118Sn118Sn8.812stable5.32119Snstable119Sn119Sn8.77stable5.14120Snstable120Sn120Sn8.94stable1.21121Snstable121Sn122Sn9.84stable0.916123Sn129.2d123Sb124Sn13.40stable7.84125Sn9.64d125TeAll:95.7079Se6.576.5E4 56.80Sestable80Se458.6131.77Se1.15stable28.178Sestable78Se78Se2.73stable4.779Se6.5E4y80Se80Se15.02stable0.92881Se18.4m81Br82Se37.86stable0.79583Se22.3m83KrAll:63.33 As a measure of the magnitude of the storage problem, we have indicated the minimum diluting volume Vmin, in m3, required by the US Regulations (U.S. Nuclear Regulatory Commission, “Licensing Requirements for Land Disposal of Radioactive Wastes”, Code of Federal Regulations, 10 CFR Part 61.55, May 19, 1989) for Low Level Waste and surface or shallow depth permanent storage, Class A (which means without active surveillance and intrusion protection). We review each element of Table 1 in order of decreasing storage volume: (1) Technetium (99Tc, 843 kg, 535×103 GBq/reactor) is the most offending FF element, as evidenced by the large value of the storage volume, 48181 m3/reactor. Technetium is also soluble in water as Tc4+, and during its long half-life (2.11×105 years) it will presumably drift out of the Repository into the environment, and hence into the biological cycle (see “Nuclear Wastes, Technologies for Separation and Transmutation”, National Academy Press 1996). It is known that plants (algae; dataon Fucus Vescicolosous indicate a ratio of accumulation with respect to the surrounding water between 21000 and 89000—see F. Patti et al. ,“Activités du Technétium 99 mesurées dans les eaux résiduaires, l'eau de mer (Littoral de la Manche, 1983), in “Technetium and the Environment” edited by G. Desmet et al, Elsevier Publishers, 1984, p. 37—and of the order of 14000 and 50000. in the points more distant; in the Greenland waters, this ratio is from 250 to 2500), fresh water and marine organisms (in the Greenland water, the ratio with respect to surrounding waters is from 1000 to 1400 for lobsters, and from. 100 to 2.00 for red abalone) accumulate the element out of the surrounding waters, so that it may end up in the humans through food. Organic matter becomes a Geochemical sink for 99Tc in soils and sediments. The physiological effects of Technetium have been poorly studied (see K. E. Sheer et al, Nucl. Medicine, Vol. 3(214), 1962, and references therein). When Technetium is injected, it reaches almost all tissues of the organism, and it is retained by the stomach, blood, saliva and in particular by the thyroid gland (12 to 24%) (see K. V. Kotegov, Thesis, Leningrad LTI, 1965). Concentration of Technetium with a long life in the organism is very dangerous, since it may lead to lesions of the tissues by β-radiation. Its release in the Oceans is an irreversible process on the human time scale, and its long-term effects are largely unknown. The diffusion of 99Tc in the sea water is evidenced by the discharges arising from the reprocessing plants of nuclear fuel, which amount to date to about 106 GBq (the quantity due to nuclear weapon testing is about 10 to 15% of this value). Substantial amounts of animal and vegetal contamination, which are particularly strong in the immediate vicinity of the, reprocessing plants of Sellafield and La Hague (see E. Holm et al. “Technetium-99 in Algae from Temperate and Arctic Waters in the North Atlantic”, in “Technetium and the Environment” edited by G. Desmet et al, Elsevier Publishers, 1984, p.52), have been discovered all the way to Greenland (see A. AArkrog et al. “Time trend of 99Tc in Seaweed from Greenland Waters”, in “Technetium and the Environment” edited by G. Desmet et al, Elsevier Publishers, 1984, p.52) (the transfer time from Sellafield to Greenland has been measured to be 7 years). Fortunately, Technetium is a pure isotope with a large resonant cross-section, leading to the stable 100Ru. Therefore, its elimination is the easiest, and for the above-mentioned reasons, it should be transmuted with the highest priority. (2) Iodine activation is small (129I, 196.2 kg, 1.28×103 GBq/reactor), only 2.40×10−3 the one of Technetium, but Iodine is also soluble in water and, presumably (see “Nuclear Wastes, Technologies for Separation and Transmutation”, National Academy Press 1996), will drift out of the Repository into the biological cycle. This is why, in spite of the small activity, Iodine requires, according to US regulations, a large diluting volume, i.e. 4327 m3/reactor. Studies on 131I, which of course are also applicable for 129I, show for instance that the transfer to goat's milk from blood is for Iodine 100 times larger than for Technetium. The transfer from contaminated pasture to milk is 5600 times larger than for Technetium. Therefore, it is of importance that also Iodine be transmuted. Iodine is produced by the LWR as a two-isotope mixture, with 76.7% of 129I, the rest being stable 127I. The stable Iodine isotope transforms under neutron capture into 128I, which decays with a half life of 24.99 m to 128Xe (Xenon gas can be easily periodically purged from the device) which is stable (93.1%), and to 129Te (6.9%) which is decaying into 129I, adding slightly to the initial sample. Therefore, transmutation can be performed with Iodine chemically separated from the FFF's, though with a number of neutroncaptures slightly larger (+23%) than in the case of an isotopically pure 129I sample. (3) Zirconium has large produced (chemical) mass (4.2 ton), with about 75.48×103 GBq/reactor of 93Zr (19.3% by weight). The Class A storage volume is small: 583 m3, about 1.2% of the one of 99Tc. In addition, being a metal, it can be diluted for instance in Lead or Copper, and be kept out of the biological cycle essentially indefinitely. Notwithstanding, it would be possible to transmute it, but in practice only with prior isotopic separation. Since the other Zr isotopes are stable and the specific activity of 93Zr is small (0.00251 Cie/g), isotopic separation is costly but not difficult. In view of the small environmental impact of Zr, the necessity to transmute the element is questionable. (4) Cesium (135Cs, 442 kg, 18.87×103 GBq/reactor), is a rather delicate case, since it is mixed with 137Cs with a high specific activity (87 Cie/g) and which is one of the most intense components of the FF's activity at short times. Straight transmutation of the chemical isotopic mixture is possible, but it will not affect appreciably the 137Cs, which has a very small capture integral (Ires=0.616 b). But both the stable 133Cs (49% by weight) and the unwanted 135Cs (17.7%) have to be transmuted, with a correspondingly greater neutron expense, 2.78 times larger than if a prior isotopic separation is introduced in order to extract a pure 135Cs. The simultaneous transmutation of both isotopes, with large Ires is technically feasible, since they lead to short-lived elements which end up in a short time to stable Barium isotopes. However, handling large amounts of strongly active material (29 Cie/g for the chemical element) for the incineration procedure is not without problems and it should be discouraged. On the other hand, the Class A dilution volume is small, 510 m3, but some concern has been expressed about the possibility that leaks may occur from the repository to the environment during the long life of the isotope. If those concerns were to be confirmed, transmutation of Cesium will become necessary. It could be performed in a few hundred years from now, when the 137Cs has sufficiently decayed and if deemed necessary at this point. (5) Tin (126Sn, 29.5 kg, 31.01×103 GBq/reactor) is a low-activity metal, for which a small volume, 239 m3, class A storage is required. The resonance integral, Ires=0.139 b, is too small for a realistic transmutation rate. Hence, our method is not immediately applicable to this element. Fortunately, the nature of the element is such that it ensures good containment in an appropriate metallic matrix, and therefore it appears entirely safe to keep it as Class A indefinite storage. (6) Selenium (79Se, 6.57 kg, 16.97×103 GBq/reactor) is also a low-activity material, for which a small volume, 131 m3, class A storage is required. The dominant Ires=56 b is the one of the element to be transmuted, the other isotopes being either of small concentration or of smaller Ires. Incineration could proceed with the chemical mixture, also taking into account the small size of the stockpile, 63.3 kg after 40 years of operation. Isotopic separation is also possible, since the specific activity of 79Se is 0.07 Cie/gr. Little is known on Selenium diffusion in the environment, though it may be significant, since it is similar to Sulfur. In case of doubt, transmutation is perfectly feasible. For these reasons it would seem appropriate to give high priority to the transmutation of 99Tc and 129I. The residual Class A definitive storage volume is thus reduced from 53971 m3 to 1463 m3, namely by a factor 37. Transmutation of 79Se may also be advisable, especially in view of the small quantities. Transmutation is not possible with 126Sn; for 135Cs, if needed at all, it must be delayed by several centuries in order to wait for the 137Cs to decay, unless an arduous, isotopic separation is performed. The characteristics of the source are evidently application-dependent. We concentrate first on the requirements of the Activator mode of operation of the Transmuter. The requirements of the Transmuter operated to decontaminate waste will be considered next. The Activator for medical and industrial purposes demands relatively small neutron intensities, though the required activity of the newly created radio-nuclide and the corresponding size of the initial sample to be activated depend strongly on the specific application and on the subsequent procedures of extraction and use. Many different types of compact neutron sources of adequate strength are commercially available, and may be relevant in various Activation applications with the present method. We list amongst them, in increasing function of the neutron intensity (1) Radioactive sources, like for instance Am—Be and similar, which produce currently about 2.1×108 neutrons for 100 Cie of α-source, or Actinide sources like 252Cf which have spontaneous fission probability and produce about 3.0×109 n/Cie. Though the neutron intensity generated with sources is more modest than the one achievable with Accelerators, the device is completely passive and offers much greater simplicity and consequently lower cost. (2) High voltage sets based on D-T or D-D reactions, which produce up to 1010 n/sec for 100 μA of accelerated current to some 300 keV. . (3) Small accelerators (Cyclotrons, RF-Q, LINAC's) with ultimate current capability of several mA, which produce typically ≧1013 n/sec with the help of accelerated currents of the order of 100 μA at several MeV, and which are already widely used in hospitals for isotope production, for instance for PET applications. (4) Spallation sources from high energy proton beams hitting a Lead or Bismuth target block. As shown later on, the Activator target for large beam power has to be liquid to ensure appropriate cooling of the beam-dissipated power which, in the example, is of the order of several hundred kWatt. High energy protons are extremely prolific neutron sources. For a possible application of the Activator on a large industrial scale and as a dedicated machine, one might consider a 100-200 MeV LINAC or compact Cyclotron and an average current of a few mA. Neutron production rates well in excess of S0=1016 n/sec can be easily obtained with such arrangement. The corresponding neutron flux in which the activation sample is normally located, is of the order of 1014 n/cm2/sec, quite comparable with the flux of the largest Power Reactors. Taking into account the fact that the capture process is-further enhanced by resonance crossing, it is evident that the present method becomes largely competitive with Reactor-driven activation. This is in particular valid for 99Mo (99mTc), which is plagued by a very small capture cross-section of 140 mb for thermal (reactor) neutrons, but with a large resonance cross-section, and for which a much more complicated extraction from the 235U-fission fragments obtained from spent reactor fuel is currently used. (5) Leakage neutrons from the core of a critical (Reactor) or Sub-critical (Energy Amplifier) device. Since these devices produce vast amounts of power (GigaWatts), the residual neutron flux is very large. Because these neutrons are anyway lost from the Core, the Transmuter can be run “parasitically”. The neutron energy spectrum must however be matched to the application. If, as most likely, resonance-driven captures are exploited in a Lead diffusing environment, the core must produce fast neutrons, with energies which are well above the resonances to be exploited. The neutron source for a Waste Transmuter must be much stronger, since, as already mentioned, the sample must undergo a complete transformation. Neutrons may be directly produced by a Spallation source of the type (4) above or, even better, by a “leakage” source of type (5). In addition, neutrons must be efficiently captured by the elements to be transmuted. The minimal amount of captured neutrons required in ideal conditions is listed in Table 2, where neutron units are kilograms (1 kg of neutrons corresponds to 5.97×1026 neutrons) and elements are the ones listed in Table 1. In reality, an even larger number is required since the capture and subsequent transmutation probability αt is less than unity. The proposed scenario in which only 99Tc, 129I and 79Se are transmuted requires, according to Table 2, an ultimate 11.29/αt kg of neutrons dedicated to transmutation. TABLE 2Minimal neutron requirements for full transmutationof most offending, long-lived FF's of the fulldischarge (40 years) of a standard LWR.Neutrons (kg)for fullIsotopic massChemical MasstransmutationElementkg% all FFkg% all FFIsotopicChemical99Tc843.2.2158432.2158.518.51129I196.20.515255.420.6711.521.9893Zr810.42.1294198.411.038.7145.14135Cs442.21.1622502.86.5773.27—126Sn29.480.07795.700.251——79Se6.570.01763.330.1660.08320.802 In the case of a source of type (4) above, one needs generally a higher energy and higher current proton beam. For proton kinetic energies of the order of or larger than 1 GeV and a Lead Spallation Target, the neutron yield corresponds to 40 MeV/neutron, i.e. 6.4'10−12 Joule/n. One kg of neutrons will then require 1.061×109 kWh, or 3.029 MWatt of average beam power during the illustrative 40 years of operation. Assuming an acceleration efficiency of 0.5, this corresponds to 6.05 MWatt of actual electric power. The ultimate 11.29 kg of neutrons will therefore require 68.40 MWatt of electric power for the whole duration of the LWR operation, corresponding to 6.8% of the electricity produced by the plant. Including capture efficiency etc., the fraction of electric power produced by the LWR needed to produce an equivalent transmutation of the selected long-lived FF's is of the order of 10% of the produced power. Evidently, off-peak energy production could be used. This installed power and the associated large scale Accelerator represents a considerable investment and running costs. It would be more profitable to make direct use of fission-driven neutron multiplication intrinsic in the necessary parallel elimination of the TRU's (which has the additional advantage of being eso-energetic) i.e. choosing a source of the type (5) above. The simultaneous, complete incineration of the TRU's (10.178 ton) will produce a number of neutrons of the order of 106.02×αf kg, where αf is the fraction of neutrons generated per fission (in these indicative considerations, we have assumed that the average neutron multiplicity/fission is 2.5) which is made available to transmutation of FF's. We conclude that, in order to proceed concurrently with the TRU (the complete fission of the TRU's will produce an additional amount, of FF's (10.178 ton), which will have to be transmuted as well, in addition to the 38.051 ton of FF's from the waste of the LWR's ; this will be discussed in more detail later on) and FF elimination, αt×αf=0.106, implying a very efficient utilisation of surplus neutrons from the TRU's incineration process. It will be shown, however, that, it can be attained thanks to the present method. With the help of the method here described, high rate of neutron captures can be achieved with relatively modest neutron fluxes. As a consequence, a practical, neutron-driven Activator can be achieved with simple and relatively cheap, small Accelerators which do not require large installations, like for instance is the case for Nuclear Reactors. The environmental impact and safety are far easier, since the Activator is not critical and it produces little extra activity apart from the one in the sample. The activation of the Lead block is limited mainly to the 209Pb isotope, which decays with a half-life of 3.2 hours into the stable 209Bi. Activation of the Graphite and of the Steel structures are also equally modest. The large Lead block constitutes a natural shielding to this activity, mostly concentrated in the centre of the Activator. All activated materials at the end of the Life of the installation qualify for direct LLW-Class A for surface storage, which is not the case for the Nuclear Reactor spent fuel. Licensing and operation of a low energy accelerator are infinitely easier than in the case of a Reactor. In view of these considerations, of the growing need for radio-isotopes for medical and industrial applications and of the comparable efficiency of activation, the accelerator-driven neutron Activator based on the proposed flux enhancement method constitutes a valid alternative to the current radio-isotope production processes. Considering the variety of short-lived isotopes needed, for instance, for medical applications (see Tables 7, 8 and 9), a general-purpose accelerator can simultaneously produce those radio-isotopes for which charged particle activation is best suited and also those isotopes for which neutron capture is most convenient by means of an Activator as disclosed herein, thereby eliminating the need to rely on Nuclear Reactors in a general-purpose (local or regional) facility. This can be realised with relatively modest means and smaller environmental impact. In the case of a Waste Transmuter, more powerful neutron sources are needed for the complete transmutation into stable elements of unwanted, long-lived radioactive waste. This can be achieved in principle with larger Accelerators and Spallation sources. In the case of the spent fuel from LWR's, since these elements have in general to be eliminated concurrently with the fissionable TRU waste, one can use the extra neutrons produced by their fission as a source for the Waste Transmuter, adding the Waste Transmuter to a fast Energy Amplifier or a Fast Reactor dedicated to the burning of the TRU's. The high efficiency of the present method ensures that both unwanted stockpiles can be effectively and simultaneously eliminated in the process. In order to illustrate the method, we present first some simple, analytic considerations. These qualitative results are approximate. However, they provide some insight in the dynamics of the method. More detailed computer simulations will be reported further on. Assume a large volume of transparent, diffusing medium, large enough in order to contain the neutron evolution. The source, assumed point-like, is located at its centre. Consider a neutron population in a large, uniform medium of N scattering centres per unit volume, with very small absorption cross-section cabs and a large scattering cross-section σsc. All other cross-sections are assumed to be negligible, as it is generally the case for neutrons of energy substantially smaller than 1 MeV. Since the angular distribution of these collisions is almost isotropic, they also have the important function of making the propagation of neutrons diffusive, and therefore maintain the neutrons “cloud” within a smaller containment volume. The neutron flux φ(x,y,z) in such a volume is defined as the number of neutrons crossing the unit-area from all directions per unit time. At this point, the energy spectrum of the neutrons is not considered, namely the flux (and the corresponding cross-sections) are averaged over the energy spectrum. The reaction rate ρx, defined as the number of events per unit time and unit volume, for a process of cross-section σx is given by ρx=φNσx=φΣx, where Σx=Nσx stands for the macroscopic cross-section for the process x (x=sc for neutron elastic scattering, x=abs for neutron absorption, x=capt for neutron capture). For a steady state, Fick's law leads to the well-known differential equation: ∇ 2 ⁢ ϕ - Σ abs D ⁢ ϕ = - S D [ 3 ] where S is the neutron source strength, defined as the number of neutrons per unit volume and time, and D=1/(3εsc) is the diffusion coefficient for isotropic scattering. For anisotropic scattering, a correction must be introduced, i.e. D=1/[3Σsc(1-μ)], where μ=<cos θ> is the mean value of the cosine of the diffusion angle (note that for relatively slow neutrons and high A, μ≈0). As already pointed out in Paragraph 1.1, two indicative materials—amongst many—can be exemplified as practical diffusing media for the present method, namely Carbon (using the density of reactor-grade graphite, d=1.70 g/cm3 and thermal neutrons cross-sections), for which D=8.6 mm and Lead, for which D=10.1 mm. These media exemplify the alternatives of quickly and slowly thermalising media, respectively. In order to achieve an effective rate of activation, the neutron flux must be as high as possible. If we place a point source at the origin of the coordinate system, Equation [3] will hold everywhere with S=0, except at the source. The approximate solution of the differential equation is: ϕ ⁡ ( r ) = S 0 ⁢ ⅇ - κ ⁢ ⁢ r 4 ⁢ ⁢ π ⁢ ⁢ Dr ; κ = Σ abs D = 3 ⁢ ⁢ Σ SC ⁡ ( 1 - μ ) ⁢ Σ abs [ 4 ] where S0 is the rate of neutrons from the source per unit of time (n/sec). The elastic scattering cross-section being large and the absorption cross section very small, D is a small number (of the order of the centimetre), while 1/κ is large (of the order of meters). For a region close to the source, namely κr<<1, the flux is given by φ(r)≈S0/(4πDr), namely is considerably enhanced with respect to the flux in absence of diffuser φ0(r)≈S0/(4πr2). For a typical sample distance of r=30 cm, the enhancement factor F=φ(r)/φ0(r)=r/D is very, substantial for instance for Carbon where F=30/0.86=34.88 and for Lead where F=30/1.01=29.7. The diffusing medium is acting as a powerful flux enhancer, due to multiple traversals. In addition, the energy spectrum of neutrons is preferably matched to the largest values of the capture cross-section of the relevant isotope. The energy spectrum of a bare source is not optimal because its energy is generally too high to produce an effective capture rate. Therefore, an energy matching (moderation) must be performed before utilisation. Examples already given in which the interesting cross-sections lay in the resonance region are the cases of Iodine activation and the production of 99Mo(99mTc) from a Molybdenum target. As already pointed out, in this case the transparent, diffusing material must have in addition a large atomic number. The energy E of the neutrons is then progressively shifted in a multitude of small steps by a large number of multiple, elastic collisions (as already pointed out, below a few hundred keV and in a transparent medium, the only dominant process is elastic scattering). The minimum emerging kinetic energy T′ min (i.e. for a maximum energy loss) of a neutron of energy T0 in collision with a nucleus of atomic number A is given by T min ′ = T 0 ⁡ ( A - 1 A + 1 ) 2 [ 6 ] which evidently suggests the largest possible A to minimise the rate of energy loss. For large A, isotropic scattering is an excellent approximation. The average, logarithmic energy decrement ξ is then ξ = - ln ⁢ 〈 T ′ 〉 T 0 = 1 - ( A - 1 ) 2 2 ⁢ A ⁢ ln ⁡ ( A + 1 A - 1 ) [ 7 ] The logarithmic energy decrement for Lead is very small ξ=9.54×10−3. The average number ncoll of collisions to slow down from 0.5 MeV to 0.025 eV (thermal energies) os mcoll=lm (0.5 MeV/0.025 eV)/ξ=1.76×103. The elastic cross-section, away from the resonances, is about constant down to thermal energies and large (σsc=11 b). The total path length lcoll to accumulate ncoll collisions is then the enormous path of 53.4 meters. The actual displacement is of course much shorter, since the process is diffusive. As a consequence of the property that neutrons loose at each step a constant fraction of their energy, the energy spectrum, generated by a high energy neutron injected in the diffuser is flat when plotted in the variable dE/E=d(log(E)). Neutrons scan progressively the full energy interval down to thermal energies, “seeking” for large values of the capture cross-section of the added impurities due to strong resonances. This method is evidently profitable provided that strong resonances exist elsewhere than at thermal energies. It is a fortunate circumstance that this is the case for several of the isotopes of practical interest. If a small amount of impurity to be activated is added to the transparent medium, it will capture some neutrons. In general the absorbing cross-section has a complicated behaviour and it varies rapidly as a function of the neutron energy, due to the presence of resonances. We introduce the survival probability Psurv(E1,E2), defined as the probability that the neutron moderated through the energy interval E1→E2 is not captured. The probability that a neutron does not get captured while in the energy interval between. E and E+dE is [1−(Σabs/Σabs+Σsc))(dE/Eξ)] where Σsc and Σabs are respectively the macroscopic elastic scattering and absorption cross-sections. Such probability is defined for a large number of neutrons in which the actual succession of energies is averaged. Combining the (independent) probabilities that it survives capture in each of the infinitesimal intervals, Psurv(E1,E2) is equal to the product over the energy range: P surv ⁡ ( E 1 , E 2 ) ≅ ⁢ ∏ E 1 E 2 ⁢ ( 1 - Σ abs Σ sc + Σ abs ⁢ ⅆ E ξ ⁢ ⁢ E ) = ⁢ exp ⁡ [ - 1 ξ ⁢ ∫ E 2 E 1 ⁢ Σ abs Σ sc + Σ abs ⁢ ⅆ E E ] ≈ ⁢ exp ⁡ [ - 1 σ sc Pb ⁢ ξ ⁢ ( N imp N Pb ⁢ I res ( imp ) ⁡ ( E 1 , E 2 ) + I res ( Pb ) ⁡ ( E 1 , E 2 ) ) ] [ 8 ] where NPb and Nimp are the number of nuclei per unit volume for Lead and the added impurity, respectively, and in the good approximation that the elastic scattering on Lead is dominant and approximately constant, namely Σsc≈σscPbNPb=const>>Σabs. The resonance integrals Ires (E1,E2) for Lead and the added impurity are defined as I res ( x ) ⁡ ( E 1 , E 2 ) = ∫ E 2 E 1 ⁢ σ abs ( x ) ⁢ ⅆ E E ; x = Pb , imp [ 9 ] The (small) probability of absorption in the same energy interval is given by P abs ⁡ ( E 1 , E 2 ) = 1 - P surv ⁡ ( E 1 , E 2 ) ≈ 1 σ sc Pb ⁢ ξ ⁢ ( N imp N Pb ⁢ I res ( imp ) ⁡ ( E 1 , E 2 ) + I res ( Pb ) ⁡ ( E 1 , E 2 ) ) [ 10 ] which exhibits the separate contributions to capture of the diffusing medium and of the added impurity, weighted according to their respective resonance integrals. The value of the normalizing cross-section in the denominator is σscPbξ=0.105 b, to be compared with the integral over the resonances Ires=150 b for 127I, Ires=310 b for 99Tc and Ires=0.115 b for natural Lead. For instance, in the case of the 99Tc Waste Transmutation, the capture probability will be enhanced over the fractional atomic concentration of the impurity N Nimp/NPb by a factor (310 b)/(0.105 b)=2.95×103. In order to reach equal capture probabilities, in 99Tc and Lead, the diffused impurity atomic concentration needed is only Nimp/Npb=(0.115 b)/(310 b)=3.70×10−4, namely 1.76×10−4 by weight. The resonance integral as a function of the energy interval for the main elements of Table 1 and relevant to the application as Waste Transmuter is given in FIG. 1, where the quantity Ires(x)(Emin, 1 MeV) is plotted as a function of the lower energy limit Emin. The value for any energy interval can be easily worked out through the obvious formula Ires(x)(E1, E2)=Ires(x)(E1, 1 MeV)−Ires(x)(E2, 1 MeV). The Figure evidences the large values of the resonance integrals for all relevant elements, with the exceptions of 126Sn (this confirms the unsuitability of 126Sn for the present transmutation method) and of natural Lead. It is also apparent that, while the main contribution to the integral in the case of Lead comes for energies >1 keV, the elements to be transmuted have dominant resonance captures (steps in the graph) which are dominant at lower energies. FIG. 1 also displays the values of Ires(Emin, 1 MeV)/σscPbξ, a dimensionless quantity (see Formula [10]) which gives the capture probability once multiplied by Nimp/NPb. For instance, the Iodine preparation for medical analysis to be irradiated in the Activator is likely to be a specific chemical compound with a variety of other elements in it (see Tables 7 and 8). In compounds made of several elements, a simple generalisation of Formula [10] indicates that the capture probabilities will be proportional to the values of the resonance integrals given in Appendix 1, weighted according to the atomic concentrations of each element. The compound to be exposed in the mentioned example is most likely Sodium Iodide (NaI). Fortunately, the Na resonance integral, Ires=0.26 b is much smaller than the one of Iodine, Ires=150 b. The activation (24Na) of Sodium will therefore be only 1.73×10−3 of the one of Iodine. The additional dose given to the patient is completely negligible. In addition, the half-lives of the two compounds, the wanted 128I and the unwanted 24Na, are 24.99 m and 14.96 h, respectively, i.e. in the ratio 2.78×10−2. The activity of the latter will then be 1.73×10−3×2.78×10−2=4.83×10−5 that of the former, of no effect for the measuring devices. In the case of Molybdenum (98Mo, Ires=7.0 b), in the form of a salt, for instance Na2MoO4, some captures occur in 23Na, leading to the unstable 24Na. The resonance integral of 23Na is more significant than in the previous example, since the 98Mo resonance integral is smaller (Ires=6.54 b), and it may constitute a problem, though the half-life of 24Na is of 14.96 h, i.e. shorter than the one of 99Mo. However, in the separation of the decay product 99mTc, the Na is generally retained. Some care must be exercised in order to ensure that a sufficiently small amount of 24Na is ending up in the patient, as a leakage through the dissolution process and subsequent preparation of the clinical sample. If the irradiated sample is either metallic Mo or MoO3, such a problem does not arise, at the cost however of some additional chemical handling at the end of the exposure. Other most likely elements in chemical compounds are Carbon (Ires=0.0016 b) (this is valid both for the leading isotope 12C and the tiny, natural concentration (1.1%) of 13C ; the small, natural concentration of 13C produces through capture radioactive 14C, though in very small amounts since its resonance integral is small), Oxygen (Ires=0.0004 b), Nitrogen (Ires=0.85 b) and Hydrogen (Ires=0.150 b). Small amounts of captures in these elements fortunately with small Ires—are harmless. In particular, 14N produces 15N, 12C produces 13C and Hydrogen produces Deuterium, which are all stable elements. The Deuterium contamination in natural Hydrogen (0.015%) can produce Tritium, but fortunately the resonance integral of Deuterium is extremely small, Ires=2.3×10−4 b. The small isotopic concentration (0.37%) of 15N in natural Nitrogen has a extremely small resonance integral, and is β-decaying to 16O with a half-life of 7.13 s, too short to reach the patient. Another element which could be present is Phosphorus. Its resonance integral is extremely small, Ires=0.0712 b. It leads to the 14.26 d isotope 32P, which is a pure β-emitter, with <Eβ>=695 keV and no γ-emission. Finally, we mention the case of Chlorine. Captures in 35Cl (75.77%, Ires=12.7 b) lead to the very long-lived 36Cl (τ1/2=3.01×105 y, β-, no γ) element which is completely harmless, and 37Cl (24.23%, Ires≈2.47 mb) has an extremely low production cross-section for 38Cl (τ1/2=37.24 m). Other chemicals which may be deemed necessary must be separately examined in view of their capture probability and the possibility of introducing harmful radioactive isotopes in the patient. The above formulae are only very approximately valid, and give only the qualitative features of the, phenomena. For instance, in such linear approximation, each element is contributing, so to say, independently. However, if a resonance is strong enough to absorb a major fraction of neutrons, it may “shield” other resonances occurring at lower energy. Then, the element which has a dominating resonance group at higher energies can void the captures of the elements “downstream”. This effect may be very important. The lethargy is modified by the elastic part of the resonance. The flux is locally decreased (dip) due to the shorter path needed to make the collision. Finally, the complexity of the geometry of a realistic device cannot be easily accounted for anlytically. In practice, computer simulations with the appropriate time evolution, are the only valid methods to predict with precision the performance of the device. These calculations use a Montecarlo method and the actual cross-sections for the interactions of particles inside the medium to simulate the propagation of the neutrons in the actual geometry of the Transmuter. A complete simulation programme has been developed in which the best known nuclear cross-sections have been used to follow the evolution of initially injected neutrons in a medium made of the appropriate mixture of isotopes and a definite geometrical configuration. Thermalization is taken into account, introducing the Maxwellian distribution of velocity for the target nuclei. Cross-sections from Nuclear Data bases have been used, and secondary decays have been included. A large number of neutrons are thus followed in their fate inside the device. The validity of the programme has been verified by comparing its predictions with a large number of different experimental data. These simulations have been found in excellent agreement (to better than the present uncertainties, of the order of ±15%) with experimental results obtained at the CERN-PS (Experiment TARC-P211). We consider first the application of the Transmuter as Activator. In Table 3, we exemplify some of the results of such computer simulations, normalised to 1013 neutrons produced by the source (23 MeV protons on a thick Beryllium target) and injected in the Activator with the geometry described in Table 6. We have chosen a Molybdenum salt Na2MoO4 (other salts may be used instead, for instance derived from the Molybdic Phosphoric Acid H7[P(Mo2O7)6] nH2O; see Paragraph 5.3 herebelow for more details) in order to evaluate the effects of the other chemical elements and their activation. Out of the injected neutrons, 91.5% are captured inside the device and 8.5% escape. These neutrons are absorbed in the surrounding shielding materials. The bulk of the captures occur in the Iron box (36.0%) and in the Lead (46.8%). Most of these captures produce stable elements, with the exception of captures in 54Fe (2.40%) which give origin to 55Fe with a half-life of 2.73 years and in 208Pb (0.43%) which produces 209Pb, which decays with a half-life of 3.25 hours into the stable 209Bi. The captures in the graphite Moderator are small (0.51%) and produce a tiny amount of 14C through captures of the natural isotope 13C (3.25×10−4). TABLE 3Example of computer simulation for the Activator loaded with Na2MoO4.Captures are given for 1013 neutrons produced. Only radio-isotopeswith a half-life longer than 1000 s are listed.ElementMass (kg)CapturesCapt/gramDaughterelement12C347.55.181E101.491E513Cstable13C4.18803.250E9 7.760E514C5730y16O0.2213——17Ostable23Na0.15941.690E9 1.060E724Na14.95h54Fe3739.02.397E116.411E455Fe2.73y56Fe61330.03.48812 5.688E457Festable57Fe1497.01.015E116.780E458Festable58Fe193.91.459E107.524E459Fe44.5d92Mo0.04731.536E8 3.247E693Mo4.9E3y92Mo0.0473<<1.0E5 <<2.0E3 93mMo6.85h94Mo0.03011.100E8 3.652E695Mostable95Mo0.05241.485E102.835E896Mostable96Mo0.05552.150E9 3.874E797Mostable97Mo0.03211.650E9 5.142E798Mostable98Mo0.08191.360E9 1.660E799Mo65.94h100Mo0.03344.100E8 1.229E7101Mo14.61m204Pb702.35.539E117.887E5205Pbstable206Pb12210.05.348E114.380E4207Pbstable207Pb11250.04.102E123.646E5208Pbstable208Pb26800.04.284E101.599E3209Pb3.25h205Pb0.00311.000E7 3.270E6206PbstableTotals118074.09.155E12 Therefore, the activation of the structures is modest and leads to no specific problem even after long exposures. As expected, the activation of a complex chemical sample produces several undesirable, unstable elements which will be reviewed in more detail later on for specific examples. The energy spectrum of the neutrons captured in 98Mo is shown as a solid line (left-hand ordinate scale) in FIG. 2. The integrated capture probability (dotted line, right-hand ordinate scale) is further displayed as a function of the upper energy value of the integration. The thermal neutron contribution is very small, and resonant capture dominates, extending all the way to the highest energies. The phenomenology of the neutron capture process is nicely visualised by the behaviour of the energy spectrum near a strong resonant absorption (FIG. 3a). Calculations refer to the activation of a block of metallic Tellurium in the Activation Volume of the Activator of Table 6. Capture probabilities in the body of the Activator (Pb, Fe, etc.) are, as expected, essentially unchanged with respect to the previous example. The specific capture rate in 130Te, leading to 131I, is η=3.54×10−5 kg−1 of natural Tellurium. A dip (indicated with an arrow, at 23 eV) occurs due to local depletion due to the main 123Te isotope: neutrons from neighbouring regions rush in, but only after a number of scattering events which are needed to displace the flux, and which induce a significant energy shift because of the lethargy of the material. After recovery from the dip, the spectral level is lower, due to depletion of the neutrons due to captures. The energy spectrum of captures in 123Te (solid line, left-hand ordinate scale), and the integrated capture probability (dotted line, right-hand ordinate scale) are shown in FIG. 3b. The presence of the prominent peak at 23 eV and of other satellite peaks is evident. Finally, in FIG. 3c, we display the same quantities, but for the captures in 130Te. The capture rate is suppressed in correspondence of the dominant peak of 123Te, but the flux is later recovered and captures can occur also at thermal energies. Resonant captures of 130Te occur at relatively high energies, prior to the 123Te absorbing action. These captures will be preserved even if, because of larger Tellurium samples, the flux will be more significantly depleted. This example shows the delicate interplay in the succession of resonant captures in different elements of a compound. Finally, we briefly discuss the application as a Waste Transmuter. The computer programme has been used to describe the time evolution of the neutron fluxes and of the element compositions in the EA (see C. Rubbia, “A High Gain Energy Amplifier Operated with Fast Neutrons”, AIP Conference, Proceedings 346, International Conference on Accelerator-Driven Transmutation Technologies and Applications, Las Vegas, July 1994) The coupling between these two, models is essential to understand the operation of the Waste Transmutation, coupled with the EA. The EA is cooled with molten Lead, which surrounds the core. In this otherwise empty volume, the conditions described for the Transmuter develop naturally. This is evidenced by the neutron spectrum shown in FIG. 4, plotted at various distances above the core for a small cylindrical volume coaxial to the core cetre and about 1 metre from the axis. The first 5 spectra (labeled 1-5) correspond to different vertical segmented levels of the core, starting from the medium plane and rising each time by 15 cm. One can observe a very hard spectrum, which is required for instance in order to fission the TRU's. The subsequent five spectra (6-10), correspond to different vertical segmented levels in the Lead surrounding the core, in steps of 40 cm. All spectra are average spectra over the vertical bin. The spectra in the surrounding Lead show the characteristic flattening due to the iso-lethargic condition, and enrich dramatically the part of the spectrum which is relevant to transmutation (1 to 1000 eV). In segments 8 and 9, we have introduced a small, diffused contamination of 99Tc at the density of 2.686 mg/cm3, equivalent to a mass concentration of only 260 p.p.m. with respect to the Lead. The capture lines corresponding to the leading 99Tc resonances are prominent, corresponding to a strong absorption as indicated by the large drop of the flux in the resonance crossing. This is better evidenced in FIG. 5, where the spectrum in segment 8 (volume 0.409 m3) is plotted in linear scale. In particular, one can see the diffusive refill of the spectrum, due to the rushing in of the neutrons from the region with no. 99Tc doping. The programme can be used to study both the time evolution of the burning inside the EA and the subsequent reactions in the Transmuter. This is evidenced in FIG. 6, where the concentration of relevant elements as a function of the burn-up in the EA is shown for segment 8 (0.409 m3) in which the 99Tc doping is inserted initially. While the 99Tc, initially with a density of 2.686 mg/cm3, is rapidly transmuted with a 1/e constant of 82 GWatt day/ton, the daughter element 100Ru builds up correspondingly. The large transformation rate of the 99Tc into the stable element 100Ru is followed by small capture rates to form 101Ru, and possibly some 102Ru. It is noted that all the indicated Ruthenium isotopes are stable The subsequent elements which may be produced by successive captures are also favourable: 103Ru and 104Ru are stable, while 105Ru quickly decays into the stable 105Pd. Also, 106Pd is stable, the first long-lived isotope being 107Pd, which has a half-life of 6.5×106 years. However, its production rate is truly negligible, taking into account that as many as eight successive neutron captures must occur in the same nucleus. The decay constant for transmutation of 99Tc is about 82.1 GWatt day/ton, corresponding to less than 3 years for the nominal EA power (1.0 GWatt, thermal). These curves evidence the feasibility of complete elimination of Technetium in the periphery of an EA with a reasonable time constant. More detailed configurations and actual rates of transmutation will be discussed later on. Incidentally, we also remark that if the materials to be transmuted were directly inserted in the core, the transmutation rate would be much smaller, since there the neutron flux is concentrated at energies in which the captures by the long-lived FF's have a very tiny cross-section. The size and the kind of the neutron source are clearly related to the application. We consider first the case of the Activator. The main parameter is the angularly integrated neutron production rate S0, since the actual angular distribution at the source is quickly made isotropic by the Lead Diffuser (see Chapter 4 herebelow for more details). Likewise, the energy spectrum of the initially produced neutrons is relatively unimportant since, as already explained, the inelastic processes in the Diffuser quickly damp the neutron energy down to about 1 MeV, where the lethargic slow-down of the neutrons is taking over. Therefore, the neutron capture efficiency for activation η and, more generally, the geometry of the Activator are relatively independent of the details of the realisation of the source. In the case of the activation of natural Iodine, it is likely that a small sample—of the order of a fraction of a gram—must be activated for each exposure to a level requiring a cyclotron or similar accelerator with a neutron production rate of few times 1013 neutrons over the full solid angle. This can be obtained with an energy of the order of 10 to 30 MeV and a beam current of the order of mA's, which is also suited for production of isotopes for PET examinations. Therefore, a combined facility may be envisioned. In the case of a large industrial production of radio-nuclides, like for instance 99Mo (99mTc), 131I or of Fissium from Uranium fissions it may be worth considering similar currents but higher proton energies, in the region of a few hundred MeV, with a correspondingly larger S0. Activation, which is proportional to S0, can then be performed within much smaller samples, which is, as will be seen, a considerable advantage especially in the case of portable 99Mo (99mTc) dispensers. At the other end of the scale, the production of small activation with a simple device using a neutron-emitting radioactive source is worth mentioning, since it might be of interest for applications which require a very weak source (<<mCie) of radio-isotopes, but at low cost and operational simplicity. The overall neutron yield from a thick Be target bombarded with a beam of protons of energy Ep=23 MeV is reported in the literature (see H. J. Brede et al, Nucl. Instr. & Methods, A274, (332), 1989 and references therein). Integration over the angular distribution (M. A. Lone et al, Nucl. Instr. & Methods 143, (331), 1977 ; see also M. A. Lone et al, Nucl. Instr. & Methods 189, (515), 1981) gives the total neutron yield S0=1.66×1014 n/sec/mA (for energies greater than 0.4 MeV′, corresponding to a neutron flux φ(r)=0.654×1012 cm−2 s mA−1 at r=20 cm from the source, according to the formula φ(r)≈S0/(4πDr), which exhibits the Lead enhancement factor (D=1.01 cm). It is also noted that the flux is fallina like the inverse of the distance (1/r), i.e. more slowly than in empty space where the flux is proportional to the solid angle from the source (1/r2) . Already for a current of 10 mA, which can be generated by modern cyclotrons, our system leads to the remarkable flux φ(r)=6.5×12 cm−2.s−1, typical of a Reactor. TABLE 4Neutron yield for energies >0.3 MeV, integrated over all angles.Integrated flux, S0ReactionEnergy (MeV)(1013 n/sec/mA)9Be(p, n)14.86.818.010.223.016.69Be(d, n)8.01.514.88.618.012.323.019.67Li(p, n)14.85.118.08.123.010.37Li(d, n)8.01.014.87.718.012.123.019.5 Other target materials can be used, in particular 7Li, with comparable yields. However, in view of the lower melting point, Lithium targets are more complicated. A summary of yields for different beams and (thick) targets is given in Table 4. The neutron yield is a growing function of the proton kinetic energy Ep. Fitting of measurements at different energies leads to the simple empirical formula S0(Ep)=4.476×1011×Ep1.8866 valid for neutrons of energy greater than 0.4 MeV. For instance, for a proton kinetic energy Ep=50 (15) MeV, the neutron yield is increased (decreased) by a factor 4.33 (0.45) when compared to Ep=23 MeV. Since the beam power E0 for a current ip is ipEp, the neutron yield for a given beam power is rising proportionally to E00.886. Neutrons can be produced also with other incident particles, in particular deuterons and alpha particles. For a given incident energy, the forward neutron yield of deuterons is substantially higher than for protons, but as relevant in our application, the angle integrated flux is comparable to the one of protons, as shown in Table 4. For instance, at Ed=23 MeV, the integrated, yield is S0=1.96×1014 n/sec/mA. The yield for incident α-particles is substantially lower. In view of the associated simplicity and their high neutron yield, proton beams seem to be optimal for the present application. An important technical element is the beam power to be dissipated in the target. The many different types of targets which are commonly used in association with particle beams of the characteristics considered here are generally applicable to our case. The effective beam area is typically of the order of several squared, centimetres. We note that the target thickness required to stop the beam is relatively small, i.e. of the order of 4 mm for Ep=25 MeV. The thermal conductivity of Beryllium is large (k=2.18 W.cm−1.° C.−1) and its melting point conveniently high (1278° C.). Over the thickness L chosen equal to the particle range, the temperature drop ΔT due to conductivity, for a surface power density q due to the beam (W/cm2), is given by ΔT=qL/2k, neglecting the variation of the ionisation losses due to the Bragg peak (including this small effect will actually improve the situation since the energy losses are largest at the end of range, which is closer to the cooling region). Setting q=5×103 W/cm2 and L=0.4 cm, we find ΔT=458° C., which is adequate. Cooling of the face of the target opposite to the beam can be performed in a variety of ways. Assuming water circulation (it has been verified that the presence of the water coolant has negligible effects on the neutronics of the device), the required water mass flow w is w=Wbeam/ΔTcρc, where Wbeam is the beam power (Watt), ΔTc is the allowed temperature change of the coolant and ρc (4.18 Joules/cm3/° C.) the heat capacity of the water coolant. Setting Wbeam=25 kWatt (1 mA @ 25 MeV), ΔTc=70° C., we find w=0.085 litre/sec, which is a modest value. For higher beam powers, it is convenient to tilt the target face with respect to the beam direction. If φ is the incidence angle of the beam on the target plane (φ=90° for normal incidence), the actual target thickness is reduced by a factor L×sinφ, and the beam surface power density by a factor q×sinφ, with consequent advantages in the target heat conductivity and cooling surface. Two types of standard neutron sources appear interesting. In the first type of sources, the neutrons are produced by the (α,n) reaction on Beryllium mixed as powder with a pure α-emitter, like for instance 241Am, 238Pu, 244Cm and so on. The main disadvantage of this source is the small neutron yield, typically 2.1×106 neutrons/s for 1 Curie of α-source. Therefore, a pure α-emitter of as much as 500 Cie is required to achieve the flux of 109 n/sec. The decay heat generated by such a source is 17.8 Watt. Another attractive type of source is an Actinide with high probability of spontaneous fission, like for instance 252Cf, which is an α-emitter with 3.1% probability of spontaneous fission, thus generating 0.031×2.8=0.087 fission neutrons at each disintegration. The above-quoted flux is then obtained with a much smaller source, of 109/(3.7×1010×0.087)=0.311 Cie. The half-life of the source is 2.64 years. For instance, a 10 Cie source of 252Cf produces 3.2×1010 neutrons/s, which has sufficient intensity to produce 0.01 GBq samples of 99mTc with a natural Molybdenum activator of 20 gram. In some diagnostic applications (see Table 9), smaller activities may be sufficient. Intermediate between the performance of the Accelerators and of the sources are the D-T high voltage columns, which produce 14 MeV neutrons at some 300 keV, with the reaction (d,n) on a Tritium-enriched target. Much higher neutron fluxes are possible with proton beams of high energy impinging a Spallation target. High energy protons will simply be absorbed in the Lead Buffer Layer, which will also act as spallation target. In view of the large power deposited by the beam on a relatively large volume of the spallation target, appropriate design is required. For highbeam powers E0, the best arrangement is the one of liquid metal target. This technology and. associated geometry will be discussed later on. The spallation neutron yield produced by a high energy proton in a Lead Block of the indicated size is listed in Table 5, as a function of the incident proton kinetic energy Ep. TABLE 5Neutron yield with energies >1.0 MeV, integrated over all anglesfor the spallation process in Lead induced by a high-energy protonE0 (kWatt)ip (mA)ΦEpS0forfor(cm−2s−1mA−1)(MeV)n0(n/sec/mA)3 1016 n/s3 1016 n/s(r = 30 cm)100.00.3992.49E151203.012.036.55E12150.00.8985.61E15801.85.351.47E13200.01.7881.12E16536.92.682.93E13250.02.7631.73E16434.31.744.54E13300.04.1562.60E16346.51.156.82E13350.05.2913.31E16317.50.918.68E13400.06.9394.34E16276.70.691.14E14 The neutron multiplicity n0, defined as the average number of neutrons produced for each incident proton of kinetic energy Ep, is a rapidly rising function of the proton energy, which can be fitted above 100 MeV with an approximate empirical formula n03.717×10−5×Ep2+3.396×10−3×Ep with Ep in MeV. The integrated specific neutron yield S0 is a correspondingly fast rising function of Ep, of the order of 1.12×1016 n/sec/mA at Ep=200 MeV. At this energy, a beam current ip of the order of ip=2.68 mA is required for a neutron yield of the order of S0=3.0×1016 n/sec. It is therefore possible to achieve fluxes which are at least two orders of magnitude higher than the ones of the intermediate energy accelerator. The neutron flux φ at r=30 cm from the centre, where the activation sample is normally located, is of the order of 0.78×1014 n/cm2/sec, quite comparable with the flux of a large Power Reactor. Taking into account the fact that the capture process is greatly enhanced by resonance crossing (see Formula [10]), it is evident that our method becomes largely competitive with Reactor-driven activation. This is in particular valid for 99Mo (99mTc), which is plagued by a very small capture cross-section of 140 mb for thermal (reactor) neutrons, and for which the alternative but much more complicated extraction from the 235U-fission fragments from a Reactor is currently used. Evidently, these currents and energies are appropriate for an industrial implantation for large scale production of radio-isotopes, and in particular of 99Mo (99mTc), for which a large market exists. The activated Molybdenum (half-life of 65 hours), as described later on, is transported to the point of use (Hospital) with the help of an Alumina container, from which the 99mTc is extracted whenever needed. An industrial Accelerator capable of producing a beam energy of the order of several mA at an energy of the order of 150 to 200 may consist in a compact cyclotron of modest size (radius=few meters) fed with a High Voltage column of about 250 keV, as suggested by P. Mandrillon. Negative ions (H−) are accelerated instead of protons, since the extraction can be easily performed with a stripper. An alternative Accelerator design, proposed by LINAC SYSTEMS (2167 N. Highway 77 Waxahachie, Tex. 75165, USA), foresees a compact (average gradient 2 MeV/m) LINAC which is capable of currents of the order of 10 to 15 mA at energies in excess of 100 MeV. As already pointed out, the considerable beam power to be dissipated in the Spallation-Target diffuser suggests the possibility of using molten Lead (melting point 327° C.) or a eutectic Lead-Bismuth (melting point 125° C.) target. The operation is facilitated by the fact that the energy of the beam, because of its higher proton energy and range, is distributed over a considerable length. The liquid flow and the corresponding cooling can be realised with the help of natural convection alone. Power in excess of 1 MWatt can be easily dissipated in the flowing, molten metal. The operating temperature is of the order of 400° C., temperature at which corrosion problems are minimal. The beam penetrates the molten liquid environment through a window. In order to avoid damage to the window due to the beam, the beam spot at the position of the window is appropriately enlarged, typically over a diameter of some 10 cm. The neutron yields S0 achievable by proton Accelerators and different targets for a 1 mA proton current are summarised in FIG. 8. The alternatives of a Beryllium target and of a heavy Spallation target are displayed. We refer to the configuration for simultaneous elimination of the TRU waste and of the. Transmutation of long-lived FF's according to the previously described scenario (Paragraph 1.4). The source is preferably an Energy Amplifier (EA), although a Fast Breeder (FB) configuration may also be employed. In this scenario, the transmutations of both offending kinds of waste must be performed concurrently, namely at rates which are predetermined by the composition of the waste which has to be decontaminated. As already pointed out in paragraph 1.5, this implies that the product of the fraction αt of the fission neutrons which are made available for transmutation and of the fraction αf of these neutrons which are actually captured in the impurity, be of the order of αt×αf=0.106. In practice it is possible to “leak out” of the order of 20 to 25% of the neutrons of the core, without affecting appreciably the TRU incineration process which demands a sub-critical multiplication constant of the order of k=0.96 to 0.98. Similar considerations apply to a Fast Breeder, though the requirement of full criticality may be more demanding in terms of neutrons destined to the Core. This implies that αf≧0.5, which is a large number, but, as we shall see, achievable with the present method. The practical realisation of the activation device is schematically illustrated in FIG. 7a for the intermediate energy beam, and in FIG. 7b for the high energy beam and spallation source, respectively. Dimensions are approximate and they are not critical. The overall shape has been chosen somewhat arbitrarily to be cylindrical of roughly equal dimensions in the three axes (length=diameter). Obviously, any other shape is also possible. The device may be divided in a number of concentric functional layers, starting from the centre, where the neutron producing target is, located. (1) In the case of FIG. 7a, the Target 1, assumed to be of small size, is hit by the beam 8 of the Accelerator, transported through the evacuated Beam Channel 2. Of course, the Beam Channel 2 is unnecessary if the neutrons, are produced by a radioactive source. In the latter case, the tube 2 may be needed to extract the source from the device. The Beam Channel is surrounded by a first Buffer Layer 3. The purpose of this layer (r0≈25 cm of Lead, but not critical) is to provide a first diffusion V enhancement and isotropisation of the neutron flux from the source. The distribution of the flux is made largely independent of the actual angular distribution of the neutron-producing reaction. Most of the possible neutron sources have an energy spectrum which extends to several MeV, much too high to lead to a practical activation. The buffer layer provides as well a first substantial and quick reduction in the energy spectrum, which is naturally achieved through inelastic scattering processes like (n,n′), (n, 2n), (n,3n) and so on. These last two processes introduce as well a small but significant increase of the flux by neutron multiplication, typically of the order of several percent and which is enhanced for higher energy sources, like for instance in the case of 14 MeV neutrons from the D-T production reaction. At the exit of the Buffer Layer, the energy spectrum in the capture resonance region of the samples has become largely independent of the nature and initial spectrum of the source. The ideal material for the Buffer Layer is Lead or Bismuth, because of its small, diffusion coefficient D, large transparendy below the inelastic threshold (the Buffer layer must also be very transparent to the lower energy neutrons which diffuse throughout the volume of the Activator) and large inelasticity of the cross-sections in the MeV range. In the case of high energy Accelerator and Spallation, neutrons (see FIG. 7b), the beam 9 traveling in an evacuated pipe 10 is sent-directly through a Window 11 to the Molten Lead 12 which acts simultaneously as (thick) Target and Buffer. Because of the considerable power dissipated by the beam (up to several hundreds of kWatt), the Target/Buffer Layer is best realised, with molten Lead, or eutectic Lead/Bismuth mixture. The molten liquid is circulated by natural convection at speeds of the order of 1 m/s through a pipe 13 in which are inserted a Heat Exchanger 14 and a Supplementary (electric) Heater 15, in order to ensure circulation and a temperature adequate to prevent the liquid from solidifying also when the Accelerator is off. The rest of the Activator Block 16 is in accordance with that of FIG. 7a and with, e.g., the parameters of Table 6. (2) The Activation Region 4 surrounds the Buffer Layer. In such a region—again best realised with Lead because of its small D value and high neutron transparency—are embedded the samples to be activated, for instance inside narrow, thin tubes. Samples must be easily introduced and extracted from the block with a suitable tool, such as a pantograph tool. These samples must be finely distributed over the whole volume of the Activation Region in order (i) to make use of the whole flux. In correspondence with very strong resonances, the sample becomes completely absorptive, and all neutrons having the appropriate energy within the volume are absorbed. If the sample is concentrated in a small volume, only the relatively few neutrons present within the volume with the right energy will be absorbed. This can cause saturation phenomena. (ii) to avoid self-screening of the sample in the large cross-section energy regions which are the most efficient in the activation. The sample holders may need structural supports. For this purpose, low-activation, neutron-transparent materials like for instance Steel, Zircalloy or Carbon compounds or, preferably, some more Lead should be used. The thickness of the Activation layer 4 may be application-dependent. Typically, it may be a layer of thickness r1 in the 5-10 cm range, concentric to the Buffer Layer 3. Since the scattering length in Lead is very short, the conditions of absorption by the resonance do not propagate appreciably from the point of occurrence. The absorption of neutrons at the (strong) resonances of the sample is a “local” phenomenon. (3) The device must be as compact as possible. If the outer volume were to be completed only with diffusing Lead, because of its small lethargy it would become rather bulky and require many hundreds of tons of material. Furthermore, since the energy losses occur in very small steps and the resonance integral is not negligible, this lengthy process would produce a significant depletion in the flux due to resonant self-absorption in the Lead itself. On the other hand, as pointed out, the activation of the wanted sample is a local condition which does not immediately propagate in the whole device. Therefore, one can introduce a Moderation Region 6 made of a thin (Δr in the 5-10 cm range, d=2.25 g/cm3) region made for instance with Carbon (Graphite) immediately beyond the Activation Volume 4, preferably preceded by a thin (r2 of the order of a few centimetres, i.e. r2>D) Lead Buffer Layer 5. The presence of the Moderation Region 6, acting both as a “reflector” and as an “energy moderator” has very beneficial effects on the energy spectrum in the Activation Volume. In FIG. 9, the calculated differential energy spectrum in the Activation Region is plotted in the variable dn/d(log(E)) since, in this variable and for an idealised iso-lethargy behaviour, it is constant and energy-independent: deviations from flatness imply changes from iso-lethargic ideal behaviour. The four curves correspond to different thicknesses of the Carbon layer, Δr=0, 2.5, 5.0 and 15.0 cm, respectively. It is noted that, in the energy region where resonances are expected, the flux is substantially enhanced with respect to the case of zero thickness of the Carbon layer. A broad optimum is achieved for a thickness Δr of the order of 5 to 10 cm. If larger thicknesses are used, the thermal energy peak becomes prominent. The activation probability for a given (weak) sample, for instance in the case of 127I, is more than doubled with the use of a 5 cm Carbon Layer. The overall size of the device is also substantially reduced. The alternative of a Moderation region between the Buffer Layer and the Activation region has also been explored and it gives much worse results. The conclusion of these studies is that the thickness of the Moderation Region, within reasonable limits, is not critical with respect to the flux in the resonance region. A thicker Carbon moderator enhances the fraction of neutrons in the thermal region. The optimal amount of thermal neutron captures depends evidently on the actual energy and location of the resonances of the sample. A very thick Carbon slab will quickly move the spectrum to thermal energy, which could be beneficial in some cases. At any rate, the use of Lead near the sample is recommended in all cases, since it produces the best flux enhancement. (4) The Moderation Region is followed by a Lead Reflector 7, and the whole device is enclosed in a thick Iron Box (not shown) to guarantee mechanical stiffness and shield the remaining neutrons. Additional, absorbing material, like concrete or similar materials, possibly loaded with Boron to efficiently capture the few escaping neutrons may be used to ensure full radio-protection of the device. The actual dimensions of a typical device are listed in Table 6, with reference to some specific activation tasks. In practice, some of the parts may be fixed and some others may be changed according to the application which is selected. The neutron spectra in the various parts of the Activator, plotted in the variable dn/d(log(E)) are shown in FIG. 10 for the parameters of Table 6 and no appreciable capturing sample. One can remark the general, remarkable flatness of the spectra, showing that the system is close to the idealised iso-lethargy conditions. The flux is roughly constant in the central region, and it drops in the Lead Reflector 7 and even more in the Iron Box. The sharp peaks are due to resonant behaviour of Lead and Iron of the Activator. TABLE 6Typical dimensions of the components, as used in the computersimulations. All elements are concentric cylinders, see FIG. 7a.OuterOuterlengthradiusMaterial(cm)(cm)RemarksBeam Tube 2Steel4.0Thin, evacuatedtubeBuffer Layer 3Lead8025Activator 4Lead +8030Samples insertedSampleinsideLead Buffer 5Lead9035C- Moderator 6Graphite10040average density1.9 gr/cm2Out Reflector 7Lead20090Containing BoxSteel300120Shield & support In order to exemplify our method, the performance of the Activator for medical isotope production is briefly summarised. As already pointed out, transmutation rates are largely independent of the chemical binding and isotopic composition of the materials inserted in the Activator. They are also almost independent on the source geometry and on the process used for the neutron production, provided that the initial neutron energy is sufficiently high (>0.4 MeV). The asymptotic activation, in GBq/gram, of the activation material as a function of the neutron yield from the source is shown in FIG. 11 for the specific examples discussed above. The main radio-isotopes used in Medicine and the corresponding domains of application are listed in Tables 7, 8 and 9. We shortly review these applications, in the light of the new possibilities offered by the Activator. A main change which becomes possible is the systematic replacement in the Iodine applications related to diagnosis with the much short-lived 128I, with the following main advantages: (1) the much smaller dose to the patient, essentially limited to the time of the examination, since the half-life is only 25 m. (2) the possibility of activating in situ an already prepared appropriate chemical compound of pharmacological quality, which is directly introduced in the patient after passing through the Activator for a short exposure (the radiation damage of the preparation is negligible, in view of the shortness of the neutron exposure). The decay scheme of the 128I has a 7% electron capture probability with K-shell soft photons, which makes it similar to 123I (which has also a γ-line at 159 keV (83.3%)). The rest is a β-γ transition with <Eβ>=737 keV and with a γ-line at 442.9 keV (16.9%). It is also similar to 131I (with 131Xe (11.9 d)), which has a γ-line at 364.8 keV (81.2%) and <Eβ>=182 keV. Therefore, these three elements have all similar diagnostics potentials, for which the γ-lines are relevant. Table 7 summarises the diagnosis data relative to Iodine radio-isotopes. The variety of products used and the general applicability of the Pre-activation method are to be emphasised. TABLE 7Main Diagnosis Applications of 131I (half-life 8.02 days, γ-line at 364.8keV (81.2%)) and of 123I (half-life 13.2 hours,decay mode EC and a γ-line at 159 keV (83.3%)). Iodine-basedDOSESuggestedPROCEDUREpreparation(GBq)MethodTUMOR131I-variesvaries128I Activationof preparationADRENAL131I-iodomethyl-0.555-128I ActivationCORTEXnorcholesterol0.74 of preparationADRENAL131I-miodobenzyl0.0018128I ActivationMEDULLAguanidineof preparationKIDNEYS131I-oiodohip- 0.00074-128I Activationpurate 0.00148of preparation(HIPPURAN)THYROID131I-sodium iodide 0.000018128I ActivationUPTAKEof preparationTUMOR131I-sodium iodide0.185-128I Activation0.37 of preparationTHYROID SCAN131I-sodium iodide 0.00015-128I Activation(substernal) 0.00037of preparationTHYROID SCAN131I-sodium iodide0.37 128I Activation(body survey)of preparationBRAIN123I-HIPDM 0.185 128I ActivationPERFUSIONof preparationBRAIN123I-IMP0.111-128I ActivationPERFUSION0.185 of preparationADRENAL123I-miodobenzyl-0.185-128I ActivationMEDULLAguanidine0.37 of preparationTHYROID SCAN123I-sodium iodide 0.00148128I Activationof preparationTHYROID123I-sodium iodide 0.00074128I ActivationUPTAKEof preparation TABLE 8Main Therapy Applications of 131I (half-life8.02 days, γ-line at 364.8 keV (81.2%)).DOSESuggestedPROCEDUREI-based product(GBq)MethodTHYROID THERAPYsodium iodide3.7-131I production(carcinoma) 8.325by 130Te (n, γ),FissiumTHYROID THERAPYsodium iodide 0.185-131I production(Graves)0.37by 130Te (n, γ),FissiumTHYROID THERAPYsodium iodide 0.925-131I production(hot nodule)11.063by 130Te (n, γ),Fissium TABLE 9Main Diagnosis Applications of 99mTc.DOSEPROCEDURE99mTc-BASED PRODUCT(Gbq)LYMPHO-antimony trisulfide0.0018-0.74SCINTIGRAPHYcolloid SPLEENdamaged RBC's0.185KIDNEYSdimercaptosuccinic0.185acid (DMSA)HEPATOBILIARYdisofenin (DISIDA)0.111-0.296BRAIN LESIONSDTPA0.555-0.925KIDNEYSDTPA0.37-0.555LUNG VENTILATION0.185BRAIN PERFUSIONECD0.555-0.925BRAIN LESIONSglucoheptonate0.555-0.925KIDNEYSglucoheptonate0.185-0.37HEPATOBILIARYHIDA0.111-0.296BRAIN PERFUSIONHMPAO0.555-0.925(BLOOD POOL)human serum albumin0.555-0.925(HSA)BONE IMAGINGhydroxymethylenedi-0.555-0.925phosphonate (HDP)ABSCESSleukocytes0.37-0.555VENOGRAMMAA0.185-0.37LUNG PERFUSIONmacroaggregated0.074-0.148albumin (MAA)HEPATOBILIARYmebrofenin0.111-0.296(CHOLETEC)KIDNEYSmercaptoacetyltri-0.185glycine (MAG3)BONE IMAGINGmethylenediphos-0.555-0.925phonate (MDP)SPLEENMIAA0.185-0.37BONE MARROWMIAA 510LIVERmicroaggregated0.185-0.37albumin (MIAA)GASTRIC EMPTYINGoatmeal (solid0.0011-0.0018phase)GASTRIC EMPTYINGovalbumin (solid0.0011-0.0018phase)BRAIN LESIONSpertechnetate0.555-0.925CYSTOGRAMpertechnetate0.444MECKEL'Spertechnetate0.37DIVERTICULUMPAROTIDSpertechnetate0.37THYROID SCANpertechnetate0.37TESTICLESpertechnetate0.555(Torsion)INFARCT (MYOCARD.)PYP0.555-0.925BONE IMAGINGpyrophosphate0.555-0.925(PYP)CARDIOVASCULARRBC's0.555-0.925HEMANGIOMARBC's0.555-0.925TESTICLESred cells0.925(Varicocele)GASTRIC EMPTYINGresin beads in0.0011-0.0018food (solid phase)(MYOCARDIUM)sestamibi0.555-0.925PARATHYROIDSsestamibi0.37BONE MARROWsulfur colloid0.185-0.37CYSTOGRAMsulfur colloid0.444GE REFLUXsulfur colloid0.0011-0.0018LIVERsulfur colloid0.185-0.37LYMPHO-sulfur colloid0.00185-0.74SCINTIGRAPHYSPLEENsulfur colloid0.185-0.37(MYOCARDIUM)teboroxime0.555-0.925 The main Therapy applications of Iodine compounds are listed in Table 8. Doses are much higher and the shortness of the. 128I will require correspondingly larger activities of the injected sample. Therefore, 131I produced by Te activation in general seems more appropriate. The dominant use of radio-isotopes in Medicine is presently concentrated on the use of 99mTc, as shown in Table 9. As already discussed, our activation method can produce large amounts of 98Mo activation, and therefore all these procedures can be in general performed with the proposed Activator. The activation method may be used to produce as well several other products. The activation reaction by neutron capture cannot be easily used to produce a variety of isotopes, amongst which 67Ga, 111In, 81Kr, 82Rb and 201Tl, and the short-lived positron emitters for PET scans, for which charged particle activation are preferable. The general availability of a particle accelerator could however foresee their production as well, but with conventional methods. The performance of the device is of course determined is by the choice of the accelerator. We assume two schematic configurations: (1) a “local” production of radio-isotopes within the premises of a Hospital, in which presumably the Accelerator is also used to produce PET isotopes by direct irradiation or other therapy programmes. The Activator is used to produce 128I and 99Mo (99mTc) The amount of 99mTc required for a single analysis is typically of the order of 1 Gbq. The simple extraction process from Molybdenum is performed near the Accelerator. The Accelerator is a compact cyclotron or a LINAC with 23 MeV protons, and the nominal current of 1 mA. The target is a thick, Beryllium target, water-cooled to absorb the beam-dissipated power (23 kWatt). The beam is spread over a surface of the order of a few square centimetres, to facilitate cooling. According to Table 4, the integrated yield is S0=1.66×1014 n/sec. The Activator has the geometry described in Table 6. With the help of an appropriate insertion tool, such as a pantograph tool, several different targets can be simultaneously inserted in the device. (2) a “regional”, industrial scale production of radio-isotopes, to be transported and used in the appropriate form at different Hospitals, located relatively near the activation plant. The transport time excludes the use of 128I, and 131I is to be used instead. We remark that for Thyroid therapy, rather than diagnosis, a large dose (up to 10 Gbq, see Table 8) must be given to the patient, and therefore the use of 131I has less counter-indications than in the case of diagnosis, where obviously the dose must be minimal and for which, as already pointed out, the use of 128I, is preferable. In addition, we have considered the production of 99Mo (99mTc) which can be transported in a Alumina dispenser, following the standard procedure used today. The amount of initial 99Mo activation required is of the order of 10 to 100 Gbq. In order to limit the mass of Molybdenum and hence the one of the Alumina in the transport, the activation density must be as large as possible. It is therefore assumed that a larger Accelerator is used and that neutrons are produced by the spallation process on Lead or eutectic Pb/Bi mixture. These complications are acceptable in view of the larger, “factory”-type scale of the operation and the larger amounts of radio-isotopes to be produced. The Accelerator is a compact cyclotron or a LINAC with 200 (150) MeV protons and the nominal current of 2.68 (5.35) mA, resulting in an integrated neutron yield, S0=3.0×1016 n/sec. The beam power to be dissipated in the molten metal target is 537 (802) kWatt. The Activator has the geometry described in Table 6, but with a significantly enlarged Buffer Layer to allow for the installation of the spallation Target. With the help of an appropriate insertion tool such as a pantograph tool, as in the previous case, several different targets can be inserted in the device. Since the fraction of the neutrons used for the activation is extremely small, many samples can be simultaneously irradiated in the Activator. The target is made either of isotopically enriched 98Mo or, if this is not available, of Natural Molybdenum containing 24.13% of 98Mo, in a chemical form discussed later on. The short-lived 99Mo (r1/2=65.94 h) is activated, in turn decaying into 99mTc. The Mo must be very pure. In particular, it must not contain Rhenium, which complicates the extraction of Molybdenum, since Rhenium has chemical properties similar to those of Technetium. In general, the presence of impurities may lead to unwanted radio-nuclides. The yield of 99Mo according to Table 3 and for a constant irradiation of 1 gram of 98Mo (4 g of Natural Mo) for a time t is 1.66×10−6×[1-exp(−t/95.35 h)]×S0 GBq, where S0 is the neutron yield of the source. For a continuous exposure of 100 hours, 1.07×10−6×S0 GBq/gr of 99Mo are activated. The extraction of Technetium (1 GBq of 99mTc corresponds to 5.13 ng of metal) out of Molybdenum matrix is a relatively simple process, vastly documented in the literature (see, for instance, A. K. Lavrukhina and A. A. Pozdnyakov, “Analytical Chemistry of Technetium, Promethium; Astatine and Francium”, Academy of Sciences of the USSR, Israel Program for Scientific Trenslations, Jerusalem 1969; and also R. D. Peacock, “The chemistry of Technetium and Rhenium” Elsevier Publishing Company, 1966). Though it is not part of the activation procedure, for completeness we briefly mention the separation on organic sorbents, especially Aluminium Oxide (Al2O3) which is widely used. An efficient process of extracting micro-amounts of 99mTc from irradiated Molybdenum has been discussed by Mixheev N. B., Garhy M. and Moustafa Z., Atompraxis, Vol 10 (264), 1964. These authors propose that Molybdenum be sorbed by Al2O3 as anion H4[P(Mo2O7)6]3−. The exchange capacity is about 8 gr/100 gr of Al2O3. According to this last method, the irradiated Molybdenum in the form of Sodium phosphomolybdate is converted into the complex salt K3H4[P(Mo2O7)6]nH2O by the reaction with KCl at pH 1.5 to 2.0. The precipitate is dissolved in 0.01 N HCl at 50° C. and the solution obtained is passed through a column filled with Al2O3 which has been washed by 0.1 N HCl. The phosphomolybdate colours the sorbent yellow. To elute the 99mTc, an isotonic NaCl solution is used. When 40 ml (figures refer to a 10.5 cm×0.5 cm column filled with 20 gr of Al2O3 ) of the elutent are passed, about 70 to 80% of the 99mTc is eluted from the column. The purity of the element is 99.9%. To elute the Molybdenum from the column, 10 to 20 ml of 0.1 N NaOH are used. The recovered Molybdenum can be re-injected in the Activator. Evidently, columns of different sizes can be used, depending on the specific activity required, and taking into account the exchange capacity. In order to limit to a minimum the handling of radioactive products, it is convenient to insert directly in the Activator the complex salt K3H4[P(Mo2O7)6]nH2O. In this way, after irradiation, the activated compound can be simply inserted in the 99mTc dispenser, without chemical handling. After the activity of the 99Mo has decayed below useful level, the salt is recovered (eluted) with 0.1 N NaOH, resulting in Sodium phospho-molybdate, which is regenerated with the above-mentioned reaction with KCl at pH 1.5 to 2, thus closing the cycle. Therefore, the target material can be reused indefinitely. TABLE 10Parameters of the Tc separator with Alumina (from MixheevN. B. et al, Atompraxis, Vol 10 (264), 1964)AluminaAl2O320grExchange capacityMo1.6grMo adsorbedMo160mgSolution0.01 KCl250mlColumn diameter0.5cmColumn length10.5cmChromogram strip1cmElutentNaCl40mlExtractingNaOH15ml An obvious drawback of using complex compounds in the Activator is the possible creation of spurious elements. The main radio-contaminants produced in the salt K3H4[P(Mo2O7)6]nH2O are 32P (δ=0.00968, τ1/2=14.26 d) and 42K(δ=0.0381, τ1/2=12.36 h), where δ is defined as the activity with respect to 99mTc in the sample after a long (asymptotic) irradiation and for a natural Molybdenum target. These small contaminants are not expected to be appreciably eluted in the 99mTc sample. If the highest purity is needed, obviously it would be best to use either metallic Molybdenum or oxide, MoO3. The compound can be in transformed into the complex salt after irradiation, using the previously described procedure to extract 99mTc or, alternatively, the extraction of 99mTc can be performed directly from the irradiated sample, for instance using an inorganic sorbent, such as Aluminium oxide as in the previous example. The procedures are described in W. D. Tucker, M. W. Green and A. P. Murrenhoff, Atompraxis, Vol 8 (163), 1962, for metallic Mo, and in K. E. Scheer and W. Maier-Borst, Nucl. Medicine Vol. 3 (214), 1964 for MoO3. In the alternative (1) of local production of 99mTc (point 2 in FIG. 11), the time delay between production and use is relatively short, but the activation is correspondingly smaller, because of the lower intensity and energy of the accelerator. Assuming indicatively a loss of activity of a factor 2 for handling delays, and a final sample of 1 Gbq, with the indicated irradiation of 100 h of a 23 MeV, 1 mA beam, we arrive at a sample of 98Mo of 11.26 g (46.6 g of Natural Mo). Elution of 99mTc from this sample will require 140 g (590 g) of Alumina, according to figures of Table 10. Though this column is probably too large for a portable dispenser, it is perfectly adequate for a fixed installation. The final solution of 99mTc can be easily concentrated before use, evaporating the excess water for instance under vacuum. The alternative (2) of a portable dispenser (point 3 in FIG. 11) is primarily characterised by a correspondingly smaller Alumina volume and hence a higher Mo activation. With the figures given above for the accelerator, and for an initial 99Mo activity of 50 GBq (the commercial Elutec™ Technetium Generator offers activation from 6 to 116 Gbq, calibrated on the 4th day after production), we find a sample of 98Mo of 1.56 g (6.4 g of Natural Mo), which will fit within the parameters of the Table 10. In view of the larger scale of the operation, it would be possible to irradiate a sample of MoO3, which is free of spurious activation and to transform the oxide into salt before introducing it into the Alumina dispenser. As before, the Mo could be recycled repetitively in the Activator, once the produced activation has sufficiently decayed, eluting it from the Alumina with the appropriate NaOH elutent. It has been verified that the activity of long-lived radio-nuclides, which could eventually accumulate in the sample is not appreciable. The short life of the 128I (τ1/2=24.99 m) precludes the transport, so that only the accelerator option (1) is retained (point 1 in FIG. 11). Fortunately, the resonance integral of 127I, is very large Ires=148 b, and therefore the activation is very efficient, even for relatively low neutron fluxes. Assuming an activation exposure of 30 min (½ of asymptotic activation), followed by a pause of 30 minutes before the imaging procedure (50% surviving), the activation is of 1.1 Gbq/gr, which is largely adequate. Different doses can easily be obtained by changing either the exposure time or the pause between exposure and use. Calculations have been performed also in the case of 127I activation. While the capture probabilities in the body of the Activator (Pb, Fe etc.) are, as expected, unchanged, the capture efficiency in 127I leading to 128I is η=2.62×10−5 g−1. The energy spectrum of the captured neutrons (solid line, left-hand ordinate scale) and the integrated capture probability (dotted line, right-hand ordinate scale) are shown in FIG. 12. Again, the resonant captures are dominant. As already pointed out, no chemical action is required, since the sample is already prepared in the appropriate form, and it can be immediately used, as required in view of the short half-life of 128I (τ1/2=24.9 m) Captures in the other elements of the compound must be taken into account. In particular, if Sodium Iodide (NaI) is used, the resonance integral for production of 24Na, a β-emitter (the decay is accompanied by two strong γ-lines (100%) at 1368.6 keV and 2754 keV) with a half-life of 14.95 hours is very small, Ires=0.26 compared with the value Ires=148 for Iodine. Calculations give capture efficiencies in NaI of η=1.62×10−7 g−1 for 24Na activation, and of η=2.218×10−5 g−1 for 128I activation, normalised for 1 gram of the NaI compound. The number of activated Na atoms are therefore more than two orders of magnitude less than the Iodine activation, with negligible consequences for the overall dose to the patient. Taking into account the ratio of lifetimes, the counting rate from 128I is enhanced by an additional factor 36. Therefore, the spurious effects in the measurements due to the presence of the 24Na are also negligible. Most likely it is so also for the other compounds of Table 7. We have considered the case of production of 131I (τ1/2=8.04 d), which is an isotope used widely in thyroid therapy. The activating reaction is neutron capture by 130Te which is a relatively abundant isotope of Tellurium (33.87%), but having a small resonance integral, Ires=0.26 b, with the following reactions: About 10% of captures lead to the isomeric state 131Te. The smallness of the resonance integral leads to a small capture probability. Fortunately, the Tellurium is a relatively cheap element (20$/lb), and it permits a simple extraction process for the Iodine produced. Therefore, relatively large amounts of target material can be used. The illustrative extraction method envisaged consists of a simple pyro-metallurgical process in which the ingot of activated element is melted to some 500° C. (melting point 449° C.), either in a crucible or by a simple electron beam device. The Iodine produced is volatised as an element, since the Tellurium Iodide (TeI4) decomposes at such temperatures. The evaporated Iodine is then easily condensed (melting point 113.5° C.), and thus recovered. This process may be repeated indefinitely, if the ingot is recast to the appropriate shape. Large amounts of 131I (τ1/2=8.04 d) are for instance used in therapy of Thyroid diseases. The activation process proceeds through the neutron capture of an isotope of natural Tellurium, 130Te (33.87%, Ires=0.259 b) . As already pointed out, the relatively small value of the cross-section requires relatively large amounts of target. Since the compound is relatively long-lived, it does not need to be produced locally. Therefore, we consider the accelerator option (2) (point 4 in FIG. 11), though sizeable amounts can also be produced with the conditions of option (1). We assume an exposure carried out during 12 days with a 10 kg target of natural Tellurium in metallic form, inserted in the form of 32 (cast) cylinders, each 50 cm long and of 0.56 cm radius (50 cm3). The remainder of the activator volume is filled with metallic Lead, in which the holes for the target have beer made. The resulting activated radio-nuclides are listed in Table 11. In addition to the two obvious isotopes 131Te and 131mTe which are the father nuclei of 131I, a number of Tellurium isotopes are produced due to the use of a natural Tellurium target. These activated products remain in the target material during the extraction process. Particularly strong is the decay of 127Te, though with a relatively short half-life of 9.35 hours. The target material will however remain activated for a relatively long time, due to the presence of 121mTe and 123mTe, with half-life of 154 days and 120 days, respectively. These residual activities may pile up in subsequent irradiations, but with no appreciable consequence. The extracted Iodine is essentially pure 131I, with a very small contamination of the short-lived 130I with a half-life of 12.36 hours, which will be rapidly further reduced by natural decay. In addition, there will be about 6 times as many nuclei of stable 127I produced and a negligibly small contamination of 129I (half-life 1.57×107 years). The tiny contamination of 131mXe will be easily separated during the Iodine extraction process. The last isotope in Table 11 is due to the short-lived activation of the Lead of the Activator volume and will not be extracted with the Target material. The total activity at discharge of the essentially pure 131I is 7355.42 Gbq (200 Cie). TABLE 11Radio-nuclides in the 10 kg natural Tellurium activator volume atthe end of a 12 days exposure. The accelerator is option (2).ElementDecay modeLifetime (1/e)Activity (GBq)Tellurium Radio-nuclides121Teε24.26d422.27121mTeIT(88.6%), ε222.7d12.04123mTeε173.1d1685.06125mTeIT83d34.64127Teβ−13.52h17892.73127mTeβ−157.6d495.35129Teβ−1.677h306.19129mTeIT(64%), β−48.59d477.30131Teβ−36.15m214.11131mTeIT(22%), β−1.808d951.12Iodine Radio-nuclides131Iβ−11.63d7355.42130Iβ−17.87h51.02Other Radio-nuclides131mXeIT17.21d28.02209Pbβ−4.704h121.23 As already described, the extraction procedure is performed by volatilising the Iodine content in the target, by melting the metal at about 500° C. In view of the high volatility of Iodine, the extraction should be essentially complete. Tellurium iodide (TeI4) formation is inhibited, since it decomposes at such temperatures. The Iodine is then condensed, while the contamination of Xenon (28.02 Gbq) is separated out and stored until it decays. The extraction process may take of the order of 4-6 hours. After extraction, the metal can be cast again into cylinders, ready for the next exposure. Allowing for a total preparation and handling time of the order of 3 days (surviving fraction 84%), the final sample of 131I will have a nominal activity of the order of 6150 GBq. Assuming instead accelerator option (1) and a 32 kg Tellurium target, the final production rate of 100 Gbq is obtained under the same procedure conditions as above. Only a very small fraction of the neutrons are captured in the Activator target. Therefore, if deemed necessary, it would be possible to increase considerably the yield by using a correspondingly larger mass of Tellurium target. The Interstitial Radiation therapy, known also as brachy-therapy, is the direct radioactive seed implant into the tumour. This technique allows the delivery of a highly concentrated and confined dose of radiation directly in the organ to be treated. Neighbouring organs are spared excessive radiation exposure. The radioactive source is usually a low-energy (20 to 30 keV) pure internal conversion (IC) γ-emitter. The lifetime should be long enough to ensure a large tissue dose, but short enough to permit the micro-capsule containing the radioactive product to remain inside the body permanently (capsules must be made of a material compatible with the body tissues). Typical sources used are 125I (τ1/2=60.14 d, <Eγ>=27 keV) and 103Pd (τ1/2=16.97 d, <Eγ>=20 keV). For 103Pd, the target can be metallic Rh irradiated with intermediate energy protons (≈20 MeV). The cross-section has a broad maximum of about 0.5 barn around 10 MeV. The yield of 103Pd at 23 MeV and thick target (0.75 g/cm2) is 5.20×10−4 for one incident proton, corresponding to an activation rate of 132.75 GBq/mA/day. However, the power dissipated in the target is large, 19.6 kWatt/mA. Therefore, if a maximum current of 200 μA is used (4 kwatt in the target), the production rate is the rather modest figure of 26.55 GBq/day (0.717 Cie/day), much smaller than the figures given here for 125I and neutron capture (≈600 Cie/day for scenario (2)). Accordingly, 103Pd may be better produced in the conventional way, with (p,n) reaction on 103Rh (the commercial product is known as Theraseed®-Pd103 and it is used in the therapy of cancer of the prostate). Production of 125I can be done with neutron capture of 124Xe and the reaction chain The resonance integral of 124Xe is very large Ires=2950 b, and an acceptable capture rate can be realised also with a gaseous target. The capture efficiency ηv=6.40×10−4/litre in pure 124Xe at n.p.t. In view of the small fraction of 124Xe in natural Xenon, (0.1%), isotopic separation is very beneficial in order to ensure a good, efficiency, also taking into account that the target can be used indefinitely. The calculated neutron spectrum and the capture energy distribution are shown in FIGS. 13a-b. Clearly, resonant capture dominates. One can also notice the flux depletion after the (strong) resonance crossing and the structure of the dip in the spectrum. If natural Xenon is directly activated, the capture efficiency leading to 125I is ηv=1.81×10−6/litre of Xe at n.p.t. The value is about a factor 3 larger than the one of pure 124Xe, once corrected for the fractional content (0.1%), since the self-shielding of the very strong resonances in 124Xe plays a more important role in the pure compound. The other isotopes in natural Xenon do not produce appreciable amounts of short-lived radioactive isotopes other than Xenon, and therefore do not contaminate the production of Iodine. Since the Xenon is an inert gas, the extraction of Iodine is immediate, because it condenses on the walls of the container. If natural Xenon is used, roughly the same amount of stable Cesium is produced, which is probably extracted with the Iodine. The Cesium is actually slightly contaminated with 137Cs which has a half-life of 30.1 years and a negligible activity. Such a contaminant is not present in the case of isotopically-enriched Xenon. In view of the large capture efficiency, the amount of activated 125I can be quite substantial. For instance, in the scenario (2) of the regional accelerator supplying 3.0×1016 n/sec, the production rate of 125I is of 6.0 Cie/day/litre of target with pure 124Xe at n.p.t. A 100 litre Activator at n.p.t will then produce as much as 600 Cie/day of 125I. A considerable number and variety of radio-isotopes are extracted from the fission fragments resulting from the fission of Uranium in a Reactor. The word “Fissium” is used herein to designate the group of elements which are the products of 235U fissions. The present Activator can be loaded with a small amount of Uranium, either natural or preferably enriched of 235U. Obviously, the target material can be recycled indefinitely. This material can be of the form of metallic Uranium or other compound, for instance Oxide, depending on the requirements of the subsequent extraction chemistry. In this way, practical amounts of Fissium can be produced, far away from criticality conditions and using initially a small sample. A possible scenario is briefly illustrated. We assume that the target is a small amount of Uranium enriched to 20% of 235U. The actual geometry used in the calculation was based on a finely subdivided metallic target arrangement for a total mass of about 30 kg. This mass has been chosen in order to ensure the correct representation of the resonance shielding, which is important in the case of Uranium. Typical capture efficiencies for truly infinitesimal amounts of Uranium are about a factor 2 larger than what is quoted in Table 13. The 20% enrichment is set by the requirements of the Non-Proliferation Agreement which limit to 20% the allowed enrichment in order to avoid the possibility of realising a critical mass. Incidentally, the amount of Plutonium which can be produced by this method is negligibly small. The target must be enclosed in a tight envelope to ensure that there is no leak of Fissium products during the exposure. The efficiencies for capture η and Fissium production (fission) ηf referred to 1 kg of enriched compound are listed in Table 13. Fissions produce additional neutrons which enter in the general neutron economy. The neutron fraction produced is about +1.04% for each kilogram of enriched Uranium, which is very small. Thus, even in the most extreme conditions, of target loading, the device remains vastly non-critical. Assuming that a specific element is present in the Fissium with an atomic fraction λ and that the exposure time texp and the necessary reprocessing time trep are both equal to one half-life of such compound, the initial activity for 1 kg of activated sample is given by 2.5×10−10 S0ληf (Gbq/kg). More generally, for arbitrary times, the activity of the extracted compound at the end of the reprocessing period is given by Equation [2]. In the scenario (2) of the regional accelerator supplying S0=3.0×1016 n/sec, the production rate for a compound with λ=0.04, texp=trep=τ1/2 and the parameters of Table 6, is 1150 GBq/kg (31.2 Cie/kg) of target. TABLE 12Most important Fissium production for 33 kg of 20% enrichedUranium, exposed for 10 days (scenario (1)).MassElement½ LifeGBq(arb. u.)77-AS1.62d2.2782.214E−783-BR2.40h1.6861.092E−888-KR2.84h23.521.911E−785-KR4.48h30.343.756E−783-KR1.83h6.2473.085E−891-SR9.63h832.42.372E−592-SR2.71h30.132.442E−790-SR28.78y1.411.040E−389-SR50.53d222.47.805E−493-Y10.18h978.23.011E−592-Y3.54h317.43.361E−691-Y58.51d234.59.743E−491-Y0.83h455.41.116E−697-ZR0.70d13307.089E−595-ZR64.02d244.61.161E−397-NB1.20h14335.431E−695-NB34.97d25.146.517E−595-NB3.61d1.7444.666E−799-MO2.75d18303.884E−499-TC6.01h17243.335E−5105-RU4.44h37.815.732E−7103-RU39.26d185.65.856E−4106-RU1.02y3.0389.389E−5105-RH1.47d303.13.659E−5103-RH0.93h185.35.804E−7112-PD0.88d6.4524.942E−7109-PD13.70h11.085.378E−7112-AG3.13h7.5178.568E−8111-AG7.45d4.0992.645E−6113-AG5.37h1.3972.756E−8115-CD2.23d5.5241.104E−6115-IN4.49h5.9619.999E−8125-SN9.64d4.1423.895E−6121-SN1.13d7.1617.625E−7128-SB9.01h4.6841.757E−7127-SB3.85d51.181.953E−5129-SB4.40h21.914.044E−7132-TE3.20d12794.223E−4131-TE1.25d112.71.440E−5129-TE1.16h27.331.330E−7129-TE33.60d7.3172.475E−5127-TE9.35h44.781.729E−6135-I6.57h529.71.528E−5133-I0.87d16761.508E−4132-I2.30h13191.299E−5131-I8.04d589.84.849E−4135-XE9.14h14225.708E−5133-XE5.24d16939.214E−4133-XE2.19d66.311.508E−5131-XE11.90d1.8522.253E−6137-CS30.10y1.4451.698E−3140-BA12.75d935.21.303E−3141-LA3.92h159.32.864E−6140-LA1.68d801.81.470E−4143-CE1.38d17332.663E−4144-CE0.78y47.21.511E−3141-CE32.50d416.71.490E−3143-PR13.57d782.11.185E−3145-PR5.98h2827.959E−6147-ND10.98d370.84.672E−4151-PM1.18d114.41.596E−5147-PM2.62y1.6511.814E−4149-PM2.21d340.28.753E−5156-SM9.40h3.4231.633E−7153-SM1.93d47.41.092E−5156-EU15.19d2.5424.702E−6157-EU15.18h1.5561.206E−7 TABLE 13Capture and Fissium production efficienciesfor 1 kg of 20% enriched UraniumFractionalCapture eff.Fissium eff.ElementContentη (kg−1)ηf (kg−1)235U0.201.212E−33.852E−3238U0.801.676E−36.587E−5 The most important radio-nuclides out of Fissium have been calculated with the geometry of Table 6 and are listed in Table 12. The conditions are the ones of scenario (1). Figures for scenario (2) are about two orders of magnitude larger. The exposure time has been arbitrarily set to 10 days, followed by 1 day of cool-down. The target was 20%-enriched metallic Uranium of a mass of 33 kg. Only elements with final activity larger than 1 Gbq are shown. It is interesting to compare the 99Mo production from Fissium with the one by direct activation from 98Mo (Paragraph 5.3). The asymptotic yield from 20%-enriched Uranium is calculated to be 51.3 Gbq/kg of target for scenario (1) activation. The same activation will be obtained with 288 grams of 98Mo. Therefore, we achieve comparable yields. Natural Silicon is made of the three isotopes 28Si (92.23%, Ires=0.0641 b), 29Si(4.46%, Ires=0.0543 b) and 30Si (3.1%, Ires=0697 b). The only isotope leading to an unstable element by neutron capture is the 30Si, which produces 31Si, in turn decaying with τ1/2=157 m to 331P, the only isotope of natural Phosphorus. The Montecarlo-calculated capture efficiencies of the isotopes for 1 kg of natural Si are η=2.353×10−4 kg−1 for 28Si, η=8.166×10−6 kg−1 for 29Si and η=1.733×10−5 kg−1 for the interesting isotope 30Si. Assuming scenario (2) of the regional accelerator with S0=3.0×1016 n/s, the atomic P implantation rate is 2.573×1014 s−1, corresponding to 1 p.p.b. (equivalent to an implanted density of donors of 5×1013 cm−3) implanted every 10.7 hours. No harmful isotope is apparently produced, and therefore the implantation process is “clean”, once the 30Si has decayed away. If higher implantation yields are needed, in view of the special, industrial nature of the process, a stronger accelerator (current and energy) may be used. A similar procedure can be applied to Germanium crystals. The leading captures occur in the 70Ge isotope (20%), producing the acceptor 71Ga (via 71Ge). A smaller rate of captures also occurs for 74Ge (36%), producing the donor 75As (via 75Ge). Hence, acceptor doping dominates. The waste transmuter operation is exemplified according to the previously-described scenarios, and in the framework of an EA. As already pointed out, these considerations apply easily also to the case where the “leaky” neutron source is a Fast Breeder reactor core. The General Layout of an EA operated in conjunction with the Waste transmuter is shown in simplified FIG. 14a (plane view at the medium plane of the Core), and FIG. 14b (vertical cut in the medium plane). It consists of a large, robust Steel Tank 20 filled with molten Lead 21, or with a Lead/Bismuth eutectic mixture. The heat produced is, dissipated by natural convection or with the help of pumps, through heat exchangers installed on the top (not shown in figure). The proton beam which is used to activate the nuclear cascades in the Energy Amplifier Core 22 is brought through an evacuated pipe 23, and it traverses the Beam Window 24 before interacting with the molten Lead in the Spallation Region 25. For simplicity, we display a common Lead volume for the Spallation Region and the rest of the device. This solution is perfectly acceptable, but it may be otherwise advisable to separate the circulation of the Lead of the Spallation Region from the one for rest of the unit. This alternative if, of course, of no relevance to the operation of the Transmuter. The Core, in analogy with standard practice in Reactors, comprises a large number of steel-cladded pins, inside which the Fuel is-inserted as Oxide, or possibly in metallic Form. The fuel material includes a fertile element, such as 232Th, which breeds a fissile element, such as 233U, after having absorbed a neutron. The subsequent fission of the fissile element exposed to the fast neutron flux in turn yields further neutrons. That breeding-and-fission process remains sub-critical (see WO 95/12203). The fuel pins, typically 1.3 m long, are uniformly spread inside a Fuel Assembly 26, also made out of Steel, generally of hexagonal shape, with typically 20 cm flat-to-flat distance. Each Fuel Assembly may contain several hundreds of pins. Molten Lead circulates upwards inside the Fuel Assemblies and cools effectively the Pins, removing the heat produced by the nuclear processes. The typical speed of the coolant is 1 m/s and the temperature rise of about 150 to 200° C. The high-energy neutrons Spallation neutrons from the Spallation Region drift into the core and initiate the multiplicative, sub-critical, breeding-and-fission process which is advantageously used (i) to Transmute Actinides in the core region and (ii) to produce the leaking neutrons used for the Waste transmutation in the Transmuter. The Transmuter Volume 27, 29 surrounds the core as closely as possible to make an effective use of the leaking neutrons. We have used for simplicity also for the Transmuter region the same hexagonal lattice 28 used for the Core. However, in order to reduce interactions in the supporting structures, these must be as light as possible. This is simplified by the light weight of the load to be transmuted (few hundred of kilograms). Though not a necessity, the same type of assemblies would permit to make use of the same tooling (pantograph) to extract both the fuel and Transmuter assemblies. The transmuter sections above and below the Core region 29 could be combined assemblies in which both Fuel and Transmuter are held together. A Buffer Region 30 should in principle be inserted between the Core and the Transmuter Volume. The Transmuter assemblies 28 are essentially filled with the circulating molten Lead, except the finely-distributed metallic 99Tc which can be in a variety of forms, for instance wires or sheets. Since 99Tc transforms itself into Ruthenium, which is also a metal, it may be left in direct contact with the molten Lead or enclosed in fine steel tubes, like the fuel. The engineering of the sample holder are of course to be defined according to the need and to the applications. In particular, different holders are required for Iodine, which is a vapour at the operating temperature of the EA (a chemical compound could be used instead, like for instance NaI which has higher melting point of 661° C. and a boiling point of 1304° C.), and it must be contained for instance in thin steel cladding. No appreciable heat is produced in the transmutation process, and it can be easily dissipated away by the molten Lead flow, even if its speed can be greatly reduced in the Transmuter sections. 99Tc, Iodine and/or Selenium holders can be combined in a single assembly, because the strong resonances of 99Tc occur at energies which are well below the ones of the other elements, as evidenced in FIG. 1. Since the resonance integral above, say, 50 eV is comparable for the three elements, captures occur first in 79Se and 129I and the surviving neutrons are later strongly absorbed by 99Tc. Therefore, one can imagine thin, sealed stainless tubes, similar to the fuel pins except that they contain 99Tc in dispersed form of metal wires or equivalent geometry and Iodine vapours at low pressure. Iodine transforms into Xenon which may be periodically purged, while Selenium produces Bromine and Krypton. The performance of the Waste Transmuter is exemplified in the case of the 99Tc. Other elements of Table 1 which have been selected for transmutation in the scenario described in Chapter 1 give quite similar behaviours. TABLE 14Neutron balance of illustrative EA.General parametersInitial fuel mixture(Th-TRU)O2Initial Fuel mass11.6tonThermal power output1.0GWattNominal Multiplic. coefficient, k0.98Initial TRU concentration21.07%Neutron capture (all reactions) inventoryCore83.5%Plenum & structures2.22%Main Vessel0.39%Leakage out of core (core fract.)14.3 (17.1)%Leakage out of tank1.46%Main reactionsCaptures64.5%Fissions (core fract.)31.5 (37.7)%n, Xn2.31%Others, incl. escapes1.65% We list in Table 14 the typical neutron balance of an EA operated as a TRU incinerator. The EA is initially filled with a mixture of Thorium and TRU's from the waste of a LWR, either in the form of Oxides (MOX) or of metals. Concentrations are adjusted in order to reach the wanted value of the multiplication coefficient k. It is a fortunate circumstance that an appropriate cancellation occurs between the increases of reactivity due to the 233U breeding from the Thorium and the losses of reactivity due to the emergence of FF's captures, reduction of the core active mass and diminishing stockpile of TRU's. Such an equilibrium permits to extend the burning to more than 100 GWatt day/t of fuel without external interventions and the simple adjustment of the produced power with the help of the Accelerator beam. In practice, this means 2 to 3 years of unperturbed operation. At the end of this cycle, the fuel is regenerated, by extracting the most neutron-capturing FF's and the Bred. Uranium and adding to the remaining Actinides an appropriate amount of LWR waste,in order to achieve the wanted value of k. The procedure is repeated indefinitely, until the LWR waste is exhausted. After a few cycles, an “asymptotic” mixture sets in, resultant of the equilibrium condition between the various reactions in the core. Such a mixture has excellent fission probability for fast neutrons, which ensures that the process can be continued in principle indefinitely. In order to evaluate the transmutation capacity of the Waste Transmuter, the transmutation volume 27 (FIGS. 14a-b) has, been filled with 270 kg of 99Tc in metallic form and finely dispersed in the Lead matrix, corresponding to a relative concentration of 1.04×10−3. The elements 29 of FIGS. 14a-b are left for spare capacity or transmutation of other elements. The mass of 99Tc to be eliminated referred to the TRU's in the waste from a standard LWR (see Paragraph 1.4) are in the ratio [99Tc/TRU]waste=(0.843 ton)/(10.178 ton)=0.0828. The calculated rate of transmutation for typical conditions of an EA (k=0.97) gives, for a fresh fuel load (first filling), [99Tc/TRU]transm=0.0856, i.e. sufficient to keep up with the waste composition. During the successive cycles of TRU's elimination, the rate of elimination is reduced, since the TRU's having the smallest fission cross-sections accumulate, so that more neutrons are required to achieve a successful fission. Instead, the 99Tc transmutation rate is essentially constant, since it is related to the fraction of neutrons which escape the core. Integrated over many cycles, as necessary to eliminate completely the TRU's, we find [99Tc/TRU]transm=0.1284, which is amply sufficient to eliminate both the 99Tc of the Waste and the one accumulated in the meantime because of the fissions of the TRU'S. The initial concentration of 99Tc has been chosen such as to match the needed performance. In order to see the dependence on this parameter, we have varied it over a wide interval. In FIG. 15, we display the transmutation rate as a function of the 99Tc concentration. As one can see progressive saturation occurs due to the self-shielding of the 99Tc in correspondence with the resonances. This is better evidenced in FIG. 16, where the neutron spectra, averaged over the transmutation volume are displayed for all the points of FIG. 15. A strong, growing depletion of the spectrum is observed after the two main 99Tc resonances. Note also the diffusive refill occurring after the last resonance and before thermal energies are reached. As already pointed out, this refill is due to the diffusion of neutrons from regions which contain no 99Tc. It should also be pointed out that the high energy spectrum, as apparent in FIG. 16, is not affected by the concentration of 99Tc. This shows that the operation of the main EA is little affected by the Activator parameters. That effect is further confirmed in FIG. 17, where the effective multiplication factor k is displayed, again as a function of the concentration. One can see that the k value is only very slightly affected, indicating that the operation of the EA is essentially independent on the activities in the Transmuter region. The fractional transmutation rate after 100 GWatt day/ton, which is a reasonable cycle time for the EA, is displayed in FIG. 18. As expected, small 99Tc loads are more quickly transmuted. In the concentration domain of interest, some 15-20% of the 99Tc are transmuted at the end of each cycle. This long transmutation time is of no practical concern, since the Transmuter elements can be left in place over several cycles, since the neutron flux is smaller and the radiation damage of the cladding correspondingly smaller. Finally, the fraction of the neutron leaked out of the vessel as a function of the 99Tc concentration is displayed in FIG. 19. The small dependence of this fraction with the concentration indicates the local nature of the resonance driven capture, which do not affect appreciably the neutron flux in the vicinity of the walls of the tank. Likewise, the neutron flux and spectrum at a reasonable distance from the Transmuter region are not very affected by the 99Tc captures. This means that the rest of the space around the core may be used to-transmute additional Waste. We have estimated the ultimate, practical transmutation capability to about twice the one already used to eliminate the 99Tc. This is amply sufficient to also eliminate all the unwanted elements according to Table 2. A general analysis of which kind of radio-nuclides could be produced with the neutron Activator has been performed. Target elements must be natural elements which are optionally selected with an isotopic enrichment, though costly. The neutron capture process leads to a daughter element which is unstable, with a reasonable lifetime, conservatively chosen to be between one minute and one year. In turn, the next daughter element can be either stable or unstable. If it is stable, the process is defined as “activation” of the sample. Since a second isotopic separation is unrealistic, the activated compound must be used directly. A practical example of this is the 128I activation from: a natural Iodine compound (127I→128I). If, instead, the first daughter element decays into another unstable (the same time window has been used) chemical species, which can be separated with an appropriate technique, the present method may constitute a way to produce pure, separated radio-nuclides for practical applications. As practical example, one may refer to the chain 98Mo→99Mo→99mTc. The suitability of a given production/decay chain to our proposed method depends on the size of the neutron capture cross-section. Two quantities are relevant: the resonance integral Ires, which is related to the use of a high A diffusing medium such as Lead, and the thermal capture cross-section which suggests the use of a low A diffuser such as Graphite. Another relevant parameter is the fractional content of the father nuclear species in the natural compound, which is relevant to the possible need of isotopic preparation of the target sample. Natur.Reson.Therm.Activatedhalf-lifeDecayDecayNexthalf-lifeTargetIsotopeConc.Integr.X-sectIsotopeactivatedmodeBr. R.Isotopenext Isot.NaNa- 231.000.260.607Na- 2414.96hβ−100.0MgMg- 260.11010.0160.0439Mg- 279.458mβ−100.0AlAl- 271.000.1120.244Al- 282.241mβ−100.0SiSi- 300.0310.6970.124Si- 312.622hβ−100.0PP - 311.000.07120.207P - 3214.26dβ−100.0SS - 340.04210.08350.256S - 3587.51dβ−100.0SS - 360.00020.100.167S - 375.050mβ−100.0ClCl- 370.24230.00250.Cl- 3837.24mβ−100.0ArAr- 360.00341.686.0Ar- 3735.04dβ+100.0ArAr- 400.9960.2310.756Ar- 411.822hβ−100.0KK - 410.06731.441.67K - 4212.36hβ−100.0CaCa- 440.02090.321.02Ca- 45163.8dβ−100.0CaCa- 460.000.2520.85Ca- 474.536dβ−100.0Sc- 473.345dCaCa- 480.00190.3791.26Ca- 498.715mβ−100.0Sc- 4957.20mScSc- 451.009.2431.10Sc- 4683.79dβ−100.0TiTi- 500.0540.06820.204Ti- 515.760mβ−100.0VV - 510.99752.085.62V - 523.750mβ−100.0CrCr- 500.04345.9418.20Cr- 5127.70dβ+100.0CrCr- 540.02370.1670.412Cr- 553.497mβ−100.0MnMn- 551.0010.5015.40Mn- 562.579hβ−100.0FeFe- 580.00281.361.32Fe- 5944.50dβ−100.0CoCo- 591.0072.042.70Co- 6010.47mβ−0.24CoCo- 591.0072.042.70Co- 6010.47mγ99.76NiNi- 640.00910.6271.74Ni- 652.517hβ−100.0CuCu- 630.69174.475.11Cu- 6412.70hβ+61.0CuCu- 630.69174.475.11Cu- 6412.70hβ−39.0CuCu- 650.30831.962.46Cu- 665.088mβ−100.0ZnZn- 640.4861.380.877Zn- 65244.3dβ+100.0ZnZn- 680.1882.891.15Zn- 6956.40mβ−100.0ZnZn- 680.1882.891.15Zn- 6913.76hγ99.97Zn- 6956.40mZnZn- 680.1882.891.15Zn- 6913.76hβ−0.03ZnZn- 700.0060.1170.105Zn- 712.450mβ−100.0ZnZn- 700.0060.1170.105Zn- 713.960hγ0.05Zn- 712.450mZnZn- 700.0060.1170.105Zn- 713.960hβ−99.95GaGa- 690.60118.02.52Ga- 7021.14mβ−99.59GaGa- 690.60118.02.52Ga- 7021.14mβ+0.41GaGa- 710.39931.804.26Ga- 7214.10hβ−100.0GeGe- 700.2052.233.35Ge- 7111.43hβ+100.0GeGe- 740.3650.4160.482Ge- 751.380hβ−100.0GeGe- 760.0781.310.172Ge- 7711.30hβ−100.0As- 771.618dAsAs- 751.0063.505.16As- 761.097dβ−99.98AsAs- 751.0063.505.16As- 761.097dβ+0.02SeSe- 740.009575.059.40Se- 75119.8dβ+100.0SeSe- 780.2364.700.492Se- 793.920mγ99.94SeSe- 780.2364.700.492Se- 793.920mβ−0.06SeSe- 800.4970.9280.699Se- 8118.45mβ−100.0SeSe- 800.4970.9280.699Se- 8157.28mγ99.95Se- 8118.45mSeSe- 800.4970.9280.699Se- 8157.28mβ−0.05SeSe- 820.0920.7950.0506Se- 8322.30mβ−100.0Br- 832.400hSeSe-820.0920.7950.0506Se- 831.168mβ−100.0Br- 832.400hBrBr- 790.5069128.012.60Br- 8017.68mβ+8.3BrBr- 790.5069128.012.60Br- 8017.68mβ−91.7BrBr- 790.5069128.012.60Br- 804.421hγ100.0Br- 8017.68mBrBr- 810.493146.403.09Br- 821.471dβ−100.0BrBr- 810.493146.403.09Br- 826.130mγ97.6Br- 821.471dBrBr- 810.493146.403.09Br- 826.130mβ−2.4KrKr- 780.003525.107.11Kr- 791.460dβ+100.0KrKr- 820.116225.032.20Kr- 831.830hγ100.0KrKr- 840.573.470.0952Kr- 854.480hβ−78.6KrKr- 840.573.470.0952Kr- 854.480hγ21.4KrKr- 860.1730.0230.34Kr- 871.272hβ−100.0RbRb- 850.72178.680.551Rb- 8618.63dβ+0.005RbRb- 850.72178.680.551Rb- 8618.63dβ−99.99RbRb- 850.72178.680.551Rb- 861.017mγ100.0Rb- 8618.63dRbRb- 870.27842.700.137Rb- 8817.78mβ−100.0SrSr- 840.005610.400.929Sr- 8564.84dβ+100.0SrSr- 840.005610.400.929Sr- 851.127hβ+13.4SrSr- 840.005610.400.929Sr- 851.127hγ86.6Sr- 8564.84dSrSr- 860.09864.701.19Sr- 872.803hγ99.7SrSr- 860.09864.701.19Sr- 872.803hβ+0.3SrSr- 880.82580.06280.66Sr- 8950.53dβ−99.991SrSr- 880.82580.06280.66Sr- 8950.53dβ−0.009YY - 891.000.8211.48Y - 902.671dβ−100.0YY - 891.000.8211.48Y - 903.190hγ100.0Y - 902.671dYY - 891.000.8211.48Y - 903.190hβ−0.002ZrZr- 940.17380.3160.057Zr- 9564.02dβ−98.89Nb- 9534.97dZrZr- 940.17380.3160.057Zr- 9564.02dβ−1.11Nb- 953.608dZrZr- 960.0285.860.0261Zr- 9716.90hβ−5.32Nb- 971.202hZrZr- 960.0285.860.0261Zr- 9716.90hβ−94.68NbNb- 931.009.781.32Nb- 946.263mγ99.5NbNb- 931.009.781.32Nb- 946.263mβ−0.5MoMo- 920.14840.9670.0237Mo- 936.850hγ99.88MoMo- 920.14840.9670.0237Mo- 936.850hβ+0.12MoMo- 980.24136.540.149Mo- 992.747dβ−12.5MoMo- 980.24136.540.149Mo- 992.747dβ−87.5Tc- 996.010hMoMo-1000.09633.880.228Mo-10114.61mβ−100.0Tc-10114.22mRuRu- 960.05527.260.332Ru- 972.900dβ+99.962RuRu- 960.05527.260.332Ru- 972.900dβ+0.038Tc- 9790.10dRuRu-1020.3164.171.41Ru-10339.26dβ−0.25RuRu-1020.3164.171.41Ru-10339.26dβ−99.75Rh-10356.11mRuRu-1040.1876.530.37Ru-1054.440hβ−72.0Rh-1051.473dRuRu-1040.1876.530.37Ru-1054.440hβ−28.0RhRh-1031.00928.0169.0Rh-1044.340mγ99.87RhRh-1031.00928.0169.0Rh-1044.340mβ−0.13PdPd-1020.010219.203.85Pd-10316.99dβ+0.1PdPd-1020.010219.203.85Pd-10316.99dβ+99.9Rh-10356.11mPdPd-1080.2646251.09.77Pd-10913.70hβ−0.05PdPd-1080.2646251.09.77Pd-10913.70hβ−99.95PdPd-1080.2646251.09.77Pd-1094.696mγ100.0Pd-10913.70hPdPd-1100.11722.790.261Pd-11123.40mβ−0.75Ag-1117.450dPdPd-1100.11722.790.261Pd-11123.40mβ−99.25Ag-1111.080mPdPd-1100.11722.790.261Pd-1115.500hγ73.0Pd-11123.40mPdPd-1100.11722.790.261Pd-1115.500hβ−7.5Ag-1117.450dPdPd-1100.11722.790.261Pd-1115.500hβ−19.5Ag-1111.080mAgAg-1070.5184100.44.20Ag-1082.370mβ−97.15AgAg-1070.5184100.44.20Ag-1082.370mβ+2.85AgAg-1090.48161460.104.0Ag-110249.8dγ1.36AgAg-1090.48161460.104.0Ag-110249.8dβ−98.64CdCd-1060.012510.601.11Cd-1076.500hβ+0.06CdCd-1060.012510.601.11Cd-1076.500hβ+99.94CdCd-1100.124938.2012.60Cd-11148.54mγ100.0CdCd-1140.287316.900.391Cd-1152.227dβ−0.0CdCd-1140.287316.900.391Cd-1152.227dβ−100.0In-1154.486hCdCd-1140.287316.900.391Cd-11544.60dβ−99.989CdCd-1140.287316.900.391Cd-11544.60dβ−0.011In-1154.486hCdCd-1160.07491.740.0859Cd-1172.490hβ−8.4In-11743.20mCdCd-1160.07491.740.0859Cd-1172.490hβ−91.6In-1171.937hCdCd-1160.07491.740.0859Cd-1173.360hβ−98.6In-11743.20mCdCd-1160.07491.740.0859Cd-1173.360hβ−1.4In-1171.937hInIn-1130.043322.013.90In-1141.198mβ−99.5InIn-1130.043322.013.90In-1141.198mβ+0.5InIn-1130.043322.013.90ln-11449.51dγ95.6In-1141.198mInIn-1130.043322.013.90In-11449.51dβ+4.4InIn-1150.9573110.232.0In-11654.41mβ−100.0SnSn-1120.009730.401.16Sn-113115.1dβ+0.0SnSn-1120.009730.401.16Sn-113115.1dβ+100.0In-1131.658hSnSn-1120.009730.401.16Sn-11321.40mγ91.1Sn-113115.1dSnSn-1120.009730.401.16Sn-11321.40mβ+8.9SnSn-1160.145312.400.147Sn-11713.60dγ100.0SnSn-1180.24225.320.25Sn-119293.1dγ100.0SnSn-1200.32591.210.16Sn-1211.127dβ−100.0SnSn-1220.04630.9160.21Sn-123129.2dβ−100.0SnSn-1220.04630.9160.21Sn-12340.06mβ−100.0SnSn-1240.05797.840.155Sn-1259.640dβ−100.0SnSn-1240.05797.840.155Sn-1259.520mβ−100.0SbSb-1210.573213.06.88Sb-1222.700dβ−97.6SbSb-1210.573213.06.88Sb-1222.700dβ+2.4SbSb-1210.573213.06.88Sb-1224.210mγ100.0Sb-1222.700dSbSb-1230.427122.04.80Sb-12460.20dβ−100.0SbSb-1230.427122.04.80Sb-1241.550mγ75.0Sb-12460.20dSbSb-1230.427122.04.80Sb-1241.550mβ−25.0SbSb-1230.427122.04.80Sb-124*20.20mγ100.0Sb-1241.550mTeTe-1200.00122.202.69Te-12116.78dβ+100.0TeTe-1200.00122.202.69Te-121154.0dγ88.6Te-12116.78dTeTe-120.00122.202.69Te-121154.0dβ+11.4TeTe-1220.02679.903.86Te-123119.7dγ100.0TeTe-1240.04825.137.79Te-12557.40dγ100.0TeTe-1260.18958.051.19Te-1279.350hβ−100.0TeTe-1260.18958.051.19Te-127109.0dγ97.6Te-1279.350hTeTe-1260.18958.051.19Te-127109.0dβ−2.4TeTe-1280.31691.730.247Te-1291.160hβ−100.0TeTe-1280.31691.730.247Te-12933.60dβ−36.0TeTe-1280.31691.730.247Te-12933.60dγ64.0Te-1291.160hTeTe-1300.3380.2590.31Te-13125.00mβ−100.0I -1318.040dTeTe-1300.3380.2590.31Te-1311.250dβ−77.8I -1318.040dTeTe-1300.3380.2590.31Te-1311.250dγ22.2Te-13125.00mII -1271.00148.07.09I-12824.99mβ+6.9II -1271.00148.07.09I-12824.99mβ−93.1XeXe-1240.0012950.190.Xe-12516.90hβ+100.0I -12559.41dXeXe-1260.000943.902.52Xe-12736.40dβ+100.0XeXe-1260.000943.902.52Xe-1271.153mγ100.0Xe-12736.40dXeXe-1280.019110.706.13Xe-1298.890dγ100.0XeXe-130.04115.3029.80Xe-13111.90dγ100.0XeXe-1320.2694.460.517Xe-1335.243dβ−100.0XeXe-1320.2694.460.517Xe-1332.190dγ100.0Xe-1335.243dXeXe-1340.1040.5910.303Xe-1359.140hβ−100.0XeXe-1340.1040.5910.303Xe-13515.29mγ100.0Xe-1359.140hXeXe-1340.1040.5910.303Xe-13515.29mβ−0.004XeXe-1360.0890.1160.299Xe-1373.818mβ−100.0CsCs-1331.00393.033.20Cs-1342.910hγ100.0BaBa-1300.0011176.013.0Ba-13111.80dβ+100.0Cs-1319.690dBaBa-1300.0011176.013.0Ba-13114.60mγ100.0Ba-13111.80dBaBa-1320.00130.408.06Ba-1331.621dβ+0.01BaBa-1320.00130.408.06Ba-1331.621dγ99.99BaBa-1340.024224.602.30Ba-1351.196dγ100.0BaBa-1360.07852.020.458Ba-1372.552mγ100.0BaBa-1380.7170.230.413Ba-1391.384hβ−100.0LaLa-1390.999110.5010.30La-1401.678dβ−100.0CeCe-1360.001964.307.18Ce-1379.000hβ+100.0CeCe-1360.001964.307.18Ce-1371.433dγ99.22Ce-1379.000hCeCe-1360.001964.307.18Ce-1371.433dβ+0.78CeCe-1380.00253.081.25Ce-139137.6dβ+100.0CeCe-1400.88480.2350.651Ce-14132.50dβ−100.0CeCe-1420.11080.8351.15Ce-1431.377dβ−100.0Pr-14313.57dPrPr-1411.0017.1013.20Pr-14219.12hβ−99.98PrPr-1411.0017.1013.20Pr-14219.12hβ+0.02PrPr-1411.0017.1013.20Pr-14214.60mγ100.0Pr-14219.12hNdNd-1460.17192.771.61Nd-14710.98dβ−100.0NdNd-1480.057614.502.85Nd-1491.720hβ−100.0Pm-1492.212dNdNd-1500.056415.801.38Nd-15112.44mβ−100.0Pm-1511.183dSmSm-1440.0311.751.88Sm-145340.0dβ+100.0SmSm-1520.2672740.236.0Sm-1531.928dβ−100.0SmSm-1540.22735.509.64Sm-15522.30mβ−100.0EuEu-1510.4781850.10700.Eu-1529.274hβ−72.0EuEu-1510.4781850.10700.Eu-1529.274hβ+28.0EuEu-1510.4781850.10700.Eu-1521.600hγ100.0EuEu-1530.5221390.359.0Eu-15446.30mγ100.0GdGd-1520.002898.01210.Gd-153241.6dβ+100.0GdGd-1580.248463.702.86Gd-15918.56hβ−100.0GdGd-1600.21867.800.874Gd-1613.660mβ−100.0Tb-1616.880dTbTb-1591.00469.031.70Tb-16072.30dβ−100.0DyDy-1560.0006953.037.90Dy-1578.140hβ+100.0DyDy-1580.001179.049.20Dy-159144.4dβ+100.0DyDy-1640.282174.02890.Dy-1652.334hβ−100.0DyDy-1640.282174.02890.Dy-1651.257mγ97.76Dy-1652.334hDyDy-1640.282174.02890.Dy-1651.257mβ−2.24HoHo-1651.00755.076.10Ho-1661.118dβ−100.0ErEr-1620.0014520.30.Er-1631.250hβ+100.0ErEr-1640.0161143.015.0Er-16510.36hβ+100.0ErEr-1680.26840.603.19Er-1699.400dβ−100.0ErEr-1700.14958.106.73Er-1717.516hβ−100.0TmTm-1691.001700.120.Tm-170128.6dβ+0.15TmTm-1691.001700.120.Tm-170128.6dβ−99.85YbYb-1680.0013378.02660.Yb-16932.03dβ+100.0YbYb-1740.31821.079.30Yb-1754.185dβ−100.0YbYb-1760.1276.643.28Yb-1771.911hβ−100.0Lu-1776.734dLuLu-1750.9741644.029.80Lu-1763.635hβ−99.91LuLu-1750.9741644.029.80Lu-1763.635hβ+0.1LuLu-1760.0259896.02810.Lu-1776.734dβ−100.0LuLu-1760.0259896.02810.Lu-177160.4dβ−78.3LuLu-1760.0259896.02810.Lu-177160.4dγ21.7Lu-1776.734dHfHf-1740.0016295.0463.0Hf-17570.00dβ+100.0HfHf-1760.0521613.616.20Hf-17751.40mγ100.0HfHf-1780.2731910.90.Hf-17925.10dγ100.0HfHf-1790.1363540.44.70Hf-1805.500hγ98.6HfHf-1790.1363540.44.70Hf-1805.500hβ−1.4Ta-1808.152hHfHf-1800.35134.4015.0Hf-18142.39dβ−100.0TaTa-1810.9999657.023.70Ta-182114.4dβ−100.0TaTa-1810.9999657.023.70Ta-182*15.84mγ100.0WW -1800.0013248.042.80W -181121.2dβ+100.0WW -1840.306716.101.95W -18575.10dβ−100.0WW -1840.306716.101.95W -1851.670mγ100.0W -18575.10dWW -1860.286344.043.30W -18723.72hβ−100.0ReRe-1850.3741710.129.0Re-1863.777dβ−93.1ReRe-1850.3741710.129.0Re-1863.777dβ+6.9ReRe-1870.626288.087.90Re-18816.98hβ−100.0ReRe-1870.626288.087.90Re-18818.60mγ100.0Re-18816.98hOsOs-1840.0002869.03430.Os-18593.60dβ+100.0OsOs-1880.133153.05.36Os-1895.800hγ100.0OsOs-1890.161837.028.90Os-1909.900mγ100.0OsOs-1900.26424.2015.0Os-19115.40dβ−100.0OsOs-1900.26424.2015.0Os-19113.10hγ100.0Os-19115.40dOsOs-1920.416.122.29Os-1931.271dβ−100.0IrIr-1910.3731170.1100.Ir-19273.83dβ−95.24IrIr-1910.3731170.1100.Ir-19273.83dβ+4.76IrIr-1910.3731170.1100.Ir-1921.450mγ99.98Ir-19273.83dIrIr-1910.3731170.1100.Ir-1921.450mβ−0.02IrIr-1930.6271310.128.0Ir-19419.15hβ−100.0IrIr-1930.6271310.128.0Ir-194171.0dβ−100.0PtPt-190.000186.70175.0Pt-1912.900dβ+100.0PtPt-1920.0079162.012.90Pt-1934.330dγ100.0PtPt-1940.3298.151.65Pt-1954.020dγ100.0PtPt-1960.2535.950.813Pt-19718.30hβ−100.0PtPt-1960.2535.950.813Pt-1971.590hβ−3.3PtPt-1960.2535.950.813Pt-1971.590hγ96.7Pt-19718.30hPtPt-1980.07252.704.34Pt-19930.80mβ−100.0Au-1993.139dAuAu-1971.001550.113.0Au-1982.693dβ−100.0AuAu-1971.001550.113.0Au-1982.300dγ100.0Au-1982.693dHgHg-1960.0014230.3520.Hg-1972.672dβ+100.0HgHg-1960.0014230.3520.Hg-19723.80hγ93.0Hg-1972.672dHgHg-1960.0014230.3520.Hg-19723.80hβ+7.0HgHg-1980.100274.802.28Hg-19942.60mγ100.0HgHg-2020.2982.655.68Hg-20346.61dβ−100.0HgHg-2040.06850.2560.492Hg-2055.200mβ−100.0TlTl-2050.70480.6480.119Tl-2064.199mβ−100.0TlTl-2050.70480.6480.119Tl-2063.740mγ100.0Tl-2064.199mPbPb-2080.5240.610.06Pb-2093.253hβ−100.0BiBi-2091.000.2020.0389Bi-2105.013dα0.0Tl-2064.199mBiBi-2091.000.2020.0389Bi-2105.013dβ−100.0Po-210138.4dThTh-2321.0083.508.49Th-23322.30mβ−100.0Pa-23326.97d
	abstract	A Method for X-ray wavelength measurement and an X-ray wavelength measurement apparatus capable of determining absolute wavelength easily and carrying out wavelength measurement having high precision with a simple structure are provided. The present invention is a Method for X-ray wavelength measurement carried out by using a channel-cut crystal for wavelength measurement (20) in which two opposing cut planes are formed and the lattice constant of which is known, and the method diffracts X-ray in respective arrangements (−, +) and (+, −) of the channel-cut crystal for wavelength measurement (20), to determine the absolute wavelength of the X-ray from the difference between crystal rotation angles in respective arrangements. This makes the alignment simpler, and, when only a channel-cut crystal suitable for measurement can be prepared, X-ray wavelength measurement can be carried out easily and with high precision.
	description	Referring now to the drawings, particularly to FIG. 1, there is illustrated a conventional boiling water/nuclear fuel bundle, generally designated 10, comprised of a plurality of fuel rods 12 disposed in a rectilinear array thereof and extending between upper and lower tie plates 14 and 16, respectively. As illustrated, a plurality of spacers S are axially spaced one from the other at various elevations along the fuel bundle. The fuel rods and spacers are enclosed within a fuel channel 18. As conventional, the boiling water/moderator flows upwardly through the fuel bundle about the fuel rods and through the cells of the spacers S, generating steam for the generation of power. Also illustrated in FIG. 1 are part-length fuel rods 20 which terminate at their upper ends short of the upper tie plate and typically below one or more of the upper spacers. Also illustrated in FIG. 1 are a plurality of conventional swirl vanes 22 disposed on the downstream side of the spacers at spacer cell locations above the part-length rods 20. Referring to FIG. 2, there is illustrated a representative spacer S which schematically illustrates cells in a rectilinear horizontal array of cells. For convenience, the rectilinear array is characterized by perpendicularly related rows and columns of cells 24, represented by x and y axes, respectively, in FIG. 2. In the illustrated spacer, a 10xc3x9710 array of spacer cells is shown with ten cells in each row (except for water rod locations, not shown). The cells may be of the ferrule type, i.e., short cylindrical tubes for receiving the fuel rods or may comprise other types such as disclosed in U.S. Pat. Nos. 5,740,218 and 5,727,039. Suffice to say that the present invention can be used with different and various configurations of cells. In FIG. 2, there is illustrated a plurality of flow diverters 26 in combination with vortex generators 28. Referring to FIG. 3, each flow diverter 26 comprises a cylindrical tube 30, preferably corresponding in diameter, in the illustrated form, to the diameter of the cell 24 (FIG. 2). The flow diverter 26 includes a plurality of tabs 32 which project laterally outwardly of the tube 30. The tabs 32 are also inclined relative to the axis of the tube 30 in an upstream direction when disposed over a cell 24. Thus, the tabs 32 are inclined downwardly, as illustrated in FIGS. 2 and 3 and form acute angles with the axis of the tubular cells 24. The lateral margins 34 of each tab 32 are shaped complementary to the arcuate outside surfaces of diagonally adjacent cells as illustrated in FIG. 5, while the distal ends 36 of the tabs 32 are likewise arcuately configured for engagement against a diagonally adjacent cell which lies both in a different row and column as the flow diverter. As best illustrated in FIG. 5, with this configuration, the tabs 32 may be interposed in the spaces between the orthogonally related rows and columns of spacer cells. To secure the flow diverter 26 to the spacer, the diverter may be welded to each of the adjacent cells in the corresponding column and row whereby the tabs 32 abut the adjacent cells. The weld joints are denoted 35 in FIGS. 7-9. Also as best seen in FIG. 5, the distal ends 36 of the tabs 32 extend laterally outwardly of the diverter tube 30 a distance beyond a straight line 40 drawn between center lines of diagonally adjacent cells. That is, the distal ends 36 of tabs 32 extend beyond straight lines 40 extending between the center axes 42 of diagonally adjacent cells and preferably abut the cell walls of a second set of diagonally adjacent cells in columns and rows not common to the column and row containing the flow diverter. With this configuration, it will therefore be appreciated that the flow through the spaces between the diagonally adjacent cells is completely diverted into the tube 30 of the diverter 26. Also illustrated in FIGS. 2-5 is a vortex generator 28 formed of a generally cylindrical tube 50 of a diameter for reception within the tube 30 of the flow diverter 26. The lower end or base of the tube 50 has laterally outward spring tabs 52 having lower surfaces 53 inclined to the axis through the tube 50. At like spacing about tube 30 of diverter 26 as the spring tabs 52 lie about tube 50, there are provided a plurality of slots 54 through diverter tube 30. By axially displacing the vortex generator 28 onto the flow diverter 26 as illustrated in sequence in FIGS. 8 and 9, the spring tabs 52 may be deflected first inwardly by engagement of the inclined surfaces 53 against the margin of tube 30 and then deflected outwardly in the slots 54 upon registration with the slots. The spring tabs/slot engagement thus maintain the vortex generator 28 assembled on the flow diverter 26 (FIG. 4). To generate a swirling flow through the flow diverter and vortex generator assembly, a plurality of swirler vanes 60 are provided adjacent the upper end of the vortex generator 28. The vanes are formed by slitting the generally cylindrical tube 50 in an axial direction through one end thereof and twisting inwardly one edge of each vane 60. For example, where four vanes 60 are used, eight slits 61 are formed through the end of tube 50. Tube material between alternate pairs of slits is removed and the remaining vanes are twisted so that one edge inclines inwardly to meet at its end the inner edges of other vanes. By forming the vanes in this manner, the vanes lie within the tubular envelope defined by tube 50, i.e., the vanes do not extend outside the periphery of tube 50 and lie within its peripheral confines. Preferably, the tip of each vane 60 along its inner edge contacts one another at an apex 62. The apices of the vane 60 may be welded at that location 62. Thus, it will be appreciated that the vanes impart a swirling motion to the boiling water/moderator flowing upwardly in the fuel bundle through the flow diverter 26 and into vortex generator 28, i.e., flowing in a downstream direction. Thus, a swirl-type flow above the assembly and in the void region above the part-length fuel rods is achieved. Note that by employing a suitable tool, not shown, the spring tabs 52 can be deflected inwardly to release the vortex generator from the flow diverter in situ and even after irradiation. To install the flow diverter/vortex generator, the flow diverter 26 is first applied over the upper or downstream end of a cell as illustrated in FIG. 6. Note that the tabs 32 extend between the diagonally adjacent cells and that their distal ends 36 are in contact with diagonally adjacent cells (FIG. 5). Welds are formed between the sides of the tubes 30 and the adjoining cells 24 in the rows and columns to secure the flow diverter to the spacer. The vortex generator 28 is then displaced axially on top of the flow diverter as illustrated in FIG. 8. Thus, as the lower end of tube 50 is received within tube 30, the springs 52 deflect inwardly and then, outwardly into slots 54 upon registration therewith to secure the vortex generator to the flow diverter as illustrated in FIG. 9. Turning now to the embodiment hereof illustrated in FIGS. 10-12, the flow diverter 80 is constructed similarly as the flow diverter 26 except that, instead of slots 54, threads 81 are formed about the tube 30, the tabs 32 being identical as those of the previous embodiment. Vortex generator 82 is likewise formed similarly as the vortex generator 28, except that spring tabs 52 are omitted in favor of complementary threads 84. The vortex generator 82 thus may be threaded into the flow diverter 80 and released therefrom by unthreading action in situ and after irradiation. The vanes 60 of vortex generator 82 are formed similarly as the vanes of the prior embodiment. Note also that the reaction force on the vortex generator 82 indicated by the arrow 86 resulting from swirling the flow in the direction illustrated by the arrow 88 in FIG. 11 tends to tighten the threads 84 of the vortex generator 82 about the threads 81 of the flow diverter 80. The threads 81 and 84 preferably comprise male/female threads, respectively, although the threads may be arranged vice-versa. Referring now to the embodiment hereof illustrated in FIGS. 13-15, there is illustrated a combined flow diverter/vortex generator and cell assembly 96 formed of a single integral piece of tubular sheet metal stock. For example, the tabs 100 of like configuration as the tabs 32 of the first embodiment are struck from the tubular body 98 to project laterally outwardly and at an angle to the tube axis as illustrated from a review of FIGS. 14 and 15. The tabs 100 have side and end configurations similarly as in the tabs 32 of the first embodiment, but leaving holes 102 in the tubular body 98 to receive the deflected flow generated by the tabs for flow in a downstream direction within the tubular body. The upper end of the tubular body 98 comprises the vortex generator 99 (FIG. 13) which includes vanes 60 constructed similarly as the vanes of the first embodiment. In this form, the lower tubular base of the assembly below the slots 102 forms a cell of the spacer disposed above a part-length fuel rod. That is, the tubular assembly 96 is substituted for a tubular cell 24 in the spacer S and forms an integral part therewith, i.e., by welding. Consequently, the boiling water/moderator flows in a downstream direction through the tubular body 98 and is joined by the deflected flow caused by the projecting tabs 100 for flow through the assembly and the vortex generator portion thereof. The vanes 60 impart the swirling motion to the flow, as represented by the circular arrow 108 in FIG. 13, similarly as in the first embodiment. FIG. 14 illustrates the cuts 106 formed in the tubular body 98 such that the tabs may be struck and laterally projected from the body as illustrated in FIG. 15. Referring to the final embodiment of the present invention illustrated in FIGS. 16 and 17, there is provided a flow diverter 120 for use in conjunction with a cell 24 which receives a fuel rod 12. The flow diverter 120 comprises a cylindrical or tubular body 122 similarly as the tube 30 of the first embodiment for securement on the top of a cell 24, also as in the embodiment of FIGS. 2-9. Thus, body 122 includes tabs 124 constructed similarly as tabs 32 and which project into the spaces between the cells. The boiling water/moderator is thus diverted from flowing between the cells into an annular space between the flow diverter 120 and a fuel rod 12 within the cell 24 thus increasing in flow velocity and pressure, e.g., xe2x80x9csuperchargingxe2x80x9d the flow. In FIG. 17, it will be appreciated that the base of the flow diverter includes a plurality of upstream extending projections 128 whereby the flow diverter may be secured, for example, by welding along the upper margin of a cell of the spacer. The cell flow diverter 120 may be located at any position in the spacer where it is desirable to increase the flow about the fuel rod extending through that cell location. Also, the tabs 124 can be cut to different lengths on opposite sides of one another, which when bent upwards would cause the coolant flow to swirl. Thus, as illustrated in FIG. 18, the flow diverter 150 has tabs 152 which have long and short sides or edges 154 and 156, respectively. When the tabs 152 are struck from the diverter 150, the tab twists such that flow through the diverter not only increases in velocity but is also swirled in the direction of the long sides 154 of the tabs 152. Thus, the tabs 152 are twisted to provide a swirl-type flow. 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.
052590089	claims	1. A depressurization system for a nuclear reactor having a reactor vessel disposed in a containment shell and inlet and outlet piping coupled to the reactor vessel defining a coolant circuit, the depressurization system comprising: a plurality of depressurizer valves coupled in fluid communication with the coolant circuit; at least one sparger in fluid communication with an inside of the containment shell, the sparger being coupled in fluid communication with at least one of said depressurizer valves; and, means for successively opening the depressurizer valves to effect depressurization, such that coupling between the coolant circuit and the containment shell is increased as pressure in the coolant circuit decreases. a coolant circuit coupled to the nuclear reactor via inlet piping and outlet piping, for circulating cooling water through the reactor, the reactor being disposed in the containment shell; a pressurizer tank disposed in the containment shell and having a bottom head coupled by a conduit to the outlet piping and a top head coupled to a vent conduit; a refueling water storage tank, open to an inside of the containment shell, and conduit means coupling the refueling water storage tank to the coolant circuit for adding water to the coolant circuit; a sparger defining at least one orifice submerged in the storage tank, coupled to the vent conduit of the pressurizer tank; a plurality of depressurization valves coupled in parallel with one another and in series with the vent conduit, the depressurization valves including a first valve operable to open at a first level setpoint and at least one additional valve operable to open at a second level setpoint which is lower than the first level setpoint; and, a last stage depressurization valve coupled between the outlet piping of the reactor and an inside of the containment shell, the last stage depressurization valve being operable to open at a third level setpoint lower than the first and second level setpoints, the last stage depressurization valve discharging cooling water directly from the coolant circuit into the containment shell. 2. The depressurization system according to claim 1, wherein the means for successively opening the depressurizer valves comprises valve controls responsive to a coolant level, operable to open successive ones of the depressurizer valves at progressively lower level setpoints. 3. The depressurization system according to claim 1, further comprising a pressurizer tank disposed in the containment shell, the pressurizer tank having a bottom head coupled to a coolant outlet of the reactor and a top head coupled to at least one of the depressurizer valves. 4. The depressurization system according to claim 1, wherein at least one of the depressurization valves is coupled to an inside of the containment shell through a sparger, said sparger being submerged in a tank of water and the tank of water being vented to the inside of the containment shell. 5. The depressurization system according to claim 4, wherein the tank of water is an in-containment refueling water storage tank, and further comprising means for emptying the tank of water into a sump in the containment shell housing the reactor. 6. The depressurization system according to claim 4, wherein a plurality of the depressurization valves are coupled to the inside of the containment shell through spargers, and a plurality of submerged orifices are vented to the inside of the containment shell. 7. The depressurization system according to claim 6, wherein said plurality of depressurization valves are coupled to the containment shell via conduits which are progressively larger for depressurization valves opening at progressively lower levels, for increasing coupling between the coolant circuit and the containment shell as the level in the coolant circuit decreases. 8. The depressurization system according to claim 7, further comprising means for opening at least one of the depressurization valves over a period of time, whereby a peak flow through said depressurization valves is limited. 9. The depressurization system according to claim 1, wherein at least one said depressurization valve is openable at a lowest pressure and is coupled directly between the coolant circuit and the containment shell. 10. The depressurization system according to claim 9, wherein the at least one depressurization valve openable at the lowest pressure is coupled between a coolant outlet of the reactor and the containment shell. 11. The depressurization system according to claim 4, wherein the plurality of depressurization valves includes a plurality of parallel depressurization valves coupled between the coolant circuit and at least one said sparger submerged in the tank, each of the plurality of parallel depressurization valves being coupled to a progressively larger conduit and openable at progressively lower level, wherein the tank of water is an in-containment refueling water storage tank, and further comprising means for emptying the tank of water into a sump in the containment shell housing the reactor. 12. The depressurization system according to claim 2, wherein the reactor includes a high pressure makeup tank coupled to the coolant circuit, and further comprising means for sensing a level of coolant in the high pressure makeup tank, coupled to the level responsive valve controls for successively opening the depressurizer valves. 13. The depressurization system according to claim 1, wherein at least one of the depressurization valves for effecting a final stage of depressurization is openable only below a predetermined differential pressure. 14. The depressurization system according to claim I, wherein the depressurization valves for at least two of the respective stages are openable using different forms of operators, whereby depressurization can proceed in the event of a common mode failure disabling one form of operator. 15. The depressurization system according to claim 1, further comprising a plurality of test valves coupleable in series with at least some of the depressurization valves, the test valves being coupled along conduits defining restricted flowpaths, whereby the depressurization valves can be tested under conditions characterized by one of limited pressure and limited flow, during operation and shutdown status of the reactor. 16. A nuclear reactor staged depressurization system for reducing operating pressure in a nuclear reactor disposed in a containment shell, comprising: 17. The nuclear reactor staged depressurization system according to claim 16, wherein the reactor is disposed in an open sump in the containment shell, and further comprising a valve coupling the open sump in the containment shell to the conduit means coupling the refueling water storage tank to the coolant circuit for adding water to the coolant circuit, whereby water in the containment shell can be added to the reactor from the open sump. 18. The depressurization system according to claim 17, wherein said plurality of depressurization valves are coupled to the containment shell via conduits which are progressively larger for depressurization valves opening at progressively lower levels, for increasing coupling between the coolant circuit and the containment shell as pressure in the coolant circuit decreases. 19. The depressurization system according to claim 16, further comprising a high pressure makeup tank coupled to the coolant circuit, and means for sensing a level of coolant in the high pressure makeup tank, said level setpoints being determined by said means for sensing. 20. The depressurization system according to claim 16, wherein at least one of the depressurization valves for effecting a final stage of depressurization is openable only below a predetermined differential pressure. 21. The depressurization system according to claim 16, wherein the depressurization valves for at least two of the respective stages are openable using different forms of operators, whereby depressurization can proceed in the event of a common mode failure disabling one form of operator. 22. The depressurization system according to claim 16, further comprising a plurality of test valves coupleable in series with at least some of the depressurization valves, the test valves being coupled along conduits defining restricted flowpaths, whereby the depressurization valves can be tested under conditions characterized by one of limited pressure and limited flow, during operation and shutdown status of the reactor.
	description	The present application is related to and claims the benefit of the earliest available effective filing date from the following listed application (the “Priority Application”) (e.g., claims benefits under 35 USC §119(e) for provisional patent applications, for any and all parent, grandparent, great-grandparent, etc. applications of the Priority Application). For purposes of the USPTO extra-statutory requirements, the present application claims benefit of priority of U.S. Provisional Patent Application No. 61/824,821, entitled Nuclear Fuel Assembly for Long Life, naming Jesse R. Cheatham, III; Michael E. Cohen; Christopher J. Johns; Brian C. Johnson; and Philip M. Schloss as inventors, filed May 17, 2013, which is within the twelve months preceding the filing date of the present application. All subject matter of the Priority application is incorporated herein by reference to the extent such subject matter is not inconsistent herewith. The present patent application relates to nuclear fission reactors and fuel assemblies, particularly for fast reactors, such as a traveling wave reactor. Fast reactors include a reactor vessel containing a reactor core. The reactor core includes a plurality of fuel assemblies. Liquid coolant passes through the reactor core, absorbing thermal energy from the nuclear fission reactions that take place in the reactor core. The coolant then passes to a heat exchanger and a steam generator, transferring the thermal energy to steam in order to drive a turbine that generates electricity. Fast reactors are designed to increase the utilization efficiency of uranium in fission reactions. Fast reactors can capture significantly more of the energy potentially available in natural uranium than typical light-water reactors. Production of energy in the fast reactor core is intense because of the high-energy neutrons that are employed. However, the high burnup and energy intensity in fast reactors also stresses the structural materials in the fuel assembly to a greater degree relative to light-water reactors. Fuel assemblies in fast nuclear fission reactor cores traditionally include a simple solid hexagonal tube surrounding a plurality of fuel elements, such as fuel pins. The tube directs coolant past the fuel pins, which are organized into a fuel bundle. The tube allows individual assembly orificing, provides structural support for the fuel bundle, and transmits handling loads from the handling socket to the inlet nozzle. Fuel pins are composed of nuclear fuel surrounded by cladding, which prevents radioactive material from entering the coolant stream. The coolant stream may be a liquid metal, such as liquid sodium. The hexagonal tubes degrade and deform from exposure to high temperatures (e.g., 300° C. to 700° C.), intensive radiation damage, and corrosion and other chemical interactions with the liquid metal coolant. Several phenomena, including irradiation creep, void swelling, bowing, and dilation, cause tubes to deform. The interstitial gap between adjacent tube walls closes during fuel assembly service life. For high burnup assemblies, the lifetime of the assembly is limited by mismatch between fuel pin swelling and dilation, which either allows coolant bypass around the periphery of the pin bundle, or reduction of coolant channels within the assembly due to compression of the pin bundle by the tube wall. Irradiation creep occurs as high-energy neutrons impinge on the tube and displace tube particles. Irradiation creep, duct dilation due to coolant pressure, and void swelling increase the diameter of the tube (i.e. cause expansion). Similarly, tubes may bow due to gradients in temperature, pressure, and radiation dose. Such gradients cause an imbalance in the macroscopic forces along the tube face. These problems, which warp and embrittle the tube structure, also increase the force necessary to withdraw fuel assemblies from the reactor, thus limiting the fuel assembly service life. Despite these deficiencies, hexagonal tubes continue to be used in fast reactors. Under service conditions in high-burnup fast reactors, such as breed-and-burn reactors (of which one type is a traveling wave reactor (“TWR”)), a simple hexagonal duct may not be able to withstand duct wall pressure differential, void swelling, and/or subsequent irradiation induced creep. This could result in unacceptable duct face dilation and duct bowing, thereby resulting in a fuel assembly design life which might not be able to support the high burnup required to achieve an equilibrium breed-and-burn cycle with depleted uranium feed assemblies. The typical approach in a core restraint system is to manage local duct dilation by adding interstitial space between ducts to allow room for duct face dilation to occur. Additionally, to manage duct bowing caused by void swelling, the core restraint system utilizes three load planes—the inlet nozzle, the above-core load pad, and the top load pad—to permit irradiation creep to offset the effects of swelling induced fuel assembly duct bowing, yet provide space for the duct faces to dilate outward toward the adjacent duct faces. Disclosed embodiments include nuclear fuel assembly ducts, nuclear fuel assemblies for nuclear reactors, nuclear reactors, methods for manufacturing ducts for nuclear fuel assemblies, and methods for loading a nuclear reactor. According to one embodiment, a duct for a nuclear fuel assembly includes a tubular body and an elongated member. The tubular body has a sidewall with an inner face and an outer face and is configured to contain nuclear fuel within a fuel region. The elongated member extends from the outer face along at least a portion of the fuel region and has a contact surface configured to stabilize the duct during operation of the nuclear fuel assembly. According to another embodiment, a fuel assembly for a nuclear reactor includes a tubular body, a plurality of fuel pins disposed within the tubular body, and an elongated member. The tubular body has a sidewall with an inner face and an outer face. A length of the tubular body containing the plurality of nuclear fuel pins defines a fuel region. The elongated member extends from the outer face along at least a portion of the fuel region and has a contact surface configured to strengthen the tubular body as the fuel assembly is operated. According to still another embodiment, a nuclear reactor includes a reactor vessel and a nuclear reactor core disposed in the reactor vessel. The nuclear reactor core includes a first fuel assembly, a second fuel assembly, a first elongated member, and a second elongated member. The first fuel assembly includes a first tubular body having a sidewall with an inner face and an outer face and is configured to contain nuclear fuel within a fuel region. The second fuel assembly includes a second tubular body configured to be positioned alongside the first tubular body and having a sidewall with an inner face and an outer face. The first elongated member extends from the outer face of the first tubular body along at least a portion of the fuel region and has a first contact surface. The second elongated member extends from the outer surface of the second tubular body. The second elongated member has a corresponding contact surface configured to engage the first contact surface to stabilize the first tubular body and the second tubular body during operation of the nuclear reactor. According to yet another embodiment, a method of manufacturing a duct for a nuclear fuel assembly includes providing a tubular body having a sidewall with an inner face and an outer face, the tubular body configured to contain nuclear fuel within a fuel region. The method also includes defining an elongated member on the outer face along at least a portion of the fuel region, the elongated member having a contact surface configured to stabilize the duct during operation of the nuclear fuel assembly. According to another embodiment, a method of loading a nuclear reactor includes positioning a first fuel assembly within a nuclear reactor core and positioning a second fuel assembly alongside the first fuel assembly. The first fuel assembly includes a first tubular body and a first elongated member. The first tubular body has a sidewall with an inner face and an outer face and is configured to contain nuclear fuel within a fuel region. The first elongated member extends from the outer face along at least a portion of the fuel region and has a first contact surface. The second fuel assembly includes a second tubular body and a second elongated member. The second tubular body has a sidewall with an inner face and an outer face. The second elongated member extends from the outer face. The second elongated member has a corresponding contact surface configured to engage the first contact surface to stabilize the first tubular body and the second tubular body during operation of the nuclear reactor. The foregoing is a summary and thus may contain simplifications, generalizations, inclusions, and/or omissions of detail; consequently, those skilled in the art will appreciate that the summary is illustrative only and is NOT intended to be in any way limiting. In addition to any illustrative aspects, embodiments, and features described above, further aspects, embodiments, and features will become apparent by reference to the drawings and the following detailed description. Other aspects, features, and advantages of the devices and/or processes and/or other subject matter described herein will become apparent in the teachings set forth herein. Introduction In the following detailed description, reference is made to the accompanying drawings, which form a part hereof. In the drawings, the use of similar or the same symbols in different drawings typically indicates similar or identical items, unless context dictates otherwise. The illustrative embodiments described in the detailed description, drawings, and claims are not meant to be limiting. Other embodiments may be utilized, and other changes may be made, without departing from the spirit or scope of the subject matter presented here. One skilled in the art will recognize that the herein described components (e.g., operations), devices, objects, and the discussion accompanying them are used as examples for the sake of conceptual clarity and that various configuration modifications are contemplated. Consequently, as used herein, the specific exemplars set forth and the accompanying discussion are intended to be representative of their more general classes. In general, use of any specific exemplar is intended to be representative of its class, and the non-inclusion of specific components (e.g., operations), devices, and objects should not be taken limiting. The present application uses formal outline headings for clarity of presentation. However, it is to be understood that the outline headings are for presentation purposes, and that different types of subject matter may be discussed throughout the application (e.g., device(s)/structure(s) may be described under process(es)/operations heading(s) and/or process(es)/operations may be discussed under structure(s)/process(es) headings; and/or descriptions of single topics may span two or more topic headings). Hence, the use of the formal outline headings is not intended to be in any way limiting. Overview Given by way of overview, illustrative embodiments include: nuclear fuel assemblies; ducts for nuclear fuel assemblies; nuclear fission reactor cores; nuclear fission reactors; methods of accommodating nuclear fuel assembly duct swelling; methods of fabricating a nuclear fuel assembly; and methods of loading fuel assemblies into a nuclear fission reactor core. Embodiments of this new duct design help reduce life-limiting constraints while simultaneously reducing the required duct structural material and interstitial sodium, which has a significant positive compound effect on the nuclear core design. The reduction in structural material and interstitial sodium reduces the burnup requirement in the discharged fuel assemblies, allows the core designer the option to reduce core height, and allows the core designer to add sodium coolant inside the ducts, which improves the thermal hydraulic performance of the core. This new duct design is relatively simple to manufacture, roughly the same order of magnitude of difficulty as a standard hexagonal duct. Embodiments of this fuel duct design use a simple concept to efficiently force the sodium coolant past the fuel pins in a highly uniform way throughout the lifetime of very high burnup fuel with minimal structural material, acceptable insertion and withdrawal loads, low manufacturing costs, excellent operational stability, very good service lifetime dimensional stability, negligible service induced degradation, and the ability to accommodate fuel pin swelling to mitigate typical fuel pin-to-bundle interaction problems. Embodiments of this duct design have external contact features or “elongated members” designed to constructively direct duct material void swelling to specific regions of the duct perimeter while effectively eliminating creep induced duct dilation. Embodiments of the elongated member duct are designed so that elongated members of neighboring ducts come into contact at operating temperature. The elongated member then becomes a support point, which prevents mid-face dilation. The elongated member features are designed to initiate duct-to-duct contact (operational stability) at hot standby conditions and provide full-length elongated member contact when full outlet temperature is attained. This allows the internal pressure within adjacent ducts to offset one another to effectively reduce the pressure load acting on the ducts to the difference between the two adjacent duct assemblies. The elongated member reduces the unsupported span of the duct face by more than one-half so the duct thickness can be greatly reduced. The use of the “elongated member” feature eliminates low power core instability that occurs in similar core designs such as the Fast Flux Test Facility (“FFTF”) and the Clinch River Breeder Reactor (“CRBR”) by providing duct-to-duct contact in the lower region of the fuel region prior to going critical. This problem was not solved in FFTF or CRBR and in the case of a large core, such as CRBR or TWRs, results in unavoidable core compaction during power ascension as the ducts settle in or the core “locks”. Embodiments of this device provide limited compliance when the pin bundle grows later in life in that the pin bundle can expand out into the regions adjacent to the “elongated members” and promote additional creep induce duct dilation locally to accommodate the pin bundle. Nuclear Fission Reactor Referring to FIGS. 1A-1C and FIG. 2 and given by way of non-limiting overview, an illustrative nuclear fission reactor 10 will be described by way of illustration and not of limitation. As shown in FIGS. 1A-1B, nuclear fission reactor 10 includes a nuclear fission reactor core 12 disposed in a reactor vessel 14. According to one embodiment, nuclear fission reactor core 12 contains nuclear fuel within a central core region 16. As shown in FIG. 2, nuclear fission reactor core 12 includes fissile nuclear fuel assemblies 18, fertile nuclear fuel assemblies 20, and movable reactivity control assemblies 22. In other embodiments, nuclear fission reactor core 12 includes only fissile nuclear fuel assemblies 18 and fertile nuclear fuel assemblies 20. According to the embodiment shown in FIGS. 1A-1C, an in-vessel handling system 26 is configured to shuffle ones of the fissile nuclear fuel assemblies 18 and ones of the fertile nuclear fuel assemblies 20. As shown in FIGS. 1A-1C, nuclear fission reactor 10 also includes a reactor coolant system 30. Still referring to FIGS. 1A-1C and FIG. 2, embodiments of the nuclear fission reactor 10 may be sized for any application as desired. For example, various embodiments of the nuclear fission reactor 10 may be used in low power (around 300 MWe-around 500 MWe) applications, medium power (around 500 MWe-around 1000 MWe) applications, and large power (around 1000 MWe and above) applications as desired. Embodiments of the nuclear fission reactor 10 are based on elements of liquid metal-cooled, fast reactor technology. For example, in various embodiments the reactor coolant system 30 includes a pool of liquid sodium disposed in the reactor vessel 14. In such cases, the nuclear fission reactor core 12 is submerged in the pool of sodium coolant in the reactor vessel 14. The reactor vessel 14 is surrounded by a containment vessel 32 that helps prevent loss of sodium coolant in the unlikely case of a leak from the reactor vessel 14. In various embodiments, the reactor coolant system 30 includes a reactor coolant pump, shown as pump 34. As shown in FIGS. 1A-1B, reactor coolant system 30 includes two pumps 34. Pumps 34 may be any suitable pump as desired (e.g., an electromechanical pump, an electromagnetic pump, etc.). Referring still to FIGS. 1A-1B, reactor coolant system 30 also includes heat exchangers 36. Heat exchangers 36 are disposed in the pool of liquid sodium. Heat exchangers 36 have non-radioactive intermediate sodium coolant on the other side of heat exchangers 36, according to one embodiment. To that end, heat exchangers 36 may be considered intermediate heat exchangers. According to one embodiment, steam generators are in thermal communication with the heat exchangers 36. It will be appreciated that any number of pumps 34, heat exchangers 36, and steam generators may be used as desired. The pumps 34 circulate primary sodium coolant through the nuclear fission reactor core 12. The pumped primary sodium coolant exits the nuclear fission reactor core 12 at a top of the nuclear fission reactor core 12 and passes through one side of the heat exchangers 36. According to one embodiment, heated intermediate sodium coolant is circulated via intermediate sodium loops 42 to the steam generators that, in turn, generate steam to drive turbines and electrical generators. According to other embodiments, heated intermediate sodium coolant is circulated to heat exchangers for still another use. The operation and construction of nuclear reactors is described by way of example and not of limitation in U.S. patent application Ser. No. 12/930,176, entitled Standing Wave Nuclear Fission Reactor and Methods, naming Charles E. Ahlfeld, Thomas M. Burke, Tyler S. Ellis, John Rogers Gilleland, Jonatan Hejzlari, Pavel Hejzlar, Roderick A. Hyde, David G. McAlees, Jon D. McWhirter, Ashok Odedra, Robert C. Petroski, Nicholas W. Touran, Joshua C. Walter, Kevan D. Weaver, Thomas Allan Weaver, Charles Whitmer, Lowell L. Wood, Jr., and George B. Zimmerman as inventors, filed Dec. 30, 2010, the contents of which are hereby incorporated by reference. Nuclear Fission Reactor Core and Fuel Assemblies Referring to the embodiment shown in FIGS. 3-4, a nuclear fission reactor core 100 (e.g., nuclear fission reactor core 12, etc.) includes a plurality of nuclear fuel assemblies (e.g., fissile nuclear fuel assemblies 18, fertile nuclear fuel assemblies 20, movable reactivity control assemblies 22, etc.), shown as fuel assemblies 110. As shown in FIG. 3, fuel assemblies 110 are supported in part by a core support grid plate 102. Primary sodium coolant flows through fuel assemblies 110, according to one embodiment. As shown in FIG. 4, fuel assembly 110 includes a plurality of nuclear fuel pins (e.g., fuel rods, fuel elements, etc.), shown as nuclear fuel pins 112, disposed within a duct that includes a tubular body, shown as tubular body 120. As shown in FIG. 4, tubular body 120 has a hexagonal cross-sectional shape. In other embodiments, tubular body 120 has another polygonal cross sectional shape (e.g., rectangular, pentagonal, etc.). In still other embodiments, tubular body 120 has still another cross sectional shape (e.g., circular, rounded, irregular, etc.). According to the embodiment shown in FIG. 4, tubular body 120 includes a plurality of sidewalls, shown as sidewalls 130, that extend along a longitudinal axis 122 of tubular body 120. In one embodiment, tubular body 120 is manufactured of HT-9 steel. Tubular body 120 has a very low void swelling rate relative to traditional ducts such that the combined effect of irradiation creep and void swelling is dominated by creep. In turn, the elastic spring forces in tubular body 120, which determine the magnitude of the force necessary to insert and withdraw fuel assemblies, will be relatively low. As shown in FIG. 4, in some embodiments, tubular body 120 is configured to contain nuclear fuel within a fuel region, shown as fuel region 140. Fuel region 140 extends between an upper blanket end and a lower blanket end, according to one embodiment. In various embodiments, nuclear fuel pins 112 include metal fuel or metal oxide fuel (regardless of whether the fuel is fissile fuel or fertile fuel). It will be appreciated that metal fuel offers high heavy metal loadings and excellent neutron economy, which is desirable for the breed-and-burn process in the nuclear fission reactor core 100. In one embodiment, fuel region 140 is defined by a length of tubular body 120 containing the plurality of nuclear fuel pins 112. In some embodiments, a nuclear fuel pin 112 is at least one of a control assembly (not shown) and a test assembly (not shown). The control assembly and the test assembly may include control instrumentation and test instrumentation, respectively, instead of or in addition to nuclear fuel. The control assembly and/or the test assembly may be structurally similar to nuclear fuel pins 112 that include nuclear fuel. In one embodiment, at least one of a control assembly and a test assembly is disposed along the length of tubular body 120 within fuel region 140. Accordingly, the term “nuclear fuel pin” is intended to include a control assembly and/or a test assembly. Referring still to FIG. 4, at least one of a duct and fuel assembly 110 includes an elongated member, shown as elongated member 150. According to the embodiment shown in FIG. 4, elongated member 150 extends from sidewall 130 along a portion of fuel region 140. In one embodiment, elongated member 150 includes a contact surface configured to stabilize tubular body 120 as fuel assembly 110 is operated (e.g., as part of a nuclear reactor core). As shown in FIG. 4, elongated member 150 extends along longitudinal axis 122 of tubular body 120. As shown in FIG. 4, elongated member 150 extends within a support region, shown as support region 152. Extension of elongated member 150 along longitudinal axis 122 facilitates the stabilization of tubular body 120 within support region 152 and reduces the risk of disengagement between neighboring elongated members 150. In one embodiment, support region 152 is defined along a length of fuel region 140 (or, for a control assembly or a test assembly, a length that corresponds to the length of fuel region 140). As shown in FIG. 4, each elongated member 150 extends along the entire length of support region 152. In one embodiment, elongated member 150 includes a plurality of plates. The plurality of plates may define a plurality of slots that extend across the longitudinal axis of tubular body 120. In one embodiment, the plurality of plates have a uniform width. In another embodiment, the plurality of plates have a uniform length. In still another embodiment, the plurality of plates have lengths that decrease from a first length at the upper blanket end to a second length at the lower blanket end. According to another embodiment, each elongated member 150 extends along a portion of support region 152. By way of example, a plurality of longitudinally spaced elongated members 150 may be positioned end-to-end and span support region 152. In one embodiment, the plurality of longitudinally spaced elongated members have lengths that decrease from a first length at the upper blanket end to a second length at the lower blanket end. In one embodiment, nuclear fission reactor core 100 employs an improved limited free bow core restraint system design. The limited free bow core design includes above-core or middle load pads and top load pads to induce an ‘S’ shape in fuel assemblies 110. Elongated members 150 of neighboring (e.g., nearby, adjacent, etc.) fuel assemblies 110 contact one another during operation of nuclear fission reactor core 100. Elongated members 150 offer contact points along outer faces of fuel assemblies 110, thereby providing spacing and support, managing swelling and dilation, and mitigating fuel assembly degradation and deformation. Elongated member 150 minimizes the friction forces associated with fuel assembly insertion and withdrawal. Elongated member 150 provides the dimensional stability and operational stability required to meet or exceed the demands of the high fuel burnup and irradiation fluence required in a fast reactor (e.g., a travelling wave reactor). Elongated member 150 produces a barrel-shaped expansion profile for nuclear fission reactor core 100 compared with previous designs that flowered outward above the top load pads. Such a barrel-shaped expansion profile reduces the uncertainty associated with traditional limited free bow systems. As shown in FIGS. 5-8B, sidewalls 130 of tubular body 120 each have an inner face 132 and an outer face 134. As shown in FIGS. 5 and 8A-8B, elongated member 150 extends from outer face 134 of sidewall 130. FIGS. 6-8A show cross sections of tubular body 120 along the lines 6-6, 7-7, and 8A-8A, respectively. As shown in FIG. 7, tubular body 120 has a first width W1 along the portion sectioned by line 7-7, and sidewalls 130 have a first thickness T1. In one embodiment, the portion of tubular body 120 sectioned by line 7-7 is a load pad having a thickness of four millimeters. As shown in FIG. 6, tubular body 120 has a second width W2 along the portion sectioned by line 6-6, and sidewalls 130 have a second thickness T2. As shown in FIG. 8a, tubular body 120 has a third width W3 along the portion sectioned by line 8A-8A, and sidewalls 130 have a third thickness T3. In one embodiment, the second thickness and the third thickness are both equal to two millimeters. The second width and the third width both equal to 156 millimeters, and the first width is equal to 160 millimeters, according to one embodiment. According to one embodiment, elongated member 150 has at least one of a specified width, a specified thickness, and a specified length. By way of example, the specified width, the specified thickness, and the specified length may facilitate the expansion of tubular body 120 according to a predetermined profile (e.g., a profile of expansion as a function of temperature, a profile of expansion as a function of time, etc.). By way of another example, the specified width, the specified thickness, and the specified length may facilitate coordinated expansion of tubular body 120 and nuclear fuel pins 112. As shown in FIG. 8B, elongated member 150 has a thickness Et and a width Ew. The width Ew of elongated member 150 is thirty millimeters, according to one embodiment. In one embodiment, the thickness Et of elongated member 150 is equal to two millimeters such that the overall width of tubular body 120 and elongated members 150 is equal to 160 millimeters. Referring next to the partial sectional view of FIG. 9, nuclear fission reactor core 100 includes a first nuclear fuel assembly, shown as first fuel assembly 200, a second nuclear fuel assembly, shown as second fuel assembly 230, and a third nuclear fuel assembly, shown as third fuel assembly 260. As shown in FIG. 9, first fuel assembly 200 includes a first tubular body 210 having a first sidewall 212 with an inner face 214 and an outer face 216. First fuel assembly 200 includes an elongated member, shown as first elongated member 220. First elongated member 220 extends from outer face 216 along at least a portion of the fuel region. According to the embodiment shown in FIG. 9, second fuel assembly 230 includes a second tubular body 240 having a second sidewall 242 with an inner face 244 and an outer face 246. Second fuel assembly 230 includes an elongated member, shown as second elongated member 250. Second elongated member 250 extends from outer face 246 along at least a portion of the fuel region. Referring still to FIG. 9, third fuel assembly 260 includes a third tubular body 270 having a sidewall 272 with an inner face 274 and an outer face 276. Third fuel assembly 260 includes an elongated member, shown as third elongated member 280. Third elongated member 280 extends from outer face 276 along at least a portion of the fuel region. As shown in FIG. 9, first tubular body 210, second tubular body 240, and third tubular body 270 are configured to contain nuclear fuel within a fuel region. As shown in the detail view of FIG. 10, first elongated member 220 and second elongated member 250 have a first contact surface 222 and a second contact surface 252, respectively. According to one embodiment, first contact surface 222 and second contact surface 252 are configured to stabilize first tubular body 210 and second tubular body 240 during operation of first fuel assembly 200 and second fuel assembly 230. As shown in FIG. 10, second contact surface 252 is a corresponding contact surface (i.e. a surface shaped to match first contact surface 222) configured to engage first contact surface 222 to stabilize first tubular body 210 during operation of the nuclear reactor. In some embodiments, first elongated member 220 and the second elongated member 250 have the same shape and contact one another along their respective lengths during operation of the nuclear reactor. According to one embodiment, the stability provided by engagement between the contact surfaces reduces the required duct structural material and interstitial sodium. These effects reduce the burnup requirement in the discharged fuel assemblies, allow the core designer the option to reduce core height, and allow the core designer to add sodium coolant inside the ducts, which improves the thermal hydraulic performance of the core. Operation of the Nuclear Reactor Referring next to FIGS. 11A-11C, operation of the nuclear reactor deforms first tubular body 210 and second tubular body 240. As shown in FIG. 11A, first tubular body 210 may be initially installed within the reactor core along second tubular body 240. In one embodiment, first tubular body 210 and second tubular body 240 are initially installed such that first contact surface 222 abuts second contact surface 252 before operation of the nuclear reactor. Accordingly, first elongated member 220 and second elongated member 250 space the tubular bodies prior to operation of the nuclear reactor. During an initial startup phase, first elongated member 220 and second elongated member 250 reduce the low power core instability of traditional reactor designs by providing duct-to-duct contact in the lower portion of the fuel region prior to the core going critical. As shown in FIG. 11A, first sidewall 212 and second sidewall 242 are flat plates prior to operation of the nuclear reactor. In other embodiments, first sidewall 212 and second sidewall 242 are initially otherwise shaped. In other embodiments, first contact surface 222 is spaced from second contact surface 252 upon installation. The space between first contact surface 222 and second contact surface 252 may be defined as the shortest distance between first contact surface 222 and second contact surface 252. In another embodiment, the distance is defined as the shortest distance between a planar surface formed by the majority of the points on the first contact surface 222 and a planar surface formed by the majority of the points on the second contact surface 252. In one such embodiment, the distance during initial installation between the first contact surface 222 and the second contact surface 252 is less than two millimeters. First tubular body 210 and second tubular body 240 may be positioned such that first contact surface 222 is spaced from second contact surface 252. In another embodiment, first elongated member 220 and second elongated member 250 are shaped such that first contact surface 222 is spaced from second contact surface 252. As the nuclear reactor operates, heat from the nuclear fission increases the temperature of the sodium coolant. Radioactive interactions due to the emission of high-energy neutrons lead to irradiation creep, void swelling, and duct dilation of first tubular body 210 and second tubular body 240. Pressure forces from the sodium coolant against inner face 214 and inner face 244 also deform first tubular body 210 and second tubular body 240. Such radioactive interactions and pressure forces bulge the sidewalls of traditional ducts for nuclear fuel assemblies and cause duct bowing and duct twist. Withdrawal loads are applied to a first end (e.g., the handling socket) and carried by the tubular body to a second end (e.g., the inlet nozzle) of the nuclear fuel assembly. Duct dilation, duct bowing, axial growth, and withdrawal forces limit the performance of traditional fuel assemblies. Duct dilation reduces the interstitial gaps (i.e. the spaces between tubular bodies) during service life and leads to large withdrawal loads. According to one embodiment, elongated members extending from adjacent tubular bodies engage one another to stabilize the tubular bodies during operation of the nuclear reactor. Such stabilization reduces dilation, thereby improving the service life of the fuel assemblies. Engagement of the elongated members reduces the likelihood of core compaction and large reactivity step changes. Engagement of the elongated members isolates the life of the fuel assembly from the duct-to-duct gap, which collapses during operation of traditional duct assemblies and limits the life of the fuel assembly. In some embodiments, elongated members reduce the long-term degradation of the duct-to-duct gap in the limited free bow design, thereby improving the service lifetime of the fuel assembly. Elongated members mitigate local duct face dilation by reducing the peak bending stresses to near (or below) the irradiation creep activation stress levels. The majority of forces and sidewall bending stresses are eliminated by cancelation of the internal duct pressure forces (i.e. the dilation forces due to the pressure of the coolant). Duct pressure forces are canceled because outward forces from one duct are opposed by forces from a neighboring duct, which are transferred through the elongated members. The difference between duct pressure forces of neighboring ducts is relatively small due to orificing, according to one embodiment. Such load conditions facilitate the reduction in the amount of structural steel for the tubular body and interstitial sodium, thereby increasing the power density of the fuel assembly. As shown in FIG. 11B, radioactive interactions and pressure forces bulge first sidewall 212 and second sidewall 242. According to one embodiment, first tubular body 210 and second tubular body 240 do not significantly deform over the life of the nuclear reactor, thereby reducing the risk of core compaction and of decreased coolant flow through the core. Engagement of first contact surface 222 with second contact surface 252 stabilizes first tubular body 210 and second tubular body 240 by reducing the potential deformation of first sidewall 212 and second sidewall 242. In one embodiment, first elongated member 220 interlocks second elongated member 250 when first contact surface 222 engages second contact surface 252. The interlocking may further reduce slipping between first contact surface 222 and second contact surface 252, thereby limiting movement of first tubular body 210 and second tubular body 240. As shown in FIG. 11B, volumetric void swelling, the effect of void swelling in the duct wall, is reduced to inconsequential bulges (e.g., bulges up to ˜600 displacements per atom) in first sidewall 212 and second sidewall 242 adjacent the elongated members, and the resulting spring force acting on adjacent tubular body is limited by creep, which dominates the void swelling induced strains in the bulged regions. As shown in FIGS. 11A-11B, the vertices of first sidewalls 212 and second sidewalls 242 are disposed closer to one another during operation of the nuclear reactor. Accordingly, the elongated members constructively direct material void swelling to specific regions of the perimeter of first tubular body 210 and second tubular body 240, while effectively eliminating creep induced duct dilation. According to one embodiment, the tubular bodies manage the fuel pins of the nuclear fuel assemblies. As the nuclear reactor core operates, the pin bundle expands, potentially contributing to dilation of the tubular bodies. Due to engagement between the elongated members of neighboring fuel assemblies, the fuel pins may expand out into the regions adjacent to the elongated members. Such expansion promotes additional irradiation creep-induced dilation locally and accommodates swelling of the pin bundle. Accordingly, elongated members extending from the tubular body alleviate mismatch between fuel pin swelling and duct dilation. In one embodiment, a first fuel assembly includes a first load pad and a second fuel assembly includes a second load pad. First tubular body 210 and second tubular body 240 may be initially installed such that first contact surface 222 is spaced from second contact surface 252 (e.g., a corresponding contact surface) and the first load pad is spaced from the second load pad. According to one embodiment, first tubular body 210 and second tubular body 240 may be positioned such that first contact surface 222 abuts second contact surface 252 at a first temperature and the first load pad abuts the second load pad at a second temperature. In one embodiment, the first the first temperature is lower than the second temperature. Such lower-temperature engagement of first contact surface 222 with second contact surface 252 causes stable radial expansion of the nuclear reactor core. In another embodiment, the first contact surface 222 abuts second contact surface 252 at the initial loading temperature and maintains engagement during operation of the nuclear reactor core. Such engagement may reduce the uncertainty in predicting core behavior during operation and facilitate operating the nuclear reactor core at higher powers, temperatures, and additional reactivity relative to traditional fast reactors. Referring next to FIG. 11C, first tubular body 210 and second tubular body 240 contract during refueling or shuffling conditions (e.g., 320 degrees Celsius, a point where high temperature coolant no longer interacts with inner face 214 and inner surface 234). Such contraction causes first contact surface 222 and second contact surface 252 to retract from an engaged position shown in FIG. 11B. Retraction of first contact surface 222 and second contact surface 252 reduces withdrawal and insertion forces. Elongated Member Configurations Referring to FIGS. 12A-12B, second elongated member 250 defines a recess 254. As shown in FIGS. 12A-12B, recess 254 extends inward (i.e. toward a central axis of the nuclear fuel assembly) from second contact surface 252. Recess 254 reduces the surface area of second contact surface 252, thereby reducing the risk of bonding between first contact surface 222 and second contact surface 252 during operation of the nuclear reactor core. Recess 254 also reduces the mass of second elongated member 250, thereby improving the power density of the nuclear fuel assembly. In one embodiment, both first elongated member 220 and second elongated member 250 define a recess. In another embodiment, at least one of first elongated member 220 and second elongated member 250 define a plurality of recesses. The plurality of recesses may be defined according to a specified pattern (e.g., a rectangular array, a triangular array, a polar array, a spherical array, etc.). As shown in FIGS. 12A-12B, recess 254 is a blind hole. In other embodiments, recess 254 is a hole that extends through second elongated member 250. According to the embodiment shown in FIG. 12A, recess 254 is circular. In other embodiments, recess 254 is otherwise shaped (e.g., hexagonal, rectangular, a slot extending along the length of second elongated member 250, a slot extending at least partially across the length of second elongated member 250, etc.). According to another embodiment, second elongated member 250 includes a projection. The projection may extend outward from second contact surface 252. In one embodiment, the projection includes a pin. In another embodiment, the projection includes a strip extending across the length of second elongated member 250. Second elongated member 250 may include a single projection or may include a plurality of projections, according to various embodiments. Referring next to FIGS. 13A-15D, first elongated member 220 interlocks second elongated member 250. Interlocking further reduces slipping between first contact surface 222 and second contact surface 252, thereby limiting movement of first tubular body 210 and second tubular body 240. In one embodiment, first elongated member 220 interlocks second elongated member 250 during operation of the nuclear reactor as first contact surface 222 engages second contact surface 252. As shown in FIGS. 13A-15D, first elongated member 220 interlocks second elongated member 250 during operation of the nuclear reactor before first contact surface 222 engages second contact surface 252. According to one embodiment, second elongated member 250 includes a recess, and first elongated member 220 includes an interlocking projection. In one embodiment, the recess and the interlocking projection join first tubular body 210 and second tubular body 240 during operation of the nuclear reactor. As shown in FIGS. 13A-13B, second elongated member 250 includes a plurality of recesses 254 extending inwardly from first contact surface 222, and first elongated member 220 includes a plurality of interlocking projections, shown as interlocking projections 224. As shown in FIGS. 13A-13B, interlocking projections 224 extend outward (i.e. away from a central axis of the nuclear fuel assembly) from first contact surface 222. In one embodiment, ends of interlocking projections 224 are spaced from second contact surface 252 during initial installation of the nuclear fuel assemblies. As shown in FIG. 13B, interlocking projections 224 extend into corresponding recesses 254 during operation of the nuclear reactor, as first tubular body 210 and second tubular body 240 expand and first elongated member 220 and second elongated member 250 are drawn together. Twisting of first tubular body 210 or second tubular body 240 (e.g., due to unequal thermal expansion along the length of the fuel assembly) reduces the life of the fuel assembly. In one embodiment, sidewalls of interlocking projections 224 contact inner surfaces of recesses 254, thereby providing a load path that reduces the likelihood of relative movement (e.g., slipping) between first tubular body 210 and second tubular body 240. As shown in FIG. 13B, first elongated member 220 fully interlocks the second elongated member 250 when the first contact surface 222 engages the second contact surface 252. As shown in FIG. 13A, interlocking projections 224 are pins. According to the embodiment shown in FIG. 13A, the pins are cylindrical, and recesses 254 are circular blind holes. In other embodiments, interlocking projections 224 and recesses 254 are otherwise shaped (e.g., hexagonal, rectangular, etc.). In still other embodiments, recesses 254 are holes that extend through second elongated member 250. Referring next to FIGS. 14A-14B, interlocking projection 224 includes a strip extending along a length of the first elongated member 220. As shown in FIG. 14A, recess 254 is a slot extending along the length of second elongated member 250. As shown in FIG. 14B, the strip extends into the slot during operation of the nuclear reactor. In one embodiment, sidewalls of interlocking projection 224 contact inner surfaces of recess 254, thereby providing a load path that reduces the likelihood of relative movement (e.g., slipping) between first tubular body 210 and second tubular body 240. As shown in FIGS. 14A-14B, the strip has a rectangular cross-sectional shape. According to other embodiments, the strip has a trapezoidal cross-sectional shape or still another shape. While shown in FIGS. 14A-14B as extending along the length of first elongated member 220 and second elongated member 250, the strip and the slot may extend at least partially across the length of first elongated member 220 and second elongated member 250. Referring to FIGS. 15A-15D, first elongated member 220 defines a plurality of interlocking projections 224, and second elongated member 250 defines a plurality of recesses 254. As shown in FIGS. 15A-15D, the plurality of interlocking projections 224 extend from first contact surface 222, and the plurality of recesses 254 extend inward from second contact surface 252. In one embodiment, the plurality of recesses 254 and the plurality of interlocking projections 224 join the first tubular body and the second tubular body at a plurality of sites during operation of the nuclear reactor. In some embodiments, the plurality of recesses 254 and the plurality of interlocking projections 224 are defined according to a specified pattern. As shown in FIG. 15A, the plurality of recesses 254 and the plurality of interlocking projections 224 are defined in a rectangular three by two array. In other embodiments, the plurality of recesses 254 and the plurality of interlocking projections 224 are defined in a rectangular array having different dimensions. As shown in FIG. 15B, the plurality of recesses 254 and the plurality of interlocking projections 224 are defined in a triangular array. As shown in FIG. 15C, the plurality of recesses 254 and the plurality of interlocking projections 224 are defined in a polar array. As shown in FIG. 15D, the plurality of recesses 254 and the plurality of interlocking projections 224 are defined in a circular array. Referring next to FIG. 16, first elongated member 220 and second elongated member 250 define a plurality of interlocking teeth. As shown in FIG. 16, first elongated member 220 defines a first set of teeth, shown as teeth 226, and second elongated member 250 defines a second set of teeth, shown as teeth 256. According to one embodiment, teeth 226 engage teeth 256 to stabilize first tubular body 210 and second tubular body 242 during operation of the nuclear reactor. By way of example, a surface of teeth 226 may contact a surface of teeth 256, thereby providing an interface that carries loading, limits twist of first tubular body 210 and second tubular body 242, and prevents relative movement of first elongated member 220 and second elongated member 250. As shown in FIG. 16, teeth 226 and teeth 256 include tooth profiles that extend across the length of first sidewall 212 and second sidewall 242. In other embodiments, teeth 226 and teeth 256 include tooth profiles that extend along the length of first sidewall 212 and second sidewall 242 (i.e. the teeth may be longitudinal or extend laterally across the sidewalls). Referring next to the embodiment shown in FIGS. 17A-17C, first contact surface 222 and second contact surface 252 have mating shapes. As shown in FIG. 17B, first contact surface 222 is concave. As shown in FIG. 17C, second contact surface 252 is convex. First contact surface 222 engages second contact surface 252 to stabilize first tubular body 210 and second tubular body 240 during operation of the nuclear reactor. As shown in FIG. 17A, first elongated member 220 has a first cross-sectional shape within a plane 300 that is perpendicular to outer face 216 of first tubular body 210. As shown in FIG. 17B, plane 300 extends across the lengths of first tubular body 210 and second tubular body 240. Second elongated member 250 has a second cross-sectional shape within plane 300. In one embodiment, the first cross-sectional shape is a negative of the second cross-sectional shape to facilitate engagement between first contact surface 222 and second contact surface 252. According to another embodiment, first elongated member 220 has a first cross-sectional shape within a plane 310 that is perpendicular to outer face 216 of first tubular body 210. As shown in FIG. 17B, plane 310 extends along the lengths of first tubular body 210 and second tubular body 240. Second elongated member 250 has a second cross-sectional shape within plane 310. In one embodiment, the first cross-sectional shape is a negative of the second cross-sectional shape to facilitate engagement between first contact surface 222 and second contact surface 252. According to one embodiment, first elongated member 220 and second elongated member 250 have specified cross-sectional shapes within a plane orthogonal to longitudinal axes along which first tubular body 210 and second tubular body 240 extend. As shown in FIGS. 17B-17C, first elongated member 220 is plano-concave and second elongated member 250 is plano-convex. In other embodiments, at least one of first elongated member 220 and second elongated member 250 have cross-sectional shapes that are rectangular, trapezoidal, plano-convex, plano-concave, bi-convex, or bi-concave. First contact surface 222 and second contact surface 252 may be planar, substantially flat, arcuate, domed, concave, and convex. In one embodiment, first elongated member 220 and second elongated member 250 have mating cross-sectional shapes, thereby further stabilizing the nuclear fuel assemblies during operation of the nuclear reactor. Referring next to the embodiments shown in FIGS. 18A-18B, the relative positions of first tubular body 210 and second tubular body 240 varies the engagement of first elongated member 220 and second elongated member 250. As shown in FIG. 18A, first tubular body 210 is positioned parallel to second tubular body 240. As shown in FIG. 18B, first tubular body 210 is angularly offset from second tubular body 240. According to one embodiment, angularly offsetting first tubular body 210 from second tubular body 240 facilitates engagement of first elongated member 220 with second elongated member 250 along the length of the fuel assembly, thereby reducing dilation and improving the service life of the fuel assembly. While shown with a single elongated member, it should be understood that a plurality of elongated members may extend from sidewalls of the tubular bodies. Such elongated members may be angularly offset toward or away from one another, according to various embodiments, where one tubular body is angularly offset from another tubular body. Tapered Elongated Members Referring to FIGS. 19A-20, first elongated member 220 has a tapered cross-sectional shape. As shown in FIGS. 19A-19B, first tubular body 210 extends along a longitudinal axis. Plane 410 is perpendicular to outer face 216 and extends along the longitudinal axis. According to one embodiment, first elongated member 220 has a cross sectional shape that is tapered such that a first end 228 of first elongated member 220 has a first thickness T1 and a second end 229 of first elongated member 220 has a second thickness T2. As shown in FIGS. 19A-19B, the second thickness is greater than the first thickness. In other embodiments, the first thickness is greater than the second thickness. In one embodiment, the thinner portion of first elongated member 220 is positioned toward the top of the nuclear reactor core. Where the temperature of the core decreases from a higher-temperature upper region to a lower-temperature lower region, higher-temperature coolant interfaces with the upper portion of first elongated member 220, thereby causing greater thermal expansion relative to the expansion caused by the interaction of the coolant toward the bottom of the nuclear reactor core with the lower portion of first elongated member 220. In one embodiment, the thickness of first elongated member 220 varies from the first thickness to the second thickness according to a specified profile such that the contact surfaces of neighboring elongated members engage along a plane that is parallel to the extension axes of the tubular bodies. In one embodiment, the specified profile provides a thickness of first elongated member 220 that is uniform along the length of first tubular body 210 during operation of the nuclear reactor. In another embodiment, the specified profile controls the bulge of first sidewall 212 during operation of the nuclear reactor (e.g., controls the bulge of first sidewall 212 as a function of time). Controlling the bulge of first sidewall 212 may provide expansion characteristics that are matched for first tubular body 210 and the pin bundle, which may be exposed to different temperatures or flux and may be manufactured from different materials (e.g., D9 for first tubular body 210 and HT-9 for the pins of the pin bundle). The specified profile may accommodate for pins that swell faster and push or bulge the tubular bodies by having elongated members of neighboring fuel assemblies engaged or under more compression at an earlier time during irradiation. In another embodiment, the specified profile may reduce the risk that the tubular bodies will swell faster than the pins of the pin bundle, which would leave a bypass flow path for coolant around the pin bundle, by having elongated members of neighboring fuel assemblies engaged at an earlier time during irradiation. Referring to FIG. 20, first elongated member 220 has a tapered cross-sectional shape within a plane 420, which is parallel to outer face 216. According to the embodiment shown in FIG. 20, first end 228 of first elongated member 220 has a first width W1 and second end 229 of first elongated member 220 has a second width W2. As shown in FIG. 20, the second width is greater than the first width. In other embodiments, the first width is greater than the second width. In one embodiment, at least one of the first width and the second width is specified to control the bulge of first sidewall 212 during operation of the nuclear reactor (e.g., control the bulge of first sidewall 212 as a function of time). Area of Elongated Members Referring still to FIGS. 19A-20, the ratio of the surface area of outer face 216 to the surface area of first contact surface 222 defines an area ratio. The surface area of outer face 216 is determined by multiplying the height of first tubular body 210 by the width of first sidewall 212, according to one embodiment. The surface area of first contact surface 222 is determined by multiplying the length of first contact surface 222 by the width of first elongated member 220, according to the embodiment shown in FIGS. 19A-19B. According to the embodiment shown in FIG. 20, where the width of first elongated member 220 changes linearly from the first thickness to the second thickness, the area of first contact surface 222 is determined by multiplying the height of first elongated member 220 by the sum of the first width and the second width and dividing by two. In one embodiment, the area ratio is greater than one. In another embodiment, the area ratio is between one and ten. In yet another embodiment, the area ratio is between one and five. According to the embodiments shown in FIGS. 19A-20, the portion of first elongated member 220 at outer face 216 has a surface area that is equalized with the surface area of first contact surface 222. As shown in FIGS. 19A-20, first elongated member 220 includes a pair of sidewalls that are perpendicular to the outer face 216. As shown in FIG. 19A, the pair of sidewalls extends along a length of first tubular body 210. As shown in FIG. 20, the pair of sidewalls is angularly offset from the length of first tubular body 210. In one embodiment, the first contact surface 222 is hardened. By way of example, first contact surface 222 may be hardened using welding, PVD, electrochemical deposition, or still other techniques. Hardening first contact surface 222 improves the wear resistance of first elongated member 220, according to one embodiment, thereby improving the performance of the fuel assembly. Hardening first contact surface 222 may also reduce the risk of coupling neighboring elongated members, which would increase the force needed to extract a fuel assembly. Illustrative Methods FIGS. 21A-22E are a series of flowcharts depicting implementations. For ease of understanding, the flowcharts are organized such that the initial flowcharts present implementations via an example implementation and thereafter the following flowcharts present other implementations and/or expansions of the initial flowchart(s) as either sub-component operations or additional component operations building on one or more earlier-presented flowcharts. Those having skill in the art will appreciate that the style of presentation utilized herein (e.g., beginning with a presentation of a flowchart(s) presenting an example implementation and thereafter providing additions to and/or further details in subsequent flowcharts) generally allows for a rapid and easy understanding of the various process implementations. In addition, those skilled in the art will further appreciate that the style of presentation used herein also lends itself well to modular and/or object-oriented program design paradigms. Embodiments of the duct may be manufactured in one or more pieces and may be drawn with integrated load pads and elongated members, by changing the dies to effect the desired end shape. In another embodiment, the elongated members may be manufactured using a machining, welding, or fastening process. Elongated members and features added to the elongated members may be reduced in volume by machining or drilling to reduce the amount of structural material without affecting the duct performance. FIGS. 21A-21J provide illustrative flow diagrams for a method of manufacturing a duct for a nuclear fuel assembly, shown as method 500, according to one embodiment. Although the method is presented as a sequence of steps for illustrative purposes, this sequence does not limit the scope of the claimed methods, and those of ordinary skill in the art will be aware of modifications and variations that may be made to the sequence. Referring to FIG. 21A, method 500 starts at block 502. At block 504, a tubular body is provided. In one embodiment, the tubular body has a sidewall with an inner face and an outer face. The tubular body is configured to contain nuclear fuel within a fuel region. At block 506, an elongated member is defined. In one embodiment, the elongated member is defined on the outer face along at least a portion of the fuel region. The elongated member has a contact surface configured to stabilize the duct during operation of the nuclear fuel assembly. In one embodiment, method 500 stops at block 508. In other embodiments, method 500 continues. Additional method steps are set forth below by way of non-limiting example. Referring to FIG. 21B, a recess is defined in the elongated member at block 510. Referring to FIG. 21C, a plurality of recesses are defined in the elongated member at block 512. Referring to FIG. 21D, in some embodiments, defining a plurality of recesses in the elongated member at block 512 may include defining the plurality of recesses according to a specified pattern at block 514. Referring to FIG. 21E, a projection is defined on the elongated member at block 516. Referring to FIG. 21F, a plurality of projections are defined on the elongated member at block 518. Referring to FIG. 21G, in some embodiments, an area ratio of the surface areas of the outer face to the contact surface is defined at block 520. In one embodiment, the elongated member includes a plurality of plates defining a plurality of slots that extend across a longitudinal axis of the tubular body. Referring to FIG. 21H, an upper blanket end and a lower blanket end are defined along the fuel region at block 522. As shown in FIG. 21I, the lengths of the plurality of plates is decreased from a first length at the upper blanket end to a second length at the lower blanket end at block 524. Referring to FIG. 21J, the contact surface is hardened at block 526. FIGS. 22A-22E provide illustrative flow diagrams for a method 600 of loading a nuclear reactor, according to one embodiment. Although the method is presented as a sequence of steps for illustrative purposes, this sequence does not limit the scope of the claimed methods, and those of ordinary skill in the art will be aware of modifications and variations that may be made to the sequence. Referring to FIG. 22A, method 600 starts at block 602. At block 604, a first fuel assembly is positioned within a nuclear reactor core. In one embodiment, the first fuel assembly includes a first tubular body having a sidewall with an inner face and an outer face, and the first tubular body is configured to contain nuclear fuel within a fuel region. The first fuel assembly also includes a first elongated member extending from the outer face along at least a portion of the fuel region, the first elongated member having a first contact surface. At block 606, a second fuel assembly is positioned alongside the first fuel assembly. In one embodiment, the second fuel assembly includes a second tubular body having a sidewall with an inner face and an outer face and a second elongated member extending from the outer face. The second elongated member has a corresponding contact surface configured to engage the first contact surface to stabilize the first tubular body and the second tubular body during operation of the nuclear reactor. In one embodiment, method 600 stops at block 608. In other embodiments, method 600 continues. Additional method steps are set forth below by way of non-limiting example. Referring to FIG. 22B, the first tubular body is positioned parallel to the second tubular body at block 610. Referring to FIG. 22C, the first tubular body is positioned angularly offset from the second tubular body at block 612. Referring to FIG. 22D, the first contact surface is spaced from the corresponding contact surface at block 614. Referring to FIG. 22E, in some embodiments, spacing the first contact surface from the corresponding contact surface at block 614 may include spacing the first contact surface less than two millimeters from the corresponding contact surface at block 616. With respect to the use of substantially any plural and/or singular terms herein, those having skill in the art can translate from the plural to the singular and/or from the singular to the plural as is appropriate to the context and/or application. The various singular/plural permutations are not expressly set forth herein for sake of clarity. The herein described subject matter sometimes illustrates different components contained within, or connected with, different other components. It is to be understood that such depicted architectures are merely exemplary, and that in fact many other architectures may be implemented which achieve the same functionality. In a conceptual sense, any arrangement of components to achieve the same functionality is effectively “associated” such that the desired functionality is achieved. Hence, any two components herein combined to achieve a particular functionality can be seen as “associated with” each other such that the desired functionality is achieved, irrespective of architectures or intermedial components. Likewise, any two components so associated can also be viewed as being “operably connected”, or “operably coupled,” to each other to achieve the desired functionality, and any two components capable of being so associated can also be viewed as being “operably couplable,” to each other to achieve the desired functionality. Specific examples of operably couplable include but are not limited to physically mateable and/or physically interacting components, and/or wirelessly interactable, and/or wirelessly interacting components, and/or logically interacting, and/or logically interactable components. In some instances, one or more components may be referred to herein as “configured to,” “configured by,” “configurable to,” “operable/operative to,” “adapted/adaptable,” “able to,” “conformable/conformed to,” etc. Those skilled in the art will recognize that such terms (e.g. “configured to”) can generally encompass active-state components and/or inactive-state components and/or standby-state components, unless context requires otherwise. While particular aspects of the present subject matter described herein have been shown and described, it will be apparent to those skilled in the art that, based upon the teachings herein, changes and modifications may be made without departing from the subject matter described herein and its broader aspects and, therefore, the appended claims are to encompass within their scope all such changes and modifications as are within the true spirit and scope of the subject matter described herein. It will be understood by those within the art that, in general, terms used herein, and especially in the appended claims (e.g., bodies of the appended claims) are generally intended as “open” terms (e.g., the term “including” should be interpreted as “including but not limited to,” the term “having” should be interpreted as “having at least,” the term “includes” should be interpreted as “includes but is not limited to,” etc.). It will be further understood by those within the art that if a specific number of an introduced claim recitation is intended, such an intent will be explicitly recited in the claim, and in the absence of such recitation no such intent is present. For example, as an aid to understanding, the following appended claims may contain usage of the introductory phrases “at least one” and “one or more” to introduce claim recitations. However, the use of such phrases should not be construed to imply that the introduction of a claim recitation by the indefinite articles “a” or “an” limits any particular claim containing such introduced claim recitation to claims containing only one such recitation, even when the same claim includes the introductory phrases “one or more” or “at least one” and indefinite articles such as “a” or “an” (e.g., “a” and/or “an” should typically be interpreted to mean “at least one” or “one or more”); the same holds true for the use of definite articles used to introduce claim recitations. In addition, even if a specific number of an introduced claim recitation is explicitly recited, those skilled in the art will recognize that such recitation should typically be interpreted to mean at least the recited number (e.g., the bare recitation of “two recitations,” without other modifiers, typically means at least two recitations, or two or more recitations). Furthermore, in those instances where a convention analogous to “at least one of A, B, and C, etc.” is used, in general such a construction is intended in the sense one having skill in the art would understand the convention (e.g., “a system having at least one of A, B, and C” would include but not be limited to systems that have A alone, B alone, C alone, A and B together, A and C together, B and C together, and/or A, B, and C together, etc.). In those instances where a convention analogous to “at least one of A, B, or C, etc.” is used, in general such a construction is intended in the sense one having skill in the art would understand the convention (e.g., “a system having at least one of A, B, or C” would include but not be limited to systems that have A alone, B alone, C alone, A and B together, A and C together, B and C together, and/or A, B, and C together, etc.). It will be further understood by those within the art that typically a disjunctive word and/or phrase presenting two or more alternative terms, whether in the description, claims, or drawings, should be understood to contemplate the possibilities of including one of the terms, either of the terms, or both terms unless context dictates otherwise. For example, the phrase “A or B” will be typically understood to include the possibilities of “A” or “B” or “A and B.” With respect to the appended claims, those skilled in the art will appreciate that recited operations therein may generally be performed in any order. Also, although various operational flows are presented in a sequence(s), it should be understood that the various operations may be performed in other orders than those which are illustrated, or may be performed concurrently. Examples of such alternate orderings may include overlapping, interleaved, interrupted, reordered, incremental, preparatory, supplemental, simultaneous, reverse, or other variant orderings, unless context dictates otherwise. Furthermore, terms like “responsive to,” “related to,” or other past-tense adjectives are generally not intended to exclude such variants, unless context dictates otherwise. Those skilled in the art will appreciate that the foregoing specific exemplary processes and/or devices and/or technologies are representative of more general processes and/or devices and/or technologies taught elsewhere herein, such as in the claims filed herewith and/or elsewhere in the present application. While various aspects and embodiments have been disclosed herein, other aspects and embodiments will be apparent to those skilled in the art. The various aspects and embodiments disclosed herein are for purposes of illustration and are not intended to be limiting, with the true scope and spirit being indicated by the following claims.
	summary
050892185	description	A water cooled nuclear reactor 10 with integral pressurizer 62 according to the present invention is shown in FIG. 1. This is a pressurized water reactor with an indirect cooling system. The water cooled nuclear reactor 10 comprises a pressure vessel 12 within which is positioned a reactor core 14. The reactor core 14 is positioned at the lower region of the pressure vessel 12, and the reactor core 14 is surrounded by a neutron reflector 16. A thermal shield 18 is positioned below the reactor core 14 and thermal shields 20 are positioned so as to surround the neutron reflector 16. The thermal shields 18,20 protect the pressure vessel 12 from radiation emanating from the reactor core 14. A primary water coolant circuit is used to cool the reactor core 12, and the primary water coolant circuit may use a forced or a natural circulating arrangement. The primary water coolant circuit comprises a hollow cylindrical member 22 which is aligned with and positioned vertically above the reactor core 14 to define a riser passage 24 therein for the natural vertically upward flow of relatively hot primary coolant from the reactor core 14. An annular downcomer passage 26 is defined between the hollow cylindrical member 22 and the pressure vessel 12 for the natural vertically downward return flow of relatively cool primary coolant to the reactor core 14. A casing 28 is positioned in the pressure vessel 12, and divides the pressure vessel 12 into a first vertically upper chamber 30 and a second vertically lower chamber 32. The cylindrical member 22 extends towards the top of the lower chamber 32 defined by the casing 28 and the pressure vessel 12, and the upper region of the cylindrical member 22 is provided with apertures 34 for the distribution of flow of the primary water coolant from the riser passage 24 to the heat exchanger region of the annular downcomer passage 26. A secondary coolant circuit takes heat from the primary water coolant circuit to produce vapor for a turbine or hot fluid for other purposes. The secondary water coolant circuit comprises a heat exchanger, which for example is a steam generator, 40 which is annular and positioned coaxially in the upper region of the annular downcomer passage 26. The heat exchanger 40 comprises one or more tubes, which are arranged in the annular vessel 42, which receive fluid from a supply of fluid via a supply pipe and inlet header, and which supply vapor to a turbine or hot fluid for other purposes via an outlet header and a supply pipe. The steam tubes are of any suitable configuration for example, as is well known in the art, the tubes could be u-tubes or helically coiled tubes which extend between the inlet header and outlet header. The helically coiled tubes may be arranged in tube bundles arranged circumferentially with the annular vessel 42. British Patent No 1386813 discloses a pressurized water reactor which has helically coiled tubes arranged in an annular steam generator, although the primary water coolant is pumped therethrough normally, natural water circulation takes place if there is a pump failure, this arrangement does not have an integral pressurizer. The pressure vessel 12 and the cylindrical member 22 are provided with members 42 and 44 respectively which extend into the downcomer passage 26. A pump 46 is positioned in the downcomer passage 26 between the members 42 and 44 to pump the primary water coolant through the primary water coolant circuit. The pump 46 is driven by a motor 48, which is positioned outside of the pressure vessel 12. The casing 28 has an annular member 60 which extends vertically downwards from the peripheral region of the casing 28. The bottom region of the annular member 60 is secured to the pressure vessel 12. The pressurizer 62 is positioned within the pressure vessel 12 in the upper chamber 30 formed between the casing 28 and the pressure vessel 12. A movable diaphragm 64 is positioned in the upper chamber 30, and is sealingly secured to a flange 66 which extends inwardly from the pressure vessel 12. A bellows 68 arrangement secures the diaphragm 64 to the flange 66, and also forms a seal between the diaphragm 64 and flange 66. The diaphragm 64 divides the upper chamber 30 into a first water space 70 and a second fluid, or gas space 72. The water space 70 is formed below the diaphragm 64 and the fluid, or gas space 72 is formed above the diaphragm 64. The water space 70 is defined by the pressure vessel 12, the casing 28, the diaphragm 64 and the bellows 68, and the fluid space 72 is defined by the pressure vessel 12, the diaphragm 64 and the bellows 68. The bottom region of the annular member 60 is provided with a plurality of circumferentially arranged surge ports 74. The surge ports 74 fluidly communicate between the water space 70 and the annular downcomer passage 26 of the primary coolant circuit, and as shown may extend into the heat exchanger 40 or into the downcomer beneath the heat exchanger. A plurality of springs 76 are provided which are secured to the diaphragm 64 and the pressure vessel 12 so as to preload the diaphragm, alternatively springs could be built into the bellows 68. A plurality of dampers 78 are provided to control the movement of the diaphragm 64, so as to give suitable transient and frequency response characteristics. The dampers 78 are of the dashpot type and each comprises a rod 80 and piston 82, and a cylinder 84. Each rod 80 and piston 82 is secured to the diaphragm 64 and the cylinders 84 are secured to the casing 28. Each rod 80 and piston 82 is arranged coaxially of an associated cylinder 84 and is arranged to move axially therein in response to movement of the diaphragm. The cylinders 84 are filled with water to impede the movement of the pistons 82 so as to damp undesirable oscillations of the diaphragm 64. Other suitable dampers could be used to achieve the same function. In operation of the water cooled nuclear reactor 10 the fission of nuclear fuel in the reactor core 14 produces heat. The heat is carried away from the reactor core 14 by the primary water coolant circuit. The heating of the water in the vicinity of the reactor core 14 causes the water to flow in an upwards direction as shown by arrows A through the riser passage 24, the primary water then flows through the flow distribution apertures 34 in the cylindrical member 22 and apertures 44 in the annular vessel 42 of the steam generator, heat exchanger i.e. 40 to pass over heat exchanger tubes as shown by arrows B. The primary water gives heat to the secondary fluid in the tubes on passing through the heat exchanger 40. The primary water then returns to the reactor core 14 through the annular downcomer passage 26 as shown by arrow C. If there is a positive volume surge in or expansion of, the water of the primary water coolant circuit, due to an excess of reactor power over heat exchanger load, a portion of the primary coolant flows from the annular downcomer passage 26 through the surge ports 74 into the water space 70. This produces an increase of the volume of the water space 70, and a corresponding reduction of the volume of fluid space 72, by causing the diaphragm 64 and bellows 68 to move against the spring 76 load, the bellows 68 expanding. If there is a negative volume surge in or contraction of the water of the primary water coolant circuit, due to a deficit in reactor power over heat exchanger load, a portion of the water flows from the water space 70 through the surge ports 74 into the annular downcomer passage 26. There is therefore a decrease of the volume of the water space 70, and a corresponding increase of the volume of fluid space 72, by the diaphragm 64 and bellows 68 moving under the action of the spring 76. The spring 76 loaded diaphragm 64 and bellows 68 allows the primary water coolant to expand or contract as the temperature of the water varies while the water is compressed over the full operating temperature range, and accommodates all normal or abnormal changes in the water volume without overpressurizing the primary coolant circuit of the reactor. The fluid space 72 volume is arranged so that at the operating pressure of the primary coolant circuit, the diaphragm 64 and bellows 68 can expand to accommodate the largest volume surge of the water. The diaphragm 64 must be strong enough and flexible enough to keep the water compressed. The diaphragm may be an elastic membrane. The diaphragm 64 and bellows 68 pressurizer does not have a constant pressure steady state characteristic, the pressure rises or falls with the temperature. The fluid space 72 is a closed gas filled space, and in this arrangement the load on the spring loaded diaphragm and bellows is reduced. The gas space is fitted with a relief valve 52 and vent 50 to relieve over pressures. The water pressure is the sum of the gas pressure in the fluid space 72 and a pressure component due to the diaphragm 64. There is a temperature and a pressure for the gas and water for which, because of the gas pressure, there is no load on the diaphragm 64 and bellows 68, it is preferable to design for operation below these temperatures and pressures. An alternative arrangement dispenses with the relief valve leaving the gas space open to atmospheric pressure via the vent. The water cooled nuclear reactor as shown in FIG. 1 is suitable for use where vertical height, headroom, of the reactor is not a constraint. A second embodiment of a water cooled nuclear reactor 110 with integral pressurizer 162 according to the present invention is shown in FIG. 2. This is a pressurized water reactor of the type with an indirect cooling system. The water cooled nuclear reactor 110 comprises a pressure vessel 112 within which is positioned a reactor core 114. The reactor core 114 is positioned at the lower region of the pressure vessel 112, and the reactor core 114 is surrounded by a neutron reflector 116. A thermal shield 120 is positioned so as to surround the neutron reflector 116 to protect the pressure vessel 112 from radiation emanating from the reactor core 114. A circulating primary water coolant circuit is used to cool the reactor core 114. The primary coolant circuit comprises a hollow cylindrical member 122 which is aligned with and positioned vertically above the reactor core 114 to define a riser passage 124 for the natural vertically upward flow of relatively hot primary coolant from the reactor core 114. An annular downcomer passage 126 is defined between the hollow cylindrical member 122 and the pressure vessel 112 for the natural vertically downward return flow of relatively cool primary coolant to the reactor core 114. A casing 128 is positioned in the pressure vessel 112, and divides the pressure vessel 112 into a first chamber 130 and a second chamber 132. The reactor core 114 and the primary coolant circuit are arranged in the second chamber 132. The cylindrical member 122 has apertures at its upper region for the flow of primary water coolant from the riser passage 124 to the annular downcomer passage 126. A secondary fluid circuit takes heat from the primary water coolant circuit for whatever purpose the plant is designed for. The secondary fluid circuit comprises a steam exchanger 140 which is annular and positioned coaxially in the upper region of the annular downcomer passage 126. The heat exchanger 140 is arranged in an annulus. The heat exchanger 40 also comprises one or more tubes which are arranged in the annulus which receive fluid from a supply of secondary fluid via a supply pipe and inlet header, and which supply hot fluid or vapor via an outlet header and a supply pipe. The casing 128 comprises an annular member which is sealingly secured to the pressure vessel 112. The casing 128 extends downwards from the pressure vessel 112, and the casing 128 is positioned coaxially with the hollow cylindrical member 122 and extends downwards into the hollow cylindrical member 122. The pressurizer 162 is positioned within the pressure vessel 112 in the first chamber 130 formed between the casing 128 and an upper portion 112B of the pressure vessel 112. A movable diaphragm 164 is positioned in the first chamber 130, and is sealingly secured to the pressure vessel 112. A bellows 168 arrangement secures the diaphragm 164 to the pressure vessel 112, and also forms a seal between the diaphragm 164 and pressure vessel 112. The diaphragm 164 divides the first chamber 130 into a first water space 170 and a second fluid space 172. The water space 170 is formed below the diaphragm 164 and the fluid space 172 is formed above the diaphragm 164. The water space 170 is defined by the pressure vessel 112, the casing 128, the diaphragm 64 and the bellows 168, and the fluid space 172 is defined by the pressure vessel 112, the diaphragm 164 and the bellows 168. The bottom region of the casing 128 is provided with at least one surge port 174, which fluidly communicates between the water space 170 and the riser passage 124 of the primary coolant circuit. The bellows 168 has built in springs to preload the diaphragm 164. The diaphragm 164 itself acts as a damper in this arrangement, the diaphragm 164 acts as a piston of a dashpot damper, and the casing 128 acts as a cylinder of the dashpot damper. This embodiment operates in substantially the same manner as the embodiment in FIG. 1, the bellows 168 however contracts when there is a positive volume surge, and expands when there is a negative volume surge. Water flows between the riser passage 124 of the primary coolant circuit and the first water space 170. The fluid space 172 has a vent 190 to connect the gas in the fluid space to atmosphere. The fluid space could be either connected to atmosphere permanently, or through a relief valve 192 positioned in the vent 190 to open at a predetermined pressure. In the latter the space between the diaphragm and the relief valve would be filled with a predetermined mass of gas. A pump 194 may be provided to assist the primary water coolant circulation. The water cooled nuclear reactor shown in FIG. 2 is suitable for use where there is limited vertical height, headroom, available for the reactor, i.e. produces a more compact reactor arrangement. Water cooled nuclear reactors with integral pressurizers according to the present invention cannot have control rod mechanisms suspended from the top of the pressure vessel, which extend downwards through the pressurizer, casing and riser passage to the reactor core, because of the diaphragm. Other methods of controlling the reactor core are required, such as by adjusting neutron absorption in the neutron reflector in the case of a very small reactor. The neutron reflector controls the escape of neutrons from the reactor core, and thus can be used to control the reactivity of the reactor core, such a control process is well known in the art, and is disclosed in U.S. Pat. No. 3,687,804. Alternatively hydraulic control rod actuators may be provided which do not require mechanical penetration through the diaphragm. Water cooled nuclear reactors with integral pressurizers according to the present invention could be used to provide low grade heat or used with moderate temperature organic or steam Rankine cycle power conversion to provide relatively low cost, low power plants between 200 KW and 10 MW of electricity. Water cooled nuclear reactors with integral pressurizers are particularly suitable for operation at relatively low temperatures and pressures. Typical temperature and pressure range is 100.degree.-200.degree. C. and 2 to 30 bars. An advantage of the water cooled nuclear reactor with integral pressuriser is that it accommodates changes in the primary water coolant volume or pressure without overpressurising the primary coolant circuit of the reactor. Also the pressuriser does not flood with water, and lower system pressure can be used. Dispersed or separate pressurizers may be provided with a diaphragm, a bellow arrangement, springs and dampers within a pressure vessel to operate in substantially the same manner. FIG. 3 illustrates a pressurized water cooled nuclear reactor 210 in which the pressurizer 262 is separate, or dispersed from the pressurized water cooled nuclear reactor with an integral primary water coolant circuit arrangement. The pressurized water cooled nuclear reactor 210 comprises a pressure vessel 212 within which is positioned a reactor core 214. The reactor core 214 is again positioned at the lower region of the pressure vessel 212, and the reactor core 214 is surrounded by neutron reflector 216. A thermal shield 220 is positioned so as to surround the neutron reflector 216 to protect the pressure vessel 212 from radiation emanating from the reactor core 214. A circulating primary water coolant circuit is used to cool the reactor core 214. The primary water coolant circuit comprises a hollow cylindrical member 222 which is aligned with and positioned vertically above the reactor core 214 to define a riser passage 224 for the natural vertically upward flow of relatively hot primary coolant from the reactor core 214. An annular downcomer passage 226 is defined between the hollow cylindrical member 222 and the pressure vessel 212 for the natural vertically downward return flow of relatively cool primary coolant to the reactor core 214. The cylindrical member 222 has flow distribution apertures 234 at its upper region for the flow of primary water coolant from the riser passage 224 to the annular downcomer passage 226. A heat exchanger 240 of a secondary fluid coolant takes heat from the primary water coolant circuit for whatever purpose the plant is designed for. The heat exchanger 240 is annular and is positioned coaxially in the upper region of the annular downcomer passage 226. The pressurizer 262 is positioned above the pressure vessel 12, and comprises a separate pressure vessel 263. The pressurizer may equally well be arranged level with the pressure, vessel. A movable diaphragm 264 is positioned in the pressure vessel 263 and is sealingly secured to the pressure vessel 263. A bellows 268 arrangement secures the diaphragm 264 to the pressure vessel 263, and also forms a seal between the diaphragm 264 and the pressure vessel 263. The diaphragm 264 divides the interior of the pressure vessel 263 into a first water space 270 and a second fluid space 272. A surge line 274 interconnects the water space 270 of the pressurizer 262 with the downcomer passage 226 of the primary water coolant circuit. Alternately the surge port 274 may interconnect with the primary water coolant circuit at the top of the pressure vessel 212 or other suitable position. The fluid space 272 has a vent 290 and a valve 292 to connect the gas in the fluid space to atmosphere. The embodiment in FIG. 2 operates substantially the same as the embodiments in FIGS. 1 and 2. A separate, or dispersed pressurizer with a diaphragm as shown in the embodiment in FIG. 3 may equally well be applicable to a low pressure dispersed, or loop type, pressurized water nuclear reactor i.e. a pressurized water nuclear reactor in which the heat exchanger of the secondary coolant fluid circuit is not integral in the pressure vessel of the plant.
	abstract	A fuel rod and a fuel assembly for light water reactors, in which crack penetration to a fuel cladding tube or an end plug can be prevented, are provided. The fuel rod 10a includes: a cylindrical cladding tube 11 formed of a ceramic base material; a connection 21 formed of the same material as the cladding tube 11; and an end plug 12a having a concave portion 12f of a continuously curved surface shape adapted to house the connection 21. The end plug 12a is formed of the same material as the cladding tube 11. A slanted surface 11a formed at an end portion of the cladding tube 11, and a slanted surface 12d formed at an end portion of the end plug 12a are joined in contact with each other with a metallic joint material 20. The joint is supported by the connection 21.
055454275	summary	The present invention relates to a process for the preparation of lithium aluminosilicate or gamma lithium aluminate ceramics, which are isostructural compounds of gamma lithium aluminate more particularly usable as a tritium-producing covering material for controlled thermonuclear fusion reactors. It is known that ceramics based on lithium oxide and in particular gamma lithium aluminate and its isostructural compounds are good candidates as tritium-producing covering material for nuclear fusion reactors. However, for such applications it is important to control the microstructure (grain size and porosity) of the ceramic in order to be able to then control the release kinetics of the tritium formed under neutron irradiation from the covering material. The known processes for preparing gamma lithium aluminate either use reactions in the solid state based on gamma aluminate, or liquid phase reactions, e.g. aluminium and lithium alkoxide hydrolysis reactions. Solid phase preparation reactions using alumina and lithium compounds such as lithium hydroxide or carbonate have e.g. been described by Kinoshita et al in Mat. Res. Bull., vol. 13, pp 445-455, 1978. These preparation procedures are not complicated, but generally suffer from the disadvantage of requiring complimentary crushing and sintering stages for the LiAlO.sub.2 powder. Moreover, as described by Bernard Rasneur in Advances in Ceramics, vol. 27, 1990, pp 63-76, the density obtained after sintering remains low and the microstructure of the material is dependent on the stoichiometry. In addition, liquid phase preparation processes have been developed and in particular those of the sol-gel type, which make it possible to prepare beta or gamma lithium aluminate powder by hydrolysis of lithium and aluminium alkoxides or alcoholates, followed by calcination. Processes of this type are e.g. described by Turner et al in Advances in Ceramics, vol. 25, 1989, pp 141-147. In order to carry out said hydrolysis, it is possible to use different lithium and aluminium alkoxides in an alcohol, but the hydrolysis reaction generally takes place with the alkoxide in the dispersed phase, because most lithium and aluminium alkoxides are solid and not very soluble in alcohols. However, it is possible to use secondary aluminium butoxide, which is liquid at ambient temperature and react it with a lithium hydroxide solution. However, with these hydrolysis processes, it is necessary to then convert the amorphous or slightly crystalline powder in the beta LiAlO.sub.2 into sintered gamma lithium aluminate. This can be carried out by a heat treatment, but the weight losses during the latter are approximately 40%, which implies the prior calcination of the powder before the actual sintering in order to prevent shrinkage and bursting problems during sintering. In addition, in the case where use is made of secondary aluminium butoxide, the authors indicate that the latter can only be diluted in secondary butanol in order to avoid prejudicial fast exchanges between another alcohol and said alkoxide. Hydrolysis can also take place on the basis of a mixture of lithium and aluminium alkoxides as described by Hirano et al, J. Am. Ceram. Soc., vol. 70, 1987, pp 171-174. In this case use is made of a mixture of lithium and aluminium ethoxides in ethanol with a concentration of 0.1 mole/l for each alkoxide, followed by hydrolysis by the addition of decarbonated distilled water. After hydrolysis, the precipitate is refluxed for 24 hours, then cooled and subjected to ultrafiltration under a nitrogen pressure of 500 kPa before being microwave dried. The powder obtained, which is beta lithium aluminate, then undergoes calcination at 750.degree. C. for conversion into gamma lithium aluminate and the weight loss recorded during the heat treatment is approximately 20%. Thus, this calcination stage must be carried out prior to the compression and final sintering of the powder at a temperature of 1000.degree. to 1450.degree. C. and under oxygen. The known processes consequently suffer from the disadvantage of requiring either a high temperature crushing and sintering stage, or a calcination stage followed by pressing and sintering stages, which does not make it possible to accurately control the microstructure (density, grain size) and stoichiometry of the gamma lithium aluminate ceramics obtained. The present invention specifically relates to a process for the preparation of ceramics of gamma lithium aluminate or its isostructural compounds making it possible to obviate the disadvantages of the known processes. According to the invention, the process for the preparation of lithium aluminosilicate or aluminate ceramics of formula EQU Li.sub.4+x Al.sub.4-3x Si.sub.2x O.sub.8 with 0.ltoreq.x<0.28, comprises the following stages: a) mixing in a short chain anhydrous alcohol a liquid, unpolarized aluminium alkoxide and, in the case where x is different from 0, a silicon alkoxide, with a hydrated or unhydrated lithium hydroxide, PA0 b) adding water to the mixture obtained in stage a) in order to hydrolyze it, PA0 c) drying at a temperature below 300.degree. C. the hydrolyzed product obtained in stage b) in order to evaporate the alcohols and water and obtain a crystalline powder with a structure identical to that of beta LiAlO.sub.2, PA0 d) shaping the powder obtained in stage c) by isostatic or non-isostatic cold pressing, by pouring a slop, by spinning or by extruding and PA0 e) subjecting the shaped powder to a thermal sintering treatment at a temperature of 800.degree. to 1200.degree. C. in order to obtain a sintered aluminosilicate or gamma lithium aluminate ceramic. In this process, through the use as the starting alkoxide of an unpolymerized liquid alkoxide and through the performance of the reaction in the presence of a short chain anhydrous alcohol, it is possible to obtain a high hydrolyze reactivity favouring the formation by hydrolysis of well crystallized gamma or beta lithium aluminate. Thus, aluminium alkoxides Al(OR).sub.3 and in particular those in which R is a lower alkyl group such as methyl, ethyl or propyl, have a marked tendency to polymerize in order to form alkoxy-type, stable molecular associations: ##STR1## Therefore, these polymerized alkoxides are generally solid at ambient temperature and only slightly soluble in alcohols. Therefore the reactivity of the alkoxide is reduced and the OR groups are difficult to hydrolyze. There is a reduced tendency on the part of liquid alkoxides in which R is a more cumbersome alkyl radical, e.g. secondary butoxide to form molecular associations with stable alkoxy bridges, but their reactivity relative to water is low, because said reactivity decreases when the size of the alkyl group R increases. Therefore they are also only slightly reactive for hydrolysis. According to the invention reactivity losses due to molecular associations by alkoxy bridges and the size of the alkyl group R are prevented by starting with an unpolymerized liquid alkoxide having a relatively cumbersome group R and by exchanging said group R by a lower alkyl group from the short chain anhydrous alcohol just prior to the reaction in order to obtain a good reactivity leading to the formation of beta LiAlO crystallized by hydrolysis followed solely by a drying operation. Moreover, through starting with a liquid alkoxide it is possible to have a high aluminium concentration in the solution. In order to favour the exchange of alkyl groups without giving rise to molecular associations, preferably in stage a) preparation takes place of a solution of aluminium alkoxide and optionally silicon alkoxide in short chain anhydrous alcohol. Preferably working takes place under an inert atmosphere, e.g. nitrogen, in order to avoid the presence of water and the irreversible formation of oxo Al-O-Al bridges. Hydrated or unhydrated lithium hydroxide is rapidly added to this solution and mixing takes place under stirring. Thus, by adding lithium hydroxide to the freshly prepared unpolymerized liquid aluminium alkoxide solution in a short chain alcohol, it is possible to avoid the formation of molecular associations by alkoxy bridges after exchange of the alkyl groups with those of the alcohol. Preferably, working takes place under an inert gas and e.g. nitrogen atmosphere in order to avoid the presence of water. The hydrated or unhydrated lithium hydroxide can be added to the freshly prepared solution in the form of powder, solution or suspension in an alcohol. It is preferably added in the form of a suspension in the same short chain alcohol as that of the solution. Preference is given to the use of monohydrated lithium hydroxide. The aluminium oxide usable in stage a) of the process according to the invention can e.g. be aluminium isopropoxide dissolved in secondary butanol or secondary butoxide of pure aluminium. Preference is given to the use of secondary aluminium butoxide. The silicon alkoxide can e.g. be tetraethoxysilane. The short chain alcohols used with these alkoxides are e.g. methanol, ethanol or propanol. Preference is given to the use of ethanol, which is less hydrophilic than methanol. In stage a) of the present process, partial hydrolysis takes place with a temperature rise and this leads to a solution in which the lithium and aluminium are intimately mixed, optionally with silicon. Starting with very pure products, it is possible in this way to obtain by subsequent hydrolysis a very homogeneous and very pure product, which by drying leads to a fine crystalline powder with a structure identical to that of the beta LiAlO.sub.2 and which does not give rise to a significant weight loss during subsequent baking. In addition, unlike in the case of the prior art processes, this powder can be directly transformed into sintered aluminium aluminosilicate or gamma lithium aluminate ceramic without it being necessary to carry out a prior calcination. Thus, the weight loss recorded during heating is very low, probably because the reactions are more complete starting with an unpolymerized, liquid aluminium alkoxide and lithium hydroxide in a short chain alcohol, which by exchange of said alkyl groups with alkoxide makes it possible to improve the reactivity of the alkoxide to hydrolysis. The transformation of the crystalline powder with a structure identical to that of .beta.-LiAlO.sub.2 into .gamma.-LiAlO.sub.2 ceramic or lithium aluminosilicate and its sintering are then obtained by carrying out the shaping stage d) and heat treatment stage e) described hereinbefore. During the heat treatment, the powder having a structure identical to that of beta LiAlO.sub.2 is transformed into lithium aluminosilicate or gamma LiAlO.sub.2 powder at about 600.degree. to 700.degree. C., followed by a sintering of the powder with the growth of grains at temperatures exceeding 800 to 1200.degree. C. The elimination of the prior powder calcination stage is of particular interest, because said stage is eliminated, whilst having a powder which is more reactive for sintering, so that the latter can be carried out at lower temperatures permitting a perfect control of the microstructure of the sintered ceramic. Thus, with temperatures of 800.degree. to 1200.degree. C., it is possible to obtain ceramics whose density varies from 70 to 100% of the theoretical density with grains having a homogeneous size, whose average dimensions are 0.1 to 10 .mu.m, as a function of the temperature used. Generally, sintering takes place in an oxygen atmosphere, e.g. in air. Moreover, during sintering, the powder parts having a crystalline structure identical to that of beta lithium aluminate are embedded in a powder bed having the sane composition in order to provide a good stoichiometry control. In the process according to the invention, there is a control of the characteristics of the powder (stoichiometry, crystallinity, particle size, etc.) and consequently its suitability for sintering, by regulating in an appropriate manner the molar ratio of the short chain alcohol to the aluminium alkoxide and the lithium hydroxide quantity. Advantageously, the alcohol:aluminium alkoxide molar ratio is 8 to 30 at the end of stage a). The lithium hydroxide quantity generally corresponds to the stoichiometric quantity, but it is possible to regulate the density of the end product by using non-stoichiometric lithium hydroxide quantities in order to have a range extending e.g. up to 4 molar %. There is also a control of the characteristics of the hydrolyzed product by adding in stage b) a water quantity such that the molar ratio of the water to the aluminium alkoxide is 5 to 20. In stage a) of the process according to the invention, the alkoxide solution is prepared by mixing, under stirring, the said alkoxide or alkoxides with alcohol, followed by the immediate addition of lithium hydroxide and vigorous stirring takes place for an adequate time during which the temperature rises to approximately 80.degree. C., in order to obtain an intimate mixture of the lithium and the aluminium. Generally this time is between 20 and 60 minutes for a volume of 600 ml. There is a partial hydrolysis in stage a), but the complete hydrolysis is then carried out in stage b) by adding water, accompanied by stirring. For said stage b) use is preferably made of deionized and decarbonated water in order to obtain a very pure powder by hydrolysis. Preferably stage a) is carried out under an atmosphere of inert gas, e.g. nitrogen, because alkoxides are very sensitive to moisture. In stage c) the product obtained by hydrolysis is dried at a temperature below 300.degree. C. This can e.g. take place at 150.degree. C. in an oven or at a temperature above the critical point of ethanol in an autoclave, e.g. at 250.degree. C. under 7 MPa. In the process according to the invention, as has been shown hereinbefore, it is possible to simultaneously regulate the size of the grains and the density by choosing appropriate sintering conditions. It is also possible to adjust the density of the end product independently of the size of the grains, either by using a lithium hydroxide quantity such that it corresponds to a slight stoichiometry variation extending e.g. up to 4 molar %, based on the pure aluminosilicate or aluminate, or by adding to the starting solution at least one doping agent such as Na, K or Zn and a transition element. The doping agent can e.g. be added in hydroxide, nitrate or ethoxide form. Other features and advantages of the invention will become more apparent from reading the following examples given in a purely illustrative and non-limitative manner.
	abstract	The method concerns processing irradiated (spent) nuclear fuel (SNF), it is primarily aimed at isolating and trapping tritium, and can be used in nuclear power industry for treating SNF. This method provides for a two-phase voloxidation of a reaction mass using gas-air mixture, the reaction mass including fragmented uranium dioxide SNF elements with containers. The first phase is carried out at 400-650° C. in the presence of air and additional carbon dioxide. The second phase is carried out at 350-450° C. using a stream of an air-vapor mixture that can be oxygen-enriched. Both phases are carried out with a repeated mechanical activation of the reaction mass. Provided in the course of the voloxidation is the gas replacement at the hour rate of about 10-50 fold the reaction chamber gas volume. Before being introduced into the reaction chamber, the gas is preheated up to the chamber internal temperature.
046506390	summary	The invention concerns a method and an apparatus for eliminating leakage spaces between the partitions surrounding the core of a pressurized water nuclear reactor, after this nuclear reactor is brought into operation. BACKGROUND In pressurized water nuclear reactors, the core of the reactor is constituted by very long, square-sectioned fuel assemblies constituted by a bundle of rods containing the fuel material and disposed side by side, vertically and in contact at their side faces. The transverse section of the core constituted by the juxtaposition of the square sections of the various assemblies is shaped like an irregular polygon whose perimeter has many steps. The whole of the reactor core is held inside partitioning, in contact with the outwardly directed faces of the peripheral assemblies, over the whole height of these assemblies. The core and its partitioning are also surrounded by a cylindrical shell, termed core casing, which provides a space, between the partitioning and its inner surface, inside which horizontal reinforcing plates are positioned ensuring that the partitioning, which is itself constituted by flat plates almost as long as the height of the core, is assembled and kept in position. The reinforcing plates are pierced by openings allowing cooling water to circulate in the space between the partitioning and the core casing. It is necessary in practice to cool the partitioning by the circulation of water over its outer surface and this cooling water can be introduced at the upper part of this space via water inlet orifices provided in the core casing. As it circulates, the pressurized cooling water of the reactor enters the core via its lower part, passes through the assemblies vertically from bottom to top and is collected by the hot branches of the primary circuit at the upper part of the core. On its return, the cooling water from the partitioning runs through the space between the partitioning and the core casing, vertically from top to bottom before combining with the water entering the core at the lower part thereof. The loss of head of the pressurized water when it runs through the interior of the assemblies, depending on the height of the core, produces a difference in pressure, at the upper part of the core, between the water for cooling the core and the water for cooling the partitioning. In the case of cooling the partitioning by descending current, this difference in pressure is of the order of 2 bars. The vertical plates constituting the partitioning are simply juxtaposed and assembled at right angles by screws. In some pressurized water nuclear reactors currently operating, the assembly between some partitions has only a small number of screws, so that a leakage space can occur between the corresponding partitioning elements. Because of the difference in pressure existing, at least at the upper part of the partitioning, between the core region and the peripheral region between the partitioning and the core casing, pressurized jets of water directed from the exterior to the interior of the core create regions of turbulence in the neighbouring assemblies and cause vibrations in the rods adjacent to these leakage regions which can cause them to deteriorate in the long run. To solve this problem, an attempt has been made to reduce the leakage spaces between such partitions, inside nuclear reactors, by hammering the joints between partitions responsible for these phenomena. Such a hammering operation must be carried out during a reactor shutdown, under water, with special tooling since the reactor materials are contaminated after use of this reactor. Before the hammering operation, the joints with too great play must be identified and then, after hammering, a check must be made that the play is sufficiently small to practically eliminate the occurrence of pressurized jets through the partitioning. These operations are therefore relatively complex and take a long time. In addition, when the reactor is in service, the joints can be displaced again so that the pressurized jets are very likely to reappear some time after the hammering operation. SUMMARY OF THE INVENTION The object of the invention is therefore to propose a method of eliminating leakage spaces between the partitions surrounding the core of a pressurized water nuclear reactor, after this reactor has been brought into operation, in which the core constituted by vertically disposed prismatic fuel assemblies is surrounded over its entire height by vertical flat partitions connected at right angles in pairs so as to constitute a partitioning whose horizontal section inside the circular section of the cylindrical casing of the core is a polygon contiguously surrounding the core whose horizontal section is correspondingly shaped, the partitions being assembled by horizontal reinforcing pieces disposed between the outer surface of the partitions and the inner surface of the core casing pierced with openings allowing the circulation of water, vertically, in the space provided between the partitioning and the core casing, this method allowing the leakage spaces to be permanently eliminated without deforming the partitions. To achieve this, during a reactor shutdown and with the core under water: (a) the joins between two right-angled partitions, between which a leakage space is capable of creating a leakage of cooling water when the reactor is operating, in the direction of the core, are identified and then, in the case of each join, PA1 (b) a first bore is made in the two partitions where they join, the bore passing through the entire thickness of one of the two partitions and opening in the other partition, in the direction of its width, PA1 (c) the bore in the first partition is widened to form a housing with a bearing surface perpendicular to the bore, PA1 (d) the interior of the bore is screw-threaded over part of its length, inside the second partition, PA1 (e) the swarf is recovered, PA1 (f) a screw with diametrical expansion is introduced and screwed in the bore until a tightness reducing the play at the screw is obtained, the screw-head bearing on the bearing surface inside the housing, PA1 (g) the screw is expanded by displacement of a rod inside this, in the longitudinal direction, PA1 (h) operations (b) to (g) are repeated at certain locations along the join between the two partitions until the leakage is insignificant, and then at the other identified joins. To fully explain the invention, the partitioning of the core of a pressurized water nuclear reactor which is the type liable to have leakage spaces and an apparatus allowing leakage spaces in this partitioning to be eliminated will now be described, as well as an operation for eliminating leakage spaces in the partitioning using the apparatus described.
	summary
051990571	abstract	An image formation-type soft X-ray microscopic apparatus comprises:. a pulse X-ray source for applying X-rays; PA0 a single concave aspherical multilayer film condenser for reflecting the X-rays emitted from the pulse X-ray source so as to condense the X-rays on a sample; PA0 a two-dimensional X-ray imaging element; PA0 a phase zone plate objective optical system for forming a image of the sample on the two-dimensional imaging element by using the X-rays; PA0 an image processing circuit connected to the two-dimensional X-ray imaging element; and PA0 an output circuit connected to the image processing circuit for the purpose of outputting an image of the sample.
062018523	description	DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS Reference should now be made to the drawing figures, on which similar or identical elements are given consistent identifying numerals throughout the various figures thereof, and on which parenthetical references to figure numbers direct the reader to the view(s) on which the element(s) being described is (are) best seen, although the element(s) may also be seen on other views. The present invention provides an apparatus for use with a radiation-blocking liquid and a radiation source. The apparatus includes an attenuation chamber capable of containing a layer of the radiation-blocking liquid, wherein the attenuation chamber is disposed to intercept at least a portion of the radiation emitted from the radiation source, and an adjustment means for selectively metering the thickness of the radiation-blocking liquid layer. Changes in the thickness of the layer alter the amount of radiation transmitted through the attenuation chamber, thereby selectively attenuating at least part of the intercepted radiation. The adjustment means may further include a reservoir capable of containing the radiation-blocking liquid and a siphon connection means for allowing transfer of the radiation-blocking liquid between the reservoir and the attenuation chamber. The thickness of the layer in the attenuation chamber varies in response to changes in elevation of the reservoir. Changes in the thickness of the layer are preferably directly proportional to changes in elevation of the reservoir. More particularly, the thickness of the liquid layer in the attenuation chamber is a function of the difference in elevation between the bottom of the attenuation chamber and the top of the liquid in the reservoir. An increase in the thickness of the liquid layer causes a drop in the radiation transmitted through the attenuation chamber. In a particular embodiment, the adjustment means may further include a pump means, which is preferably automatically controlled, for assisting the flow in the siphon connection means. In a particular embodiment, a substantially linear increase in the thickness of the liquid layer in the attenuation chamber yields a substantially exponential drop in the radiation dose rate transmitted through the attenuation chamber. Preferably, the elevation of the attenuation chamber is substantially fixed and the reservoir is vertically moveable, whereby changes in the radiation dose rate transmitted through the attenuation chamber are a function of changes in the elevation of the reservoir. The adjustment means preferably includes a control means for controlling the movement of the reservoir, thereby providing control of the amount of transmitted radiation, the dose as well as the dose rate of the radiation transmitted through the attenuation chamber may be selectively controlled. The control means may further include means for maintaining at least a minimum liquid thickness in the reservoir, and additionally, means for preventing the liquid level in the reservoir from rising above a maximum liquid height. Preferably, the control means includes a means for specifying a desired dose rate pattern, such as a one-, two-, or three-component exponential dose rate pattern, or another does rate pattern. The adjustment means further preferably includes a movable support means for supporting the reservoir and for adjusting the elevation of the reservoir relative to the attenuation chamber. The movable support means may include a platform and drive means for vertically moving the platform. The reservoir may be attached to the platform by one or more z-axis brackets. A particular embodiment of the drive means includes a shaft connected to the platform, a stepper motor connected to the shaft, and a stepper motor control means for receiving instructions from the control means and for sending motor control signals to the stepper motor. The movable support means would then preferably include a gear reduction means connecting the stepper motor to the shaft. The gear reduction means may comprise a planetary gearbox, for example, a planetary gearbox having an approximate 100 to 1 gear reduction ratio. The shaft may comprise a lead screw. Preferably, the radiation-blocking liquid layer is liquid mercury. Mercury is known to effectively attenuate radiation, so that small changes in the thickness of a layer of liquid mercury may result in a relatively large increment in attenuation. Furthermore, the thickness of the attenuating liquid, as well as changes thereto, may be minimized. Accordingly, the siphon means is at least partially fabricated from a material exhibiting a substantial lack of reactivity with mercury, such as PVC. The siphon means enables very precise control over the metering of the mercury. Other suitable radiation-blocking or radiation absorbing or radiation opaque liquids may also be used, such as another liquid metal or solution or stable suspension of a radiation absorbing or blocking substance, such as an aqueous solution of cesium acetate. When such other suitable radiation-blocking liquids are used, the siphon means is preferably at least partially fabricated from materials exhibiting a substantial lack of reactivity to the radiation-blocking fluid utilized. Such other radiation-blocking fluids and materials which do not substantially react therewith are known in the art. Further preferably, the attenuation chamber and the reservoir are liquid-tight and airtight in order to fully contain the radiation-blocking liquid and any vapors or gases associated therewith. The apparatus may further include a mutual vent means connecting the attenuation chamber and reservoir above respective maximum liquid levels for allowing an equalization of gas pressure therebetween. The mutual vent means may include a vent tube. In a particular embodiment, the vent tube connects the top of the attenuation chamber with the top of the reservoir means. The present invention also contemplates an adjustable irradiator system for use with a radiation-blocking liquid. The system includes a radiation source and a variable attenuator means for intercepting at least a portion of the radiation emitted from the radiation source and for selectively blocking at least a part of the intercepted radiation with the radiation-blocking liquid, wherein the variable attenuator means is capable of transmitting at least a second part of the radiation intercepted from the radiation source. The system is capable of administering a metered dose or dose rate of radiation. Preferably, the system is capable of delivering exponentially varying temporal radiation dose rates. The system preferably includes a target means having at least one target station capable of receiving radiation transmitted through the attenuation chamber. The distance between the target station and the attenuation chamber is preferably adjustable. Thus, the system may include a plurality of spaced apart target stations, wherein each station is disposed a different respective distance away from the attenuation chamber, whereby the target stations are capable of simultaneously receiving different respective radiation rates through the attenuation chamber. The present invention further contemplates, in a particular embodiment, a method for delivering varying temporal radiation dose rates using an adjustable irradiator system, the system comprising a radiation source, a reservoir containing a radiation-blocking liquid, and an attenuation chamber connected to the reservoir by a siphon coupling and disposed in front of the radiation source. Preferably the radiation dose rates are temporally varied exponentially. The method includes the steps of selectively adjusting the elevation of the reservoir relative to the attenuation chamber and allowing the radiation-blocking liquid to seek a common level in the attenuation chamber and in the reservoir. The thickness of the radiation-blocking liquid in the attenuation chamber is thereby selectively adjustable, and changes in the radiation dose rate transmitted through the attenuation chamber are a function of changes in the thickness of the radiation-blocking liquid in the attenuation chamber. In at least one embodiment, a substantially constant rate of change in the liquid level in the reservoir causes a substantially constant rate of change in the liquid level in the attenuation chamber, thereby causing an exponential rate of change in the radiation transmitted or delivered through the attenuation chamber. Thus, an increase in the thickness of the liquid layer in the attenuation chamber causes a decrease in the radiation dose rate transmitted through the attenuation chamber. The method preferably includes exponentially temporally varying the radiation dose rates. A substantially linear change in the thickness of the liquid layer preferably causes a substantially exponential change in the radiation dose rate transmitted through the attenuation chamber. The method may also include maintaining a minimum liquid thickness in the attenuation chamber. The method may also include preventing the level of the liquid in the attenuation chamber from rising above a maximum liquid level. FIGS. 1-4 correspond to a first preferred embodiment of an irradiator system 10 according to the present invention. As seen in FIG. 1, a .sup.137 Cs-irradiator 12 is coupled to a computer controlled variable attenuator 14. The system 10 was designed and constructed to irradiate small animals chronically with dose rate patterns that exactly match those delivered by internal radiochemicals. A first preferred embodiment of the irradiator system 10 has three major components: a .sup.137 Cs irradiator 12, an attenuator 14, and a motion control system 16. The irradiator 12 delivers low dose rates of .sup.137 Cs gamma rays (0.01-30 cGy/h) to animal cages 18 housed below the irradiator 12. The attenuator 14 affords precise control of the dose rate by introducing a layer of highly absorbing mercury between the irradiator 12 and the cages 18. The liquid properties of mercury allow siphoning of the material between a reservoir 20 outside the irradiator 12 and an attenuation chamber 22 mounted between the irradiator 12 and the cages 18. The motion control system 16 is used to raise the reservoir 20 to add mercury to the attenuator chamber 22 (i.e. decrease dose rate) and lower the reservoir 20 to remove mercury from the attenuator chamber 22 (i.e. increase dose rate). The computer-controlled motion control system 16 automatically raises and lowers the mercury reservoir 20 to achieve the desired temporal dose rate pattern. In the first embodiment, a low-dose-rate .sup.137 Cs-irradiator 12 was custom designed for the purpose of chronic irradiation of small animals. A self-contained cabinet-like Model JL-28-8 irradiator (inner dimensions 48".times.9".times.13") as constructed by J. L. Shepherd and Associates (San Fernando, Calif.) was utilized. FIG. 1 shows the interior of irradiator cabinet 24, defining a radiation chamber, with mouse cages 18. The mercury attenuator chamber 22 is just above the top cage 18 and just below the .sup.137 Cs source 12. The water lines 26 for the mouse cages 18 can be seen on the right side. The cages 18 could be placed within the cabinet 24 and irradiated simultaneously, each cage 18 receiving a different dose rate. The irradiator 12 housed an 18 Ci .sup.137 Cs source 28 which provided a beam of 662 keV gamma rays. The beam was passed through a beam shaper to provide a uniform field. Field uniformity at a distance of 20 cm from the beam port is .+-.6% over a 6".times.6" area. The dimensions of the isodose plane increase as the distance from the beam port is increased. Shelves 30 (1/4" Lucite.RTM.) were located within the irradiator system 10 to hold animal cages 18 at different distances below the source 28, thereby providing different dose rates to each cage 18. The source-to-cage distances were capable of being varied, as desired, in 1/4" increments. The irradiator system 10 was also fitted with a day-night timed light, six-outlet flexible water supply line 26, and a ventilation system to continuously replace the air in the cabinet 24. In addition, the irradiator system 10 had an electronic interlock system to prevent opening of the door during periods of irradiation. In order to simulate exponentially decreasing dose rates, an irradiator system 10 was built using the JL-28-8 irradiator 12. The attenuator system 14 included two air-tight cambers, viz. a mercury reservoir 20 and an attenuation chamber 22. The reservoir 20 and attenuation chamber 22 were constructed of 1/2" thick clear polyvinyl chloride (CPVC). Holes were drilled and tapped in the bottom of each chamber 20, 22 and 1/8" nylon NPT elbow fittings inserted. The two chambers 20, 22 were connected with Nalgene.TM. reinforced PVC tubing 32 (3/16" ID) to allow transfer of mercury therebetween. To prevent buildup of air pressure in the chambers 20, 22, an additional NPT fitting was inserted into the side of each chamber and connected with Nalgene.TM. reinforced PVC tubing to serve as a vent. PVC was chosen for its lack of reactivity with mercury. The attenuator chamber 22 was bolted to the inside of the irradiator cabinet 24 between the irradiator 12 and the animal cages 18 and shelves 30, whereas the reservoir 20 was fixed on a computer controlled platform 34. In the absence of air in the mercury transfer line 32, the mercury thickness in the attenuation chamber 22 depends on the vertical position of the mercury reservoir 20. Mercury has a linear attenuation coefficient of about 1.49 cm.sup.-1 for the 662 keV gamma rays of .sup.137 Cs. Therefore, a 4 cm thick layer of mercury can attenuate the beam by a factor of about 200. A linear increase in the mercury thickness yields an exponential drop in the dose rate to each animal cage 18. Therefore, a constant flow rate of mercury into the attenuator chamber 22 provides an exponentially decreasing dose-rate to each cage 18 in the irradiator cabinet 24, the half-time of the decrease in dose-rate being determined by the flow rate of the mercury. Similarly, a constant flow rate out of the attenuator chamber 22 gives an exponentially increasing dose rate. Each cage location in the irradiator receives a different initial dose-rate depending on the distance from the .sup.137 Cs source 28, although the dose-rates in all of the cages 18 vary with the same half-time. If a multicomponent exponential change in the dose-rate is desired, the flow rate of the mercury can be automatically altered using the motion control system 16 described below to accommodate the half-time of each component. Finally, the hard limit switches of the Daedal cross-roller table 34 (described below) were set to ensure a minimum mercury thickness of at least 4 mm in the attenuator chamber 22, which was the minimum thickness required to cover the entire bottom of the chamber 22, and a maximum of mercury thickness of 40 mm to prevent overflow into the vent tube. The vertical position of the mercury reservoir 20 was automatically controlled using a motorized cross-roller table 34. The motorized table 34 included a Daedal (Harrison City, Pa.) Model 106061 C cross-roller table fitted with a Model 04M lead screw (0.4 mm/revolution) and Model 4990-06 z-axis brackets, a Bayside (Port Washington, N.Y.) Model PG60 planetary gearbox 36 with 100:1 ratio, and a Compumotor (Rohnert Park, Calif.) Model 567-102-MO stepper motor 38. The stepper motor 38 was controlled with a Compumotor Zeta series drive (Model 83-135) and a Compumotor AT6200 two-axis stepper controller housed in a Gateway 2000 386SX/20C computer 40. The entire motion control system 16 was powered through an American Power Conversion (APC) Back-UPS 1250 uninterruptable power supply. This high precision system 16, which utilized a 0.4 mm/revolution lead screw and 100:1 gearbox, was capable of changing the mercury thickness in the attenuator 22 by only 2 .mu.m per revolution of the stepper motor 38. In this particular embodiment, software was written in Borland TurboPascal 4.0 to control the motion of the mercury reservoir 20 via computer to provide the desired dose rate pattern, and to execute the planned motion by sending Compumotor 6000 Series commands to the motor 38. The software code accommodated one-, two-, or three-component exponential dose-rate patterns having the forms described below. For a single component exponential, which is capable of being described by the following equation: EQU r=r.sub.o e.sup.-0.693t/T.sup..sub.d , (1) the code requires input of the decrease half-time T.sub.d, i. e. the time required for the dose rate to decrease to one-half its value, ) and the initial dose rate r.sub.o required for cage position 1. As used herein, T.sub.i represents the half-time for dose-rate increase. A two-component exponential dose rate pattern, where there is an initial period of increasing dose rate followed by a period of decreasing dose rate, is capable of being described by the following equation: EQU r=r.sub.o (e.sup.-0.693t/T.sup..sub.d -e.sup.-0.693t/T.sup..sub.i ). (2) In this case, the code requires the extrapolated initial dose rate r.sub.o (12), the increase half-time T.sub.i (time required for dose rate to increase from zero to one-half of r.sub.o), and the decrease half-time T.sub.d. Finally, for a three-component pattern that simulates an increase phase and two decrease phases, the dose rate is capable of being described by the following equation: EQU r=r.sub.o {(ae.sup.-0.693t/T.sup..sub.d1 +(1-a)e.sup.-0.693t/T.sup..sub.d2 )-e.sup.-0.693t/T.sup..sub.i }. (3) The extrapolated initial dose rate r.sub.o, the increase half-time T.sub.i, and the decrease half-times T.sub.d1 and T.sub.d2, as well as the parameter a are required for the code. It should be understood that in addition to the above dose rate profiles (Eqs. 1-3), the code could be modified to accommodate any dose rate pattern, wherein the user may input desired values, or levels, or parameters, or patterns into the control means 40 so as to effect a precisely controlled attenuation of radiation, resulting in a metered radiation dose or dose rate. It should be further understood that the level of radiation blocking liquid in the attenuation chamber may be maintained at discrete or fixed levels for extended periods of time. Thus, the present invention provides a method and means for automatically administering a time-varying or temporally varying dose of radiation. The automated radiation delivery can help reduce the potential for human error. It should be understood that the present invention may comprise a control means which includes accepting user input commands corresponding to a manual override, wherein a preset temporal pattern may be interrupted by, or substituted with, real time manual commands. A Thomson-Nielson Model TN-RD-50 MOSFET dosimeter system was used to measure the absorbed dose-rate at each cage position in the radiation chamber of the cabinet 24 as a function of mercury thickness in the mercury attenuator chamber 22. The MOSFET dosimeters and bias power supply were factory customized to allow measurements at low dose-rates (<1 cGy/h) and low doses (as low as 2 cGy). Low doses could be measured with an accuracy of about 10%, whereas the accuracy of higher doses (>10 cGy) is within 5%. Dose rates were measured with the probes attached to mouse phantoms placed in the 9".times.6".times.6" polycarbonate animals cages 18 (with bedding and wire cage tops). The dosimeter system was also used to monitor the total absorbed dose received by each cage 18 of animals during exposures involving varying dose rates. A mutual vent means 42 which connects the attenuation chamber 22 and the reservoir 20 is preferably provided above respective maximum liquid levels. Thus the vent means 42 allows an equalization of gas pressure between the reservoir 20 and the attenuation chamber 22, thereby facilitating the flow of attenuating liquid therebetween. Furthermore, the vent means 42 allows the system to run as a closed system. For example, if mercury were used as the attenuating liquid, both the liquid and gas or vapor phase of the mercury would be contained substantially within the system, thereby reducing the potential of any unintentional contact with the mercury, whether by the operator, the test subjects or others. In operation, a control means or computer 40 direct stepper motor 38 to turn planetary gear box 36, which thereby raises or lowers table 34. The reservoir 20 thus is raised or lowered to adjust the level of mercury inside the reservoir 20 with respect to the level of mercury residing in the attenuator chamber 22. The layer of mercury in the attenuator chamber 22 attenuates or filters at least part of the radiation emanating from the source 28 of the irradiator 12. Radiation dosages or dose rates incident upon objects or specimens within the irradiator cabinet 24, such as in animal cages 18 or on shelves 30, may be carefully controlled, and in particular, temporally controlled. It should be understood that the present invention is capable of delivering differential doses over a desired period of time. Any time-dosage pattern may be entered into the system. For example, a test subject or patient may be exposed to a high dosage for ten minutes, then to substantially no radiation for three hours, then to two-minute dosages at low levels every hour for six hours. Furthermore, the system 10 may include a sensor means for detecting and/or recording the radiation dosage and/or dosage rate incident upon a given location. The sensor means may be used to track the amount of radiation received by an object or subject, and may also serve as a safety mechanism to prevent over or under exposure to the incident radiation. The sensor means may further be connected to the control means 40, wherein the signal or signals received from the sensor means may be utilized as a feedback signal in control scheme which controls the motion of the reservoir 20, and hence the level of radiation-blocking liquid in the attenuation chamber. Thus, the radiation dosage or dose rate may be adjusted according to a preset pattern which may be further controlled by a real-time feedback control scheme. FIG. 2 illustrates the dose rate as a function of mercury thickness in the attenuator chamber 22 for each cage position. The dose rate was exponentially dependent on the mercury thickness. Least squares fits of the experimental data for each cage position yielded a mean linear attenuation coefficient of 1.22.+-.0.02 cm.sup.-1, which represents the mean slope and standard deviation of the curves shown in FIG. 2. For a mercury density of 13.546 g/cm.sup.3, the mass attenuation coefficient was calculated to be 0.089 cm.sup.2 /g. This value is comparable to the Hubbell's theoretical value for mercury of 0.11 cm.sup.2 /g for 662 keV photons. See Hubbell, Int. J. Appl. Radiat. Isot., 33:1269-1290 (1982), which is incorporated by reference herein in its entirety. FIG. 2 also shows that the dose rate changed by a factor of about 20 from the top cage to the bottom cage regardless of the mercury thickness of the attenuator chamber 22. Hence, depending on the cage location and the mercury thickness in the attenuator chamber 22, dose rates from 0.01 cGy/h to 12 cGy/h can be delivered. Furthermore, the maximum dose rate can be increased to as high as about 30 cGy/h simply by using low-profile (5 cm in height instead of the standard cage height of 15 cm) animal cages 18 which allow the cages to be placed closer to the .sup.137 Cs source 28. To demonstrate the capabilities of the irradiator system 10, a two-component exponential dose rate pattern, corresponding to Equation 2 above, was simulated using a 1 h increase half-time, a 12 h decrease half-time, and an extrapolated initial dose rate r.sub.o of 6.0 cGy/h. FIG. 3 shows the resulting experimental dose rate measurements along with the expected dose rate pattern based on Equation 2, revealing good agreement between the experimental and expected dose rates. The data presented in FIGS. 2 and 3 show that the system 10 is capable of delivering dose rate patterns that are similar to those observed in therapeutic nuclear medicine. Given the strong dependence of biological response on dose rate, such an irradiator system 10 is an invaluable tool to assess the biological effects of exponentially varying dose rates on any given target tissue, which is a largely unexplored area of considerable importance to radioimmunotherapy and other targeted therapies. FIG. 4 is a hypothetical calibration curve for a given decrease half-time T.sub.d and increase half-time T.sub.i. The biological effect is given as a function of the extrapolated initial dose rate r.sub.o delivered by the .sup.137 Cs irradiator 12. To obtain the extrapolated initial dose rate for a given injected activity of a radiochemical having parameters T.sub.e and T.sub.eu, the experimentally determined biological effect can be used in conjunction with the calibration curve as indicated by the dashed lines. With knowledge of r.sub.o, T.sub.e, and T.sub.eu, one can readily calculate the total dose and dose rates at any given time postinjection. Inasmuch as the relative biological effectiveness of .sup.137 Cs 662 keV gamma rays are the same as that of the beta particles emitted by radionuclides relevant to therapeutic nuclear medicine, e.g. .sup.90 Y, .sup.131 I, .sup.32 P, .sup.186 Re, such an irradiator system 10 also offers a unique opportunity to calibrate biological dosimeters for bone marrow dosimetry. Examples of potential biological dosimeters include survival of bone marrow subpopulations (e.g CFU-S, CFU-GM, etc.), induction of micronuclei in lymphocytes or reticulocytes, induction of chromosome aberrations in lymphocytes, and others. Calibration of a biological dosimeter to measure absorbed dose delivered to a target tissue by a given radiochemical can be accomplished generally by the following two steps: 1. Determine dose-rate kinetics in the target tissue for the radiochemical of interest. When the dose rate to the target tissue is principally due to activity within itself (i.e. self-dose rate), the increase and decrease half-times (T.sub.i, T.sub.d) are essentially equal to the experimentally determined effective uptake half-time T.sub.eu and effective clearance half-time T.sub.e of the radioactivity in the tissue. The assumption is generally valid when the primary contribution to the target tissue dose is from particulate radiations (e.g. .sup.32 P, .sup.90 Y, .sup.212 Bi). PA1 2. Using the T.sub.d and T.sub.i established in Step 1, determine the response of the biological dosimeter as a function of extrapolated initial dose rate r.sub.o with the .sup.137 Cs irradiator 12 in system 10 (see FIG. 4). PA1 3. Obtain biological response of tissue following administration of a given activity of the radiochemical. PA1 4. Using the calibration curve based on the response of the tissue to .sup.137 Cs gamma rays delivered with same dose rate pattern, i.e., T.sub.d, T.sub.i (see FIG. 4), the extrapolated initial dose rate r.sub.o to the tissue can be extracted. With knowledge of r.sub.o, T.sub.d, and T.sub.i, the dose rate and cumulated dose to the tissue can be calculated at any time t. Generally, two additional steps are required to utilize the calibrated biological dosimeter to ascertain the extrapolated initial dose rate received by the tissue following administration of a given activity of the radiochemical, as follows: Calibration and implementation of biological dosimeters in this manner provide an effective means of accurately determining the absorbed dose and dose rate pattern received by the target tissue following administration of internal radionuclides that emit low-LET radiations. Biological dosimeters calibrated in this manner, however, are not able to provide information regarding dose and dose rate from internal radionuclides that emit high-LET radiations (e.g. alpha particles, Auger electrons). In these situations, the biological dosimeter would yield a quantity which is the product of the relative biological effectiveness (RBE) and the extrapolated initial dose rate r.sub.o. It should be noted that the irradiator system 10 described above delivers a whole-body dose and, as such, this system is particularly useful for biological dosimetry of sensitive tissues such as bone marrow and gonads. The irradiator system 10 described above utilized a custom-designed .sup.137 Cs small-animal gamma irradiator 12 and a variable attenuator system 14, wherein the irradiator system 10 was capable of delivering chronic exposures of low-linear-energy-transfer (LET) radiation with any desired variable dose rate pattern encountered with internal radionuclides. Thus, the irradiator system 10 could be designed to irradiate animals with exponentially increasing and decreasing dose rate patterns that simulate those encountered during exposure from incorporated radionuclides. The irradiator system 10 can be used to calibrate biological dosimeters, which in turn can serve as an indirect experimental measurement of the absorbed dose. Such experimental measurements of the absorbed dose can be utilized to verify the calculated absorbed doses that are presently relied upon in internal radionuclide dosimetry. In another embodiment of the present invention, an irradiator system is used in conjunction with a means for sensing radiation. The irradiator system may comprise an attenuator system which includes a liquid reservoir and an attenuation chamber, wherein the chamber and the reservoir are connected by tubing in a manner which allows transfer of liquid, such as mercury, therebetween. The attenuator system is disposed between an irradiator and the means for detecting or reading radiation, wherein the attenuation chamber is spaced apart from the radiation reading means to define an irradiation area. In operation, an object is placed in between the attenuation chamber and the reading means while the irradiator is activated. Radiation from the irradiator is filtered or attenuated by the attenuating means, wherein at least a part of the radiation which is not absorbed nor reflected from the attenuation chamber impinges upon the object. The object may in turn reflect or absorb part of the incident radiation, and part of the incident radiation may be transmitted through the object. The radiation reading means may be adapted to receive the radiation transmitted from the attenuation means and through and/or past the object. The radiation means may further filter or process its incident radiation. Thus, for example, radiation impinging upon the radiation reading means may be recorded and/or transmitted for further processing or viewing. In one particular embodiment, the irradiator emits X-rays and the radiation reading means comprises a means for sensing X-rays or a means for exposing film or other recording device which is sensitive to X-rays. In another particular embodiment, the present invention comprises a radiation examination apparatus which includes a radiation source, a detector for detecting radiation originating from the radiation source, and a radiation attenuator disposed between the radiation source and the detector. The attenuator comprises an attenuation chamber capable of containing a layer of a radiation-blocking liquid, an adjustment means for adjusting the thickness of the layer of the radiation-blocking liquid, including a reservoir capable of containing the liquid, and a siphon connection means for allowing the transfer of the liquid between the reservoir and the attenuation chamber. The adjustment means allows for the selective metering of the liquid layer thickness. The thickness of the layer in the attenuation chamber is a function of the difference in elevation between the top of the layer and the attenuation chamber and the top of the liquid in the reservoir. Changes in the thickness of the layer alter the radiation transmitted through the attenuation chamber, wherein the radiation originates from the radiation source. The detector is capable of detecting at least part of the attenuated radiation. FIG. 5 shows a schematic representation of a radiation examination apparatus according to one embodiment of the present invention. Structural elements which are similar to those found in FIG. 1 have been labeled with the same numerals. In addition, detector or reading means 50 is shown disposed at a spaced apart location from the attenuation means 22, wherein an object 100 to be irradiated or examined is placed or transported between the attenuation means 22 and the detector 50. FIG. 6 shows another embodiment of an irradiator system of the present invention, wherein structural elements similar to those of FIG. 1 have been labeled with the same numerals. The irradiator system 10 comprises an attenuation chamber 22 comprising at least one baffle 44 which separates the chamber 22 into two or more sub-chambers. The baffle prevents liquid flow between the sub-chambers. Each subchamber is supplied with a radiation-blocking liquid from its own respective reservoir 20 and motion control system 16. FIG. 6 shows all of the motion control systems 16 for each of the sub-chambers being connected to one control means 40, although each motion control system 16 may be provided with its own control means 40. Preferably the liquid levels in the sub-chambers are controlled in a coordinated fashion, although the liquid level in each sub-chamber may be controlled separately or independently of one or more of the liquid levels in the other sub-chambers. Thus, the radiation emitted from the radiation source may be selectively attenuated spatially, as well as temporally, at any given radiation dosing location or animal cage 18, or portion thereof. In one embodiment, for example, a first subchamber may contain a layer of a first radiation blocking liquid and a second subchamber may contain a second radiation blocking liquid, wherein the second liquid has a greater radiation blocking capability than the first liquid so that the first subchamber may be used for coarse adjustments in attenuation or delivery of radiation and the second subchamber can be used for fine adjustments thereof. FIG. 7 shows yet another embodiment of an irradiator system according to the present invention similar to that shown in FIG. 6 but having at least one generally vertical oriented baffle. Such an embodiment could deliver spatially varied radiation doses in a horizontal plane, for example when different radiation blocking fluids are used and/or when different levels are maintained in different subchambers. In still another embodiment, an irradiator system according to the present invention comprises an attenuation chamber 22 which includes at least one baffle for dividing the attenuation chamber into two or more sub-chambers wherein two or more sub-chambers are connected to a common reservoir. In yet another particular embodiment, the present invention comprises a filter for use with an X-ray examination apparatus. The examination apparatus comprises an X-ray source and an X-ray detector for detecting X-rays originating from the X-ray source. The filter comprises an attenuation chamber capable of containing a layer of radiation-blocking liquid, an adjustment means for adjusting the thickness of the layer of the radiation-blocking liquid, a reservoir capable of containing the liquid, and a siphon connection means for allowing the transfer of the radiation-blocking liquid between the reservoir and the attenuation chamber. The thickness of the layer in the attenuation chamber is a function of the difference in elevation between the top of the layer in the attenuation chamber and the top of the liquid in the reservoir. Changes in the thickness of the layer alter the radiation transmitted through the attenuation chamber. Thus, the filter may be used to selectively meter the amount of radiation reaching an object which passes through the X-ray examination apparatus. The object may be subjected to a temporally varying dose of radiation. Alternately, or in addition, the object may be subject to one or more discrete levels of radiation. In another particular embodiment, the present invention comprises a filter for use with an X-ray examination apparatus, such as that typically found in airports and other areas of security checking. The present invention also contemplates an irradiating system which is used in therapeutic treatment applications, such as those associated with humans, animals, or plants. The present invention further contemplates attenuation and/or delivery of radiation in the preparation and/or treatment of food stuffs. Most preferably, the adjustment means for selectively metering the thickness of a radiation-blocking layer comprises an attenuation chamber and a reservoir connected by a siphon means. It has been found that precise and repeatable control over the layer thickness can be achieved by such means or method. However, the adjustment means may alternately comprise a pump means for controlling the flows into and out of, and therefore the level of liquid in, the attenuation chamber, although precision, repeatability and/or reproducibility may not approach that achievable by the above-described embodiments. Furthermore, a pump means may be used to assist or enhance the control of the liquid level in the attenuation chamber, in conjunction with, or in parallel with, the siphon connection means. For example, a pump-assisted connection means between the attenuation chamber and the reservoir, which may include valve means and connections to the control means, may be provided in parallel with a siphon connection means to speed the addition and/or removal of the liquid from the attenuation chamber. For example, the pump means may be activated when rapid filling or emptying of the attenuation chamber is desired. Furthermore, the attenuation chamber may be provided with one or more liquid level sensors to assist in the control of the liquid level and/or the calibration of the apparatus. Preferably the attenuation chamber is adapted to possess a planar internal bottom surface which supports the radiation blocking liquid. The attenuation chamber may instead be provided with a non-planar bottom which would be necessary to achieve a desired dispersion or intensity of radiation. Preferably, the internal surfaces of the attenuation chamber that support the liquid are fixed or rigid. The present invention may be used with either ionizing radiation, such as neutrons or protons, or nonionizing radiation, such as visible light, infrared or ultraviolet radiation. Typically a suitable radiation blocking liquid would be selected which is appropriate for the type of radiation to be attenuated and the desired range of attenuation. For example, a boron rich material may be used (instead of mercury) to attenuate neutron radiation. By way of another example, light intensity may be attenuated by an opaque liquid. By way of further example, aqueous solutions of a heavy metal salt, such as cesium acetate, may be used as an attenuating liquid. The present invention may further comprise filtering and/or focusing radiation passing through the attenuation means. It is to be understood that the invention is not limited to the illustrations described and shown herein, which are deemed to be merely illustrative of the best modes of carrying out the invention, and which are susceptible of modification of form, size, arrangement of parts and details of operation. The invention rather is intended to encompass all such modifications which are within its spirit and scope as defined by the claims. It will thus be seen that the objects set forth above, among those elucidated in, or made apparent from, the preceding description, are efficiently attained and, since certain changes may be made in the above construction without departing from the scope of the invention, it is intended that all matter contained in the above description or shown on the accompanying drawing figures shall be interpreted as illustrative only and not in a limiting sense. It is also to be understood that the following claims are intended to cover all of the generic and specific features of the invention herein described and all statements of the scope of the invention which, as a matter of language, might be said to fall therebetween.
	abstract	Disclosed is a novel scanning-probe type atomic force microscope wherein false deflection of the probe is reduced. The probe of the scanning-probe type atomic force microscope moves in both the horizontal direction and the vertical direction during the scanning, while the sample is kept in order to reduce the false deflection brought to the probe due to the scanning motion, two approaches are adopted. The first is to have a focused laser spot tracking an invariant point on the probe's cantilever, which moves three-dimensionally during the scanning. The second approach is to have the laser beam, which is reflected from the moving cantilever, hitting an invariant point of the PSD, when the sample is distanced from the probe and induces no deflection. A beam tracking system wherein the scanning probe is located approximately at the focal point of an objective lens and he optical system including a laser source, an optical module, a feedback module and the probe are driven by an approach mechanism to move in synchronization.
	description	Embodiment 1 A soft X-ray reduction projection exposure system and a soft X-ray reduction projection exposure method according to Embodiment 1 of the invention will now be described with reference to FIG. 1. FIG. 1 is a rough cross-sectional view of the soft X-ray reduction projection exposure system of Embodiment 1, which includes a first chamber 110, a second chamber 120 and a third chamber 130 communicating with one another. The first chamber 110 includes a discharge type X-ray source 111 for generating a soft X-ray beam, an illumination optical system 112 for transmitting the soft X-ray beam generated by the discharge type X-ray source 111 to the second chamber 120, and a first diffusion pump 113 for reducing the pressure within the first chamber 110. The second chamber 120 includes a reflecting mask 121 on which a desired pattern is formed, a mask stage 122 for holding the reflecting mask 121, a reflecting optical system 123 for introducing the soft X-ray beam having been transmitted from the illumination optical system 112 of the first chamber 110 to the reflecting mask 121, a reduction projection optical system 124 for reducing the soft X-ray beam having been reflected by the reflecting mask 121 and transmitting the reduced soft X-ray beam to the third chamber 130, and a second diffusion pump 125 for reducing the pressure within the second chamber 120. The third chamber 130 includes a wafer 131 on which a pattern is to be formed, a wafer stage 132 for holding the wafer 131, and a third diffusion pump 133 for reducing the pressure within the third chamber 130. The soft X-ray beam transmitted from the reduction projection optical system 124 of the second chamber 120 to the third chamber 130 irradiate the surface of the wafer 131. The total pressure within the first chamber 110 can be controlled down to 1.33xc3x9710xe2x88x928 Pa by the first diffusion pump 113, the total pressure within the second chamber 120 can be controlled down to 1.33xc3x9710xe2x88x928 Pa by the second diffusion pump 125 and the total pressure within the third chamber 130 can be controlled down to 1.33xc3x9710xe2x88x928 Pa by the third diffusion pump 133. As a characteristic of Embodiment 1, the partial pressure of a carbon compound gas within the first chamber 110 is controlled to be 1.33xc3x9710xe2x88x928 Pa or less by the first diffusion pump 113, the partial pressures of the carbon compound gas within a region where the reflecting mask 121 is disposed and a region where the reduction projection optical system 124 is disposed in the second chamber 120 are respectively controlled to be 1.33xc3x9710xe2x88x928 Pa or less by the second diffusion pump 125, and the partial pressure of the carbon compound gas within the third chamber 130 is controlled to be 1.33xc3x9710xe2x88x928 Pa or less by the third diffusion pump 133. FIG. 2 shows the relationship between the partial pressure of a hydrocarbon (CxHy) gas in the vicinity of the reduction projection optical system 124 and the thickness of a carbon film adhered onto a reflecting face of the reduction projection optical system 124 obtained by introducing the soft X-ray beam having been reflected by the reflecting mask 121 to the reduction projection optical system 124 continuously for 24 hours in the soft X-ray reduction projection exposure system of FIG. 1. In FIG. 2, points shown with xcex94 indicate actually measured values and a solid line is a virtual line obtained on the basis of these actually measured values. As is obvious from FIG. 2, in the case where the partial pressure of the hydrocarbon gas is higher than 1.33xc3x9710xe2x88x928 Pa, the thickness of the carbon film deposited on the surface of the reduction projection optical system 124 through the irradiation with the soft X-ray beam is abruptly increased. In contrast, in the case where the partial pressure of the hydrocarbon gas is 1.33xc3x9710xe2x88x928 Pa or less, the thickness of the deposited carbon film is approximately 0.1 nm. The thickness of the deposited carbon film largely depends upon the partial pressure of the hydrocarbon gas. In the case where the partial pressure of the hydrocarbon gas is higher than 1.33xc3x9710xe2x88x928 Pa, a degree of carbon adhering onto the surface of the reduction projection optical system 124 is higher than a degree of carbon releasing from the surface of the reduction projection optical system 124, so that the thickness of the carbon film can be increased. In contrast, in the case where the partial pressure of the hydrocarbon gas is 1.33xc3x9710xe2x88x928 Pa or less, the degree of carbon releasing from the reduction projection optical system 124 is higher than the degree of carbon adhering onto the reduction projection optical system 124, so that the thickness of the carbon film cannot be increased. In other words, in the case where the partial pressure of the hydrocarbon gas is 1.33xc3x9710xe2x88x928 Pa or less, the thickness of the carbon film does not exceed 0.1 nm. Also, when the thickness of the carbon film is approximately 0.1 nm, lowering of the reflectance of the reduction projection optical system 124 can be suppressed to approximately 0.1%, which does not cause any practical problem. According to Embodiment 1 of the invention, the thickness of the carbon film deposited on the reflecting face of the reduction projection optical system 124 can be suppressed to approximately 0.1 nm because the partial pressure of the hydrocarbon gas is controlled to be 1.33xc3x9710xe2x88x928 Pa or less, and therefore, the optical characteristic can be prevented from degrading due to contamination of the reflecting face of the reduction projection optical system 124 with an organic substance. In this embodiment, the relationship between the partial pressure of the hydrocarbon gas within the region where the reduction projection optical system 124 is disposed in the second chamber 120 and the thickness of the carbon film deposited on the surface of the reduction projection optical system 124 is described. This relationship also holds between the partial pressure of the hydrocarbon gas within the region where the illumination optical system 112 is disposed in the first chamber 110 and the thickness of a carbon film deposited on the surface of the illumination optical system 112 and between the partial pressure of the hydrocarbon gas within the region where the reflecting mask 121 is disposed in the second chamber 120 and the thickness of a carbon film deposited on the surface of the reflecting mask 121. Specifically, the partial pressure of the hydrocarbon gas within the region where the illumination optical system 112 is disposed in the first chamber 110 and the partial pressure of the hydrocarbon gas within the region where the reflecting mask 121 is disposed in the second chamber 120 are respectively controlled to be 1.33xc3x9710xe2x88x928 Pa or less, and hence, the thicknesses of the carbon films deposited on the reflecting faces of the illumination optical system 112 and the reflecting mask 121 can be suppressed to approximately 0.1 nm. In Embodiment 1, the partial pressure of the hydrocarbon gas is controlled to be 1.33xc3x9710xe2x88x928 Pa or less. Alternatively, the partial pressure of a gas of any of hydrocarbons such as methane, ethane and propane, straight-chain organic substances such as isopropyl alcohol and polymethyl methacrylate, and cyclic organic substances such as benzene and phthalate may be controlled in order to suppress the thickness of the carbon film to approximately 0.1 nm. Although the partial pressure of the hydrocarbon gas is controlled to be 1.33xc3x9710xe2x88x928 Pa or less in all of the inside region of the first chamber 110 and the regions where the reflecting mask 121 and the reduction projection optical system 124 are respectively disposed in the second chamber 120 in Embodiment 1, the partial pressure of the hydrocarbon gas may be controlled to be 1.33xc3x9710xe2x88x928 Pa or less in at least one of these regions. As a characteristic of Embodiment 1, the total pressure within the first chamber 110 is controlled to be 1.33xc3x9710xe2x88x924 Pa or less by the first diffusion pump 113 and the total pressures within the regions where the reflecting mask 121 and the reduction projection optical system 124 are respectively disposed in the second chamber 120 are controlled to be 1.33xc3x9710xe2x88x924 Pa by the second diffusion pump 125. Since the total pressure within the first chamber 110 and the total pressures within the regions where the reflecting mask 121 and the reduction projection optical system 124 are respectively disposed in the second chamber 120 are thus controlled to be 1.33xc3x9710xe2x88x924 Pa or less, any gas other than the hydrocarbon gas, such as a gas of an inorganic substance like a metal element, can be suppressed. Therefore, the optical characteristics can be prevented from degrading due to the contamination, with an inorganic substance, of the reflecting faces of the illumination optical system 112, the reflecting mask 121 and the reduction projection optical system 124. In Embodiment 1, the total pressure and the partial pressure of the hydrocarbon gas within the first chamber 110 are controlled by the first diffusion pump 113, and the total pressure within the second chamber 120 and the partial pressures of the hydrocarbon gas within the regions where the reflecting mask 121 and the reduction projection optical system 124 are respectively disposed are controlled by the second diffusion pump 125. These diffusion pumps may be provided to the respective chambers or the respective regions, or a common diffusion pump may be provided to a plurality of chambers or a plurality of regions. In particular, the partial pressure of the hydrocarbon gas within the region of the reflecting mask 121 is disposed in the second chamber 120 and the partial pressure of the hydrocarbon gas within the region where the reduction projection optical system 124 is disposed in the second chamber 120 are both controlled by the second diffusion pump 125 in Embodiment 1. However, the partial pressures of the hydrocarbon gas within the region where the reflecting mask 121 is disposed and within the region where the reduction projection optical system 124 is disposed are preferably individually controlled by different diffusion pumps. Although the discharge type X-ray source 111 is used as the soft X-ray source in Embodiment 1, another soft X-ray source such as a laser induced plasma X-ray source may be used instead. Also, although the reflecting mirrors are used as the illumination optical system 112 and the reduction projection optical system 124 in Embodiment 1, another means may be employed instead. Furthermore, although the diffusion pumps are used as pressure reducing means in Embodiment 1, another vacuum pumping device such as a turbo pump or an ion pump may be used instead. Embodiment 2 A soft X-ray reduction projection exposure system and a soft X-ray reduction projection exposure method according to Embodiment 2 of the invention will now be described with reference to FIG. 3. FIG. 3 is a rough cross-sectional view of the soft X-ray reduction projection exposure system of Embodiment 2, which includes a first chamber 210, a second chamber 220 and a third chamber 230 communicating with one another. The first chamber 210 includes a discharge type X-ray source 211 for generating a soft X-ray beam, an illumination optical system 212 for transmitting the soft X-ray beam generated by the discharge type X-ray source 211 to the second chamber 220, a first diffusion pump 213 for reducing the pressure within the first chamber 210, and a first organic substance trap 215 disposed in a first pressure reducing path 214 connecting the first chamber 210 to the first diffusion pump 213. The second chamber 220 includes a reflecting mask 221 on which a desired pattern is formed, a mask stage 222 for holding the reflecting mask 221, a reflecting optical system 223 for introducing the soft X-ray beam having been transmitted from the illumination optical system 212 of the first chamber 210 to the reflecting mask 221, a reduction projection optical system 224 for reducing the soft X-ray beam having been reflected by the reflecting mask 221 and transmitting the reduced soft X-ray beam to the third chamber 230, a second diffusion pump 225 for reducing the pressure within the second chamber 220, and a second organic substance trap 227 disposed in a second pressure reducing path 216 connecting the second chamber 220 to the second diffusion pump 225. The third chamber 230 includes a wafer 231 on which a pattern is to be formed, a wafer stage 232 for holding the wafer 231, and a third diffusion pump 233 for reducing the pressure within the third chamber 230. The soft X-ray beam transmitted from the reduction projection optical system 224 of the second chamber 220 to the third chamber 230 irradiate the surface of the wafer 231. The first organic substance trap 215 and the second organic substance trap 227 are both cooled with liquid helium and respectively include filters for capturing a carbon compound generated in the first chamber 210 and the second chamber 220. The total pressure within the first chamber 210 can be controlled down to 1.33xc3x9710xe2x88x928 Pa by the first diffusion pump 213, the total pressure within the second chamber 220 can be controlled down to 1.33xc3x9710xe2x88x928 Pa by the second diffusion pump 225 and the total pressure within the third chamber 230 can be controlled down to 1.33xc3x9710xe2x88x928 Pa by the third diffusion pump 233. Similarly to Embodiment 1, the partial pressure of a carbon compound gas within the first chamber 210 is controlled to be 1.33xc3x9710xe2x88x928 Pa or less by the first diffusion pump 213, the partial pressures of the carbon compound gas within a region where the reflecting mask 221 is disposed and a region where the reduction projection optical system 224 is disposed in the second chamber 220 are respectively controlled to be 1.33xc3x9710xe2x88x928 Pa or less by the second diffusion pump 225, and the partial pressure of the carbon compound gas within the third chamber 230 is controlled to be 1.33xc3x9710xe2x88x928 Pa or less by the third diffusion pump 233. As a characteristic of Embodiment 2, the first organic substance trap 215 is disposed in the first pressure reducing path 214 connecting the first chamber 210 to the first diffusion pump 213 and the second organic substance trap 227 is disposed in the second pressure reducing path 216 connecting the second chamber 220 to the second diffusion pump 225. Therefore, the partial pressure of the carbon compound gas within the first chamber 210 can be rapidly controlled to be 1.33xc3x9710xe2x88x928 Pa or less, and the partial pressures of the carbon compound gas within the regions where the reflecting mask 221 and the reduction projection optical system 224 are disposed in the second chamber 220 can be rapidly controlled to be 1.33xc3x9710xe2x88x928 Pa or less. Therefore, according to Embodiment 2, the thickness of a carbon film deposited on the surface of the illumination optical system 212, the thickness of a carbon film deposited on the surface of the reflecting mask 221 and the thickness of a carbon film deposited on the surface of the reduction projection optical system 224 can be all suppressed to approximately 0.1 nm. Accordingly, the optical characteristics can be prevented from degrading due to the contamination, with an organic substance, of the reflecting faces of the illumination optical system 212, the reflecting mask 221 and the reduction projection optical system 224. In Embodiment 2, the partial pressure of the carbon compound gas is controlled to be 1.33xc3x9710xe2x88x928 Pa or less in the inside region of the first chamber 210, the region where the reflecting mask 221 is disposed in the second chamber 220 and the region where the reduction projection optical system 224 is disposed in the second chamber 220. Instead, the partial pressure of the carbon compound gas may be higher than 1.33xc3x9710xe2x88x928 Pa in the inside region of the first chamber 210, the region where the reflecting mask 221 is disposed in the second chamber 220 and the region where the reduction projection optical system 224 is disposed in the second chamber 220. In this case, a carbon compound generated in the inside region of the first chamber 210 is captured by the first organic substance trap 215, and a carbon compound generated in the regions where the reflecting mask 221 and the reduction projection optical system 224 are disposed in the second chamber 220 is captured by the second organic substance trap 227. Thus, the carbon compound is captured by the first organic substance trap 215 and the second organic substance trap 227, and hence, the thicknesses of the carbon films deposited on the surfaces of the illumination optical system 212, the reflecting mask 221 and the reduction projection optical system 224 can be reduced. Therefore, the optical characteristics can be prevented from degrading due to the contamination, with the organic substance, of the illumination optical system 212, the reflecting mask 221 and the reduction projection optical system 224. Although the filters cooled with liquid helium are used in the first organic substance trap 215 and the second organic substance trap 227 in Embodiment 2, any other device capable of capturing an organic substance, such as a filter cooled with liquid nitrogen, may be used instead. Also, the first organic substance trap 215 is disposed between the first chamber 210 and the first diffusion pump 213 and the second organic substance trap 227 is disposed between the second chamber 220 and the second diffusion pump 225 in Embodiment 2. Alternatively, the first organic substance trap 215 may be disposed in a branch path branching from the first pressure reducing path 214 and the second organic substance trap 227 may be disposed in a branch path branching from the second pressure reducing path 216. Furthermore, the partial pressure of the carbon compound gas is controlled to be 1.33xc3x9710xe2x88x928 Pa or less and hydrocarbon is captured in Embodiment 2. The carbon compound may be any of hydrocarbons such as methane, ethane and propane, straight-chain organic substances such as isopropyl alcohol and polymethyl methacrylate, and cyclic organic substances such as benzene and phthalate. The thickness of the carbon film can be suppressed to approximately 0.1 nm by controlling the partial pressure of the hydrocarbon, the straight-chain organic substance or the cyclic organic substance, and the thickness of the carbon film can be suppressed by capturing the hydrocarbon, the straight-chain organic substance or the cyclic organic substance. Also, the partial pressure of the carbon compound gas is controlled to be 1.33xc3x9710xe2x88x928 Pa or less in all of the inside region of the first chamber 210, the region where the reflecting mask 221 is disposed in the second chamber 220 and the region where the reduction projection optical system 224 is disposed in the second chamber 220 in Embodiment 2. Alternatively, the partial pressure of the carbon compound gas may be controlled to be 1.33xc3x9710xe2x88x928 Pa or less at least in one of these regions. Also in Embodiment 2, similarly to Embodiment 1, the total pressure within the first chamber 210 is preferably controlled to be 1.33xc3x9710xe2x88x924 Pa or less by the first diffusion pump 213 and the total pressure within the second chamber 220 is preferably controlled to be 1.33xc3x9710xe2x88x924 Pa or less by the second diffusion pump 225. Thus, the optical characteristics can be prevented from degrading due to contamination, with an inorganic substance such as a metal element, of the reflecting faces of the illumination optical system 212, the reflecting mask 221 and the reduction projection optical system 224. In Embodiment 2, the total pressure and the partial pressure of the hydrocarbon gas within the first chamber 210 are controlled by the first diffusion pump 213, and the total pressure within the second chamber 220 and the partial pressures of the hydrocarbon gas within the regions where the reflecting mask 221 and the reduction projection optical system 224 are disposed are controlled by the second diffusion pump 225. These diffusion pumps may be provided to the respective chambers or the respective regions, or a common diffusion pump may be provided to a plurality of chambers or a plurality of regions. In particular, the partial pressure of the hydrocarbon gas within the region where the reflecting mask 221 is disposed in the second chamber 220 and the partial pressure of the hydrocarbon gas within the region where the reduction projection optical system 224 is disposed in the second chamber 220 are both controlled by the second diffusion pump 225 in Embodiment 2. However, the partial pressures of the hydrocarbon gas within the region where the reflecting mask 221 is disposed and within the region where the reduction projection optical system 224 is disposed are preferably individually controlled by different diffusion pumps. Furthermore, the carbon compound generated in the region where the reflecting mask 221 is disposed in the second chamber 220 and the carbon compound generated in the region where the reduction projection optical system 224 is disposed in the second chamber 220 are captured preferably by different organic substance traps. Although the discharge type X-ray source 211 is used as the soft X-ray source in Embodiment 2, another soft X-ray source such as a laser induced plasma X-ray source may be used instead. Also, although the reflecting mirrors are used as the illumination optical system 212 and the reduction projection optical system 224 in Embodiment 2, another means may be employed instead. Furthermore, although the diffusion pumps are used as pressure reducing means in Embodiment 2, another vacuum pumping device such as a turbo pump or an ion pump may be used instead. Embodiment 3 A soft X-ray reduction projection exposure system and a soft X-ray reduction projection exposure method according to Embodiment 3 of the invention will now be described with reference to FIG. 4. FIG. 4 is a rough cross-sectional view of the soft X-ray reduction projection exposure system of Embodiment 3, which includes a first chamber 310, a second chamber 320 and a third chamber 330 communicating with one another. The first chamber 310 includes a discharge type X-ray source 311 for generating a soft X-ray beam, an illumination optical system 312 for transmitting the soft X-ray beam generated by the discharge type X-ray source 311 to the second chamber 320, a first diffusion pump 313 for reducing the pressure within the first chamber 310, and a first oxygen cylinder 315 serving as first oxygen gas supply means for supplying an oxygen gas not ionized to the first chamber 310 through a first massflow controller 314 for controlling the flow rate of the oxygen gas. The second chamber 320 includes a reflecting mask 321 on which a desired pattern is formed, a mask stage 322 for holding the reflecting mask 321, a reflecting optical system 323 for introducing the soft X-ray beam having been transmitted from the illumination optical system 312 of the first chamber 310 to the reflecting mask 321, a reduction projection optical system 324 for reducing the soft X-ray beam having been reflected by the reflecting mask 321 and transmitting the reduced soft X-ray beam to the third chamber 330, a second diffusion pump 325 for reducing the pressure within the second chamber 320, and a second oxygen cylinder 327 serving as second oxygen gas supply means for supplying an oxygen gas not ionized to the second chamber 320 through a second massflow controller 326 for controlling the flow rate of the oxygen gas. The third chamber 330 includes a wafer 331 on which a pattern is to be formed, a wafer stage 332 for holding the wafer 331, and a third diffusion pump 333 for reducing the pressure within the third chamber 330. The soft X-ray beam transmitted from the reduction projection optical system 324 of the second chamber 320 to the third chamber 330 irradiate the surface of the wafer 331. As a characteristic of Embodiment 3, the partial pressure of the oxygen gas within the first chamber 310 is controlled to be 1.33xc3x9710xe2x88x924 Pa through 1.33xc3x9710xe2x88x921 Pa by the first diffusion pump 313 and the first massflow controller 314, and the partial pressures of the oxygen gas within regions where the reflecting mask 321 and the reduction projection optical system 324 are disposed in the second chamber 320 are controlled to be 1.33xc3x9710xe2x88x924 Pa through 1.33xc3x9710xe2x88x921 Pa by the second diffusion pump 325 and the second massflow controller 326. FIG. 5 shows the relationship between the partial pressure of the oxygen gas in the vicinity of the reduction projection optical system 324, the thickness of a carbon film adhered onto a reflecting face of the reduction projection optical system 324 and the transmittance loss per meter of the soft X-ray beam obtained by introducing the soft X-ray beam having been reflected by the reflecting mask 321 to the reduction projection optical system 324 continuously for 24 hours in the soft X-ray reduction projection exposure system of FIG. 4. In FIG. 5, points shown with xe2x96xa1 indicate actually measured values of the thickness of the carbon film and a solid line is a virtual line obtained on the basis of these actually measured values. Also, points shown with xcex94 indicate actually measured values of the transmittance loss and a broken line is a virtual line obtained on the basis of these actually measured values. As is obvious from FIG. 5, in the case where the partial pressure of the oxygen gas is lower than 1.33xc3x9710xe2x88x924 Pa, the thickness of the carbon film deposited on the surface of the reduction projection optical system 324 through the irradiation with the soft X-ray beam is abruptly increased. In contrast, in the case where the partial pressure of the oxygen gas exceeds 1.33xc3x9710xe2x88x924 Pa, the thickness of the deposited carbon film is approximately 0.1 nm. On the surface of the reduction projection optical system 324, a reaction in which the absorbed carbon film is decomposed by oxygen atoms activated through the irradiation with the soft X-ray beam always occurs. Accordingly, the thickness of the carbon film deposited on the surface of the reduction projection optical system 324 largely depends upon the partial pressure of the oxygen gas. In the case where the partial pressure of the oxygen gas is 1.33xc3x9710xe2x88x924 Pa or more, a degree of decomposing carbon is higher than a degree of carbon depositing on the surface of the reduction projection optical system 324, so that the thickness of the carbon film cannot increase. Although the thickness of the carbon film is reduced as the partial pressure of the oxygen gas is higher, when the partial pressure of the oxygen gas is too high, a light absorbing function of the oxygen molecules becomes too large to neglect, and hence, the transmittance loss occurs. When the transmittance loss per meter of the soft X-ray beam exceeds 1%, the proportion of the soft X-ray beam generated by the discharge type X-ray source 311 to reach the surface of the wafer 331 is disadvantageously lowered. As is understood from FIG. 5, in the case where the partial pressure of the oxygen gas is 1.33xc3x9710xe2x88x921 Pa, the transmittance loss is 1%. Therefore, it seems that the upper limit of the partial pressure of the oxygen gas is 1.33xc3x9710xe2x88x921 Pa. For this reason, the partial pressure of the oxygen gas is controlled to be 1.33xc3x9710xe2x88x924 Pa through 1.33xc3x9710xe2x88x921 Pa in Embodiment 3. Therefore, the thickness of the carbon film deposited on the reflecting face of the reduction projection optical system 324 can be suppressed to approximately 0.1 nm without increasing the transmittance loss of the soft X-ray beam. Accordingly, the optical characteristic can be prevented from degrading due to the contamination of the reflecting face of the reduction projection optical system 324 with an organic substance. In this embodiment, the relationship between the partial pressure of the oxygen gas within the region where the reduction projection optical system 324 is disposed in the second chamber 320 and the thickness of the carbon film deposited on the surface of the reduction projection optical system 324 is described. This relationship also holds between the partial pressure of the oxygen gas within a region where the illumination optical system 312 is disposed in the first chamber 310 and the thickness of a carbon film deposited on the surface of the illumination optical system 312 and between the partial pressure of the oxygen gas within the region where the reflecting mask 321 is disposed in the second chamber 320 and the thickness of a carbon film deposited on the surface of the reflecting mask 321. Specifically, when the partial pressure of the oxygen gas within the region where the illumination optical system 312 is disposed in the first chamber 310 and the partial pressure of the oxygen gas within the region where the reflecting mask 321 is disposed in the second chamber 320 are respectively controlled to be 1.33xc3x9710xe2x88x924 Pa through 1.33xc3x9710xe2x88x921 Pa, the thicknesses of the carbon films deposited on the reflecting faces of the illumination optical system 312 and the reflecting mask 321 can be suppressed to approximately 0.1 nm without increasing the transmittance loss of the soft X-ray beam. Also, although the oxygen gas not ionized is introduced in Embodiment 3, an ionized oxygen gas may be introduced instead. However, the oxygen gas not ionized is preferably introduced because the surface of the reflecting mirror is minimally damaged. Although the partial pressure of the oxygen gas is controlled to be 1.33xc3x9710xe2x88x924 Pa through 1.33xc3x9710xe2x88x921 Pa in all of the inside region of the first chamber 310 and the regions where the reflecting mask 321 and the reduction projection optical system 324 are disposed in the second chamber 320 in Embodiment 3, the partial pressure of the oxygen gas may be controlled to be 1.33xc3x9710xe2x88x924 Pa through 1.33xc3x9710xe2x88x921 Pa in at least one of these regions. As a characteristic of Embodiment 3, the total pressure within the first chamber 310 and the total pressures within the regions where the reflecting mask 321 and the reduction projection optical system 324 are disposed in the second chamber 320 are controlled to be 1.33xc3x9710xe2x88x921 Pa or less. Thus, any gas other than the oxygen gas, such as a gas of an inorganic substance like a metal element, can be suppressed. Therefore, the optical characteristics can be prevented from degrading due to the contamination, with an inorganic substance, of the reflecting faces of the illumination optical system 312, the reflecting mask 321 and the reduction projection optical system 324. In Embodiment 3, the partial pressure of the oxygen gas within the first chamber 310 is controlled by the first diffusion pump 313, and the partial pressures of the oxygen gas within the regions where the reflecting mask 321 and the reduction projection optical system 324 are disposed in the second chamber 320 are controlled by the second diffusion pump 325. These diffusion pumps may be provided to the respective chambers or the respective regions, or a common diffusion pump may be provided to a plurality of chambers or a plurality of regions. In particular, the partial pressure of the oxygen gas within the region of the reflecting mask 321 is disposed in the second chamber 320 and the partial pressure of the oxygen gas within the region where the reduction projection optical system 324 is disposed in the second chamber 320 are both controlled by the second diffusion pump 325 in Embodiment 3. However, the partial pressures of the oxygen gas within the region where the reflecting mask 321 is disposed and within the region where the reduction projection optical system 324 is disposed are preferably individually controlled by different diffusion pumps. Although the discharge type X-ray source 311 is used as the soft X-ray source in Embodiment 3, another soft X-ray source such as a laser induced plasma X-ray source may be used instead. Also, although the reflecting mirrors are used as the illumination optical system 312 and the reduction projection optical system 324 in Embodiment 3, another means may be employed instead. Furthermore, although the diffusion pumps are used as pressure reducing means in Embodiment 3, another vacuum pumping device such as a turbo pump or an ion pump may be used instead. Embodiment 4 A soft X-ray reduction projection exposure system and a soft X-ray reduction projection exposure method according to Embodiment 4 of the invention will now be described with reference to FIG. 6. FIG. 6 is a rough cross-sectional view of the soft X-ray reduction projection exposure system of Embodiment 4, which includes a first chamber 410, a second chamber 420 and a third chamber 430 communicating with one another. The first chamber 410 includes a discharge type X-ray source 411 for generating a soft X-ray beam, an illumination optical system 412 for transmitting the soft X-ray beam generated by the discharge type X-ray source 411 to the second chamber 420, a first diffusion pump 413 for reducing the pressure within the first chamber 410, and a first ozone cylinder 415 serving as first ozone gas supply means for supplying an ozone gas to the first chamber 410 through a first massflow controller 414 for controlling the flow rate of the ozone gas. The second chamber 420 includes a reflecting mask 421 on which a desired pattern is formed, a mask stage 422 for holding the reflecting mask 421, a reflecting optical system 423 for introducing the soft X-ray beam having been transmitted from the illumination optical system 412 of the first chamber 410 to the reflecting mask 421, a reduction projection optical system 424 for reducing the soft X-ray beam having been reflected by the reflecting mask 421 and transmitting the reduced soft X-ray beam to the third chamber 430, a second diffusion pump 425 for reducing the pressure within the second chamber 420, and a second ozone cylinder 427 serving as second ozone gas supply means for supplying an ozone gas to the second chamber 420 through a second massflow controller 426 for controlling the flow rate of the ozone gas. The third chamber 430 includes a wafer 431 on which a pattern is to be formed, a wafer stage 432 for holding the wafer 431, and a third diffusion pump 433 for reducing the pressure within the third chamber 430. The soft X-ray beam transmitted from the reduction projection optical system 424 of the second chamber 420 to the third chamber 430 irradiate the surface of the wafer 431. As a characteristic of Embodiment 4, the partial pressure of the ozone gas within the first chamber 410 is controlled to be 1.33xc3x9710xe2x88x924 Pa through 4.00xc3x9710xe2x88x922 Pa by the first diffusion pump 413 and the first massflow controller 414, and the partial pressures of the ozone gas within regions where the reflecting mask 421 and the reduction projection optical system 424 are disposed in the second chamber 420 are controlled to be 1.33xc3x9710xe2x88x924 Pa through 4.00xc3x9710xe2x88x922 Pa by the second diffusion pump 425 and the second massflow controller 426. FIG. 7 shows the relationship between the partial pressure of the ozone gas in the vicinity of the reduction projection optical system 424, the thickness of a carbon film adhered onto a reflecting face of the reduction projection optical system 424 and the transmittance loss per meter of the soft X-ray beam obtained by introducing the soft X-ray beam having been reflected by the reflecting mask 421 to the reduction projection optical system 424 continuously for 24 hours in the soft X-ray reduction projection exposure system of FIG. 5. In FIG. 7, points shown with xe2x96xa1 indicate actually measured values of the thickness of the carbon film and a solid line is a virtual line obtained on the basis of these actually measured values. Also, points shown with xcex94 indicate actually measured values of the transmittance loss and a broken line is a virtual line obtained on the basis of these actually measured values. As is obvious from FIG. 7, in the case where the partial pressure of the ozone gas is lower than 1.33xc3x9710xe2x88x924 Pa, the thickness of the carbon film deposited on the surface of the reduction projection optical system 424 through the irradiation with the soft X-ray beam is abruptly increased. In contrast, in the case where the partial pressure of the ozone gas exceeds 1.33xc3x9710xe2x88x924 Pa, the carbon film is minimally deposited. On the surface of the reduction projection optical system 424, an oxidation/decomposition reaction caused by ozone molecules always occurs. Accordingly, the thickness of the carbon film deposited on the surface of the reduction projection optical system 424 largely depends upon the partial pressure of the ozone gas. In the case where the partial pressure of the ozone gas is 1.33xc3x9710xe2x88x924 Pa or more, a degree of decomposing carbon is higher than a degree of carbon depositing on the surface of the reduction projection optical system 424, so that the thickness of the carbon film cannot increase. Although the thickness of the carbon film is reduced as the partial pressure of the ozone gas is higher, when the partial pressure of the ozone gas is too high, a light absorbing function of the ozone molecules becomes too large to neglect, and hence, the transmittance loss occurs. When the transmittance loss per meter of the soft X-ray beam exceeds 1%, the proportion of the soft X-ray beam generated by the discharge type X-ray source 411 to reach the surface of the wafer 431 is disadvantageously lowered. As is understood from FIG. 7, in the case where the partial pressure of the ozone gas is 4.00xc3x9710xe2x88x922 Pa, the transmittance loss is 1%. Therefore, it seems that the upper limit of the partial pressure of the ozone gas is 4.00xc3x9710xe2x88x922 Pa. For this reason, the partial pressure of the ozone gas is controlled to be 1.33xc3x9710xe2x88x924 Pa through 4.00xc3x9710xe2x88x922 Pa in Embodiment 4. Therefore, the carbon film is minimally deposited on the reflecting face of the reduction projection optical system 424 without increasing the transmittance loss of the soft X-ray beam. Accordingly, the optical characteristic can be prevented from degrading due to the contamination of the reflecting face of the reduction projection optical system 424 with an organic substance. In this embodiment, the relationship between the partial pressure of the ozone gas within the region where the reduction projection optical system 424 is disposed in the second chamber 420 and the thickness of the carbon film deposited on the surface of the reduction projection optical system 424 is described. This relationship also holds between the partial pressure of the ozone gas within a region where the illumination optical system 412 is disposed in the first chamber 410 and the thickness of a carbon film deposited on the surface of the illumination optical system 412 and between the partial pressure of the ozone gas within the region where the reflecting mask 421 is disposed in the second chamber 420 and the thickness of a carbon film deposited on the surface of the reflecting mask 421. Specifically, when the partial pressure of the ozone gas within the region where the illumination optical system 412 is disposed in the first chamber 410 and the partial pressure of the ozone gas within the region where the reflecting mask 421 is disposed in the second chamber 420 are respectively controlled to be 1.33xc3x9710xe2x88x924 Pa through 4.00xc3x9710xe2x88x922 Pa, the carbon films can be substantially prevented from depositing on the reflecting faces of the illumination optical system 412 and the reflecting mask 421 without increasing the transmittance loss of the soft X-ray beam. Although the partial pressure of the ozone gas is controlled to be 1.33xc3x9710xe2x88x924 Pa through 4.00xc3x9710xe2x88x922 Pa in all of the inside region of the first chamber 410 and the regions where the reflecting mask 421 and the reduction projection optical system 424 are disposed in the second chamber 420 in Embodiment 4, the partial pressure of the ozone gas may be controlled to be 1.33xc3x9710xe2x88x924 Pa through 4.00xc3x9710xe2x88x922 Pa in at least one of these regions. As a characteristic of Embodiment 4, the total pressure within the first chamber 410 and the total pressures within the regions where the reflecting mask 421 and the reduction projection optical system 424 are disposed in the second chamber 420 are controlled to be 4.00xc3x9710xe2x88x922 Pa or less. Thus, any gas other than the ozone gas, such as a gas of an inorganic substance like a metal element, can be suppressed. Therefore, the optical characteristics can be prevented from degrading due to the contamination, with an inorganic substance, of the reflecting faces of the illumination optical system 412, the reflecting mask 421 and the reduction projection optical system 424. In Embodiment 4, the partial pressure of the ozone gas within the first chamber 410 is controlled by the first diffusion pump 413, and the partial pressures of the ozone gas within the regions where the reflecting mask 421 and the reduction projection optical system 424 are disposed in the second chamber 420 are controlled by the second diffusion pump 425. These diffusion pumps may be provided to the respective chambers or the respective regions, or a common diffusion pump may be provided to a plurality of chambers or a plurality of regions. In particular, the partial pressure of the ozone gas within the region of the reflecting mask 421 is disposed in the second chamber 420 and the partial pressure of the ozone gas within the region where the reduction projection optical system 424 is disposed in the second chamber 420 are both controlled by the second diffusion pump 425 in Embodiment 4. However, the partial pressures of the oxygen gas within the region where the reflecting mask 421 is disposed and within the region where the reduction projection optical system 424 is disposed are preferably individually controlled by different diffusion pumps. Although the discharge type X-ray source 411 is used as the soft X-ray source in Embodiment 4, another soft X-ray source such as a laser induced plasma X-ray source may be used instead. Also, although the reflecting mirrors are used as the illumination optical system 412 and the reduction projection optical system 424 in Embodiment 4, another means may be employed instead. Furthermore, although the diffusion pumps are used as pressure reducing means in Embodiment 4, another vacuum pumping device such as a turbo pump or an ion pump may be used instead. Embodiment 5 In Embodiment 5, a pattern formation method performed by using the soft X-ray reduction projection exposure system according to any of Embodiments 1 through 4 will be described. Although the soft X-ray reduction projection exposure system of Embodiment 1 is used in the following description, it goes without saying that the pattern formation method can be similarly performed by using the soft X-ray reduction projection exposure system of Embodiment 2, 3 or 4. First, after forming a multi-layer film composed of a molybdenum film and a silicon film on a glass substrate, an absorbing film of chromium is selectively formed on a portion of the multi-layer film where a desired pattern is to be formed. Thus, the reflecting mask 121 is prepared and then held on the mask stage 122. Next, a resist film with a thickness of 200 nm is formed by applying a resist material photosensitive with a soft X-ray beam onto the wafer 131 by spin coating, and the resist film is cured through annealing at a temperature of 110xc2x0 C. for 60 seconds. Thereafter, the wafer 131 having the resist film is held on the wafer stage 132. Then, a soft X-ray beam is generated by the radiating X-ray source 111 within the first chamber 110 in which the partial pressure of a carbon compound gas is controlled to be 1.33xc3x9710xe2x88x928 Pa or less. The generated soft X-ray beam are transmitted by the illumination optical system 112 to the second chamber 120 in which the partial pressure of the carbon compound gas is controlled to be 1.33xc3x9710xe2x88x928 Pa or less. In the second chamber 120, the soft X-ray beam having been transmitted from the first chamber 110 is introduced to the reflecting mask 121 by the reflecting optical system 123, and the soft X-ray beam having been reflected by the reflecting mask 121 is transmitted by the reduction projection optical system 124 to the third chamber 130 in which the partial pressure of the carbon compound gas is controlled to be 1.33xc3x9710xe2x88x928 Pa or less, so that the resist film formed on the wafer 131 disposed in the third chamber 130 can be irradiated with the soft X-ray beam for pattern exposure. Thereafter, the wafer 131 is taken out from the soft X-ray reduction projection exposure system, and the resist film is subjected to post-bake at a temperature of 110xc2x0 C. for 60 seconds. Then, the resist film is developed with an alkaline developer, resulting in obtaining a resist pattern. In this manner, a resist pattern with a pattern width of 50 nm free from pattern distortion derived from aberration can be precisely formed. Even after this pattern formation method is continuously performed for a half year or 1 year, no pattern distortion derived from aberration is caused and the resolution is not changed from that attained at the beginning. Accordingly, since the pattern exposure is performed with the inside pressures of the first chamber 110 and the second chamber 120 controlled to be 1.33xc3x9710xe2x88x928 Pa or less in the pattern formation method of Embodiment 5, the illumination optical system 112, the reflecting mask 121 and the reduction projection optical system 124 are not contaminated with an organic substance. As a result, a precise resist pattern with no distortion derived from aberration can be stably formed for a long period of time. Although the glass substrate, the multi-layer film composed of a molybdenum film and a silicon film and the chromium film are used as the materials for the reflecting mask 121, the structure of the reflecting mask is not limited to this. Also, the conditions for forming a resist film are not limited to those described above.
	summary
050013548	summary	BACKGROUND OF THE INVENTION At the present time, procedures in various medical areas such as radiology and orthopedics, require physicians to place their hands near or even in a fluoroscopic field. These procedures are relatively new and provide improved patient care, less traumatic operations and shortened recovery periods. Such procedures include percutaneous kidney stone removal or percutaneous biliary shunt placement, both of which have virtually replaced open surgical procedures and are performed by radiologists. In orthopedics, blind nailing of bones such as the femur are also much less invasive and are increasingly replacing open surgical procedures. In all of these procedures, however, the physician is required to work for prolonged periods of time under fluoroscopic guidance with the physician's hands near or in the fluoroscopic field. An active physician performing these type of procedures will rapidly accumulate high doses of ionizing radiation to the hands, often well above the limits allowed. The International Council for Radiation Protection currently recommends a dose limit of not more than 50 rems/year to the hands. During certain non-vascular procedures, exposure to the hands has been shown to be about 33 times more than the dose received at the physician's face. Thus, there is a present need for providing a surgical glove having good tactile sense in order to permit the physician to perform the procedures involved and that the glove provide fully effective hand protection to the physician against exposure to X-rays. That is, the glove must have an X-ray absorbing composition uniformly distributed throughout the glove and be free of pin holes Presently, there are available lead oxide loaded gloves produced from polyurethane or polyvinylchloride. Unfortunately, these gloves lack the tactile sense needed for surgical use since they are essentially non-stretchable due to the nature of polyurethane or other material compared with natural latex rubber, and this is exacerbated by the heavy loading with lead oxide. Examples of such gloves are disclosed in U.S. Pat. Nos. 3,025,403; 3,045,121; 3,185,751 and 3,883,749. None of these patents discloses a means for preparing gloves from natural latex which are free of pinholes. In addition, the prior art gloves are relatively ineffectual for attenuating X-rays. For example, in the ranges normally used, e.g., 80-100 kilovolts, these gloves absorb only about 20% of the incident radiation. These gloves also share the common drawback in that their production is dangerous since lead oxide is toxic. Accordingly, it would be highly desirable to provide a glove having a tactile sense which permits their use in surgery as well as a glove capable of attenuating substantially greater amounts of incident radiation as compared to presently available gloves It would be highly desirable to provide an X-ray absorbing glove from natural latex since gloves produced from latex are highly stretchable. However, there is no presently available process for producing high particle density loaded latex gloves which avoids latex coagulation and/or air entrapment while maintaining a homogeneous latex mix. Thus, there is no presently available method for making thin loaded latex gloves free of pin holes. SUMMARY OF THE INVENTION The present invention provides surgical gloves containing particles having a high specific gravity at a concentration which permits the gloves to absorb a high percentage of incident radiation. The particles can comprise a metal or a metal compound which has a specific gravity of at least 11. The metal or metal compound particles are admixed with a natural rubber latex composition under conditions which avoid latex coagulation and which avoid cavitation or any entrainment of air at all at the exposed top surface of the mixture in a container until a substantially uniform mixture of the resulting composition is produced. The glove is produced from the latex metal or metal compound mixture by dipping a glove form into the mixture and allowing the mixture to dry on the form. The glove can be formed from one or a multiplicity of layers which multiplicity of layers can include a natural rubber latex layer free of filler particles.
	description	A reactor-internal equipment handling apparatus according to an embodiment of the present invention will be explained in detail with reference to the accompanying drawings hereinafter. FIG. 1 is a longitudinal sectional view showing a reactor-internal equipment handling apparatus 18 according to an embodiment of the present invention. This reactor-internal equipment handling apparatus 18 can simultaneously load/unload all of the CR 7, the FS 8, and the CRGT 6 into/from a reactor by a remote manipulation. The reactor-internal equipment handling apparatus 18 has a main body frame 26. FIG. 1 shows a state where the main body frame 26 is properly positioned at a predetermined position in a reactor pressure vessel 1 (see FIG. 2). FIG. 2 is a longitudinal sectional view showing a state where the reactor-internal equipment handling apparatus 18 is installed inside the reactor pressure vessel 1 by a refueling machine 14. In this case, when the reactor-internal equipment handling apparatus 18 is to be utilized, four fuel assemblies 10 (see FIG. 10) which are located in an objective working area are pulled out from the core previously by the refueling machine 14. A top end of the main body frame 26 is fitted to a bottom end of a hoist rope 25 of the refueling machine 14 shown in FIG. 2 such that the main body frame 26 can be lifted up and down by there fueling machine 14. A guide member 27 is fitted to the hoist rope 25 to be faced to a top surface of the main body frame 26. To the main body frame 26 are fitted a control rod grapple (referred to as xe2x80x9cCR grapplexe2x80x9d hereinafter) 16 acting as a control rod holding means which releasably holds the CR 7 being placed inside the reactor pressure vessel 1, and fuel support/control rod guide tube grapples (referred to as xe2x80x9cFS/CRGT grapplesxe2x80x9d hereinafter) 17 each acting as a fuel support/control rod guide tube holding means which releasably holds both the FS 8, which supports the bottom end of the fuel assembly 10, and the CRGT 6, on which the FS 8 is positioned at top end. FIG. 3 is a longitudinal sectional view showing the main body frame 26 and the FS/CRGT grapples 17 in an enlarged manner. A pair of FS/CRGT grapples 17 are fitted to the main body frame 26. Each of the FS/CRGT grapples 17 comprises an orifice engaging hook (orifice engaging member) 28 which can engage edge portions of both the FS orifices 33 shown in FIG. 12 and the CRGT orifices 32 shown in FIG. 11, an orifice engaging hook linking mechanism (orifice engaging member linking mechanism) 30 which is employed to operate the orifice engaging hook 28, and an orifice engaging hook driving cylinder (orifice engaging member driving means) 19 which is employed to drive the orifice engaging hook linking mechanism 30. Preferably, each of the orifice engaging hook driving cylinders 19 is composed of an air cylinder. Also, a clamping state detecting mechanism (holding state detecting mechanism) 20, which detects holding states of the FS/CRGT grapples 17 about the FS 8 and the CRGT 6, is provided to each of the FS/CRGT grapples 17. This clamping state detecting mechanism 20 is placed on a top portion of the orifice engaging hook driving cylinder 19. The clamping state detecting mechanism 20 has limit switches (detection switches) 44 whose on/off state can be switched depending on a change in clamping states. More particularly, the clamping state detecting mechanism 20 has a limit switch 44 which is directly on/off-operated by an output axis of the orifice engaging hook driving cylinder 19 when the output axis is moved back and forth simultaneously with a motion of the orifice engaging hook 28, and a limit switch 44 which is on/off-operated by transmitting a back-and-forth motion of an output axis via a lever mechanism 62. Further, each of the FS/CRGT grapples 17 has a guide portion 29. This guide portion 29 has a function of seating the main body frame 26 on a predetermined position without fail by guiding an inside of the fuel assembly sustaining hole 31 (see FIG. 12) of the FS 8. Both an FS stepped portion 34a which comes into contact with an edge portion of the FS orifice 33 (see FIG. 12) of the FS 8, and a CRGT stepped portion 34b which comes into contact with an edge portion of the CRGT orifice 32 (see FIG. 11) of the CRGT 6 are formed on the orifice engaging hook 28. With the use of the FS stepped portions 34a and the CRGT stepped portions 34b, both the FS 8 and the CRGT 6 can be handled simultaneously. The orifice engaging hook linking mechanism 30 is so constructed that an opening/closing motion of the orifice engaging hook 28 can be disabled in the situation that the FS stepped portion 34a and the CRGT stepped portion 34b are brought into contact with the edge portions of the FS orifice 33 and the CRGT orifice 32 respectively. In more detail, when the orifice engaging hook linking mechanism 30 is shifted from its clamping state (holding state) to its releasing state (non-holding state), the orifice engaging hook 28 once protrudes outwardly from the orifices 32, 33 and then withdraws inwardly. Thus, in the situation that both the FS 8 and the CRGT 6 are being hoisted or only the FS 8 is being hoisted, a mechanical lock can be made by its own weight of the hoisted substance and the FS stepped portions 34a and the CRGT stepped portions 34b of the orifice engaging hooks 28. Therefore, even when either an actuating pressure of the orifice engaging hook driving cylinder 19 is lost at the worst or the operator operates it erroneously, the hoisted substance is never released or unengaged. As shown in FIG. 4 and FIG. 5, the reactor-internal equipment handling apparatus according to the present embodiment further comprises a stroke varying mechanism 35 which can change an operation stroke of the orifice engaging hook driving cylinder 19. The stroke varying mechanism 35 is composed of disk-like stoppers 36, stroke varying blocks 37, and an arm 38. The disk-like stoppers 36 are provided to output axes of two orifice engaging hook driving cylinders 19 of the FS/CRGT grapples 17 on the orifice engaging hook linking mechanism 30 side respectively. Each of the stroke varying blocks 37 is rotatably mounted between the disk-like stopper 36 and the orifice engaging hook driving cylinder 19 by pins 39 being provided to a cylinder case. The arm 38 can couple the stroke varying blocks 37 with each other. When the arm 38 is moved vertically, both stroke varying blocks 37 are put in and out simultaneously. A swing amount (amount of motion) of the orifice engaging hook 28 can be adjusted by changing an operating stroke of the orifice engaging hook driving cylinder 19 by the stroke varying mechanism 35. Therefore, the orifice engaging hook 28 can be set not to be engaged by the edge portion of the orifice 32 of the CRGT 6. As a result, the FS/CRGT grapple 17 cannot clamp the CRGT 6, but it can clamp only the FS 8. For example, in the case that a load applied in pulling out the CRGT 6 exceeds a limit load of the hoist of the refueling machine 14 because the core plate 3 and the CRGT 6 have stuck together, only the CR 7 and the FS 8 can be hoisted by operating the stroke varying mechanism 35 in order not to exceed the limit load of the refueling machine 14. As shown in FIG. 1, the CR grapple 16 is fitted to the main body frame 26 such that it can be slid by a predetermined width along the longitudinal direction of the CR 7. This predetermined sliding width is defined by the distance between the flange portion 70a of the movable member 70 and the inner upper surface 71a of the cap member 71. The movable member 70 is fixed to both the hoist rope 25 and the CR grapple 16. On the other hand, the cap member is fixed to the main body frame 26. In contrast, the FS/CRGT grapple 17 is secured to the main body frame 26. For this reason, the CR grapple 16 and the FS/CRGT grapple 17 can be relatively displaced mutually along the longitudinal direction of the CR 7. Therefore, when the CR 7, the FS 8, and the CRGT 6 are to be hoisted, first the CR 7 is hoisted slightly and then the FS 8 and the CRGT 6 are hoisted. In this manner, a time difference can be introduced into an application of the load, which is equivalent to the head pressure caused by the air contained in the CRD housing 4, by hoisting the CR 7 previously, and as a result the simultaneously applied load can be reduced. Accordingly, the hoisting load to unload outside the reactor can be shared much more by the head pressure to lift the CRGT 6. As shown in FIG. 6, the CR grapple 16 has a hook (handle engaging member) 41 acting as an L-shaped swingable hooking member which can engage a hoisting handle 7a (see FIG. 1) secured to the top end of the CR 7. This hook 41 has a gaff 43. The hook 41 is connected to a hook driving cylinder (handle engaging member driving means) 40 via a linking mechanism 42, and is operated by the hook driving cylinder 40 to be swung. Then, in the situation that the CR 7 is hoisted via the hook 41, a mechanical lock using its own weight of the CR 7 can be made by the linking mechanism 42 and the gaff 43 of the hook 41. Such mechanical lock can act to hold a engaged state of the hoisting handle 7a by the hook 41. In addition, as shown in FIG. 6, a clamping state detecting mechanism (holding state detecting mechanism) 60 which detects a clamping state (holding state) of the CR 7 is provided to the CR grapple 16. This clamping state detecting mechanism 60 has a limit switch (detection switch) 61 whose on/off state can be changed depending upon a swing motion of the hook 41. More specifically, an on/off switching operation of the limit switch 61 is performed by a base end portion of the hook 41. Also, as shown in FIG. 6, the reactor-internal equipment handling apparatus 18 according to the present embodiment is connected to an external power supply 55 which is arranged apart from the reactor-internal equipment handling apparatus 18. A power supply for clamping state confirming indicator lamps (holding state confirming indicator lamps) 45 and a seating state confirming indicator lamp (positioning state confirming indicator lamp) 50, which are shown in FIG. 7, can be supplied from this external power supply 55. In place of the external power supply 55, a built-in battery (not shown) can be incorporated into the reactor-internal equipment handling apparatus 18. In this case, exchange of the battery must be performed by pulling up the main body frame 26 every run-down of the battery. If the work of unloading all the CRs 7, the FSs 8, and the CRGTs 6 must be carried out in the preventive maintenance work, etc., it is preferable to supply the power supply from the external power supply 55, which is placed on the refueling machine 14, etc., since an employment term of the reactor-internal equipment handling apparatus 18 is prolonged over a long term. As shown in FIG. 7, a plurality of clamping state confirming indicator lamps 45 whose lighting state can be changed depending upon a change in clamping states (holding states) are provided to both the clamping state detecting mechanism 20 (see FIG. 3) for the FS 8 and the CRGT 6 and the clamping state detecting mechanism 60 (see FIG. 6) for the CR 7. These clamping state confirming indicator lamps 45 are attached to the top surface of the main body frame 26. The clamping state confirming indicator lamps 45 can switch their lighting states according to on/off states of the limit switches 44, 61 (FIG. 3 and FIG. 6). In more detail, at least three clamping state confirming indicator lamps 45 are provided. A lighting state of a first clamping state confirming indicator lamp 45 can be switched by the limit switch 61 (see FIG. 6) which is switched depending upon a change in the clamping state of the CR 7. Also, a lighting state of a second clamping state confirming indicator lamp 45 can be switched by the limit switch 44 (see FIG. 3). When the orifice engaging hooks 28 shown in FIG. 3 is located at the position to clamp both the FS 8 and the CRGT 6, such limit switch 44 can be switched depending upon the change in the clamping state of the FS 8 and the CRGT 6 via the lever mechanism 62. In addition, a lighting state of a third clamping state confirming indicator lamp 45 can be switched by the limit switch 44. When the orifice engaging hooks 28 is located at the position to clamp only the FS 8, such limit switch 44 can be switched by the output axis of the orifice engaging hook driving cylinder 19 depending upon the change in the clamping state of the FS 8. In this way, by checking the lighting state of plural clamping state confirming indicator lamps 45 with the naked eye, the operator can know whether or not the reactor-internal equipment handling apparatus 18 has already clamped the CR 7, the FS 8, and/or the CRGT 6. As shown in FIG. 8 and FIG. 9, the reactor-internal equipment handling apparatus 18 is equipped with a first positioning state detecting mechanism 63 (FIG. 8) and a second positioning state detecting mechanism 64 (FIG. 9), which detect a positioning state of the main body frame 26 in the reactor pressure vessel respectively. As shown in FIG. 8, the first positioning state detecting mechanism 63 consists of a seating state detecting mechanism 21, a motion limiting mechanism 23, and a motion limiting mechanism locking device 24. While, as shown in FIG. 9, the second positioning state detecting mechanism 64 consists of the seating state detecting mechanism 21, and the motion limiting mechanism 23. The seating state detecting mechanism 21 is composed of seating detecting pins 46, 47, a cam mechanism 48, and a limit switch 49. The seating detecting pin 46 of the first positioning state detecting mechanism 63 is employed to detect the positioning pin 11 (see FIG. 1). The seating detecting pin 47 of the second positioning state detecting mechanism 64 is employed to detect the top surface of the FS 8. Two seating detecting pins 46, 47 are projected from the bottom surface of the main body frame 26. Thus, when the main body frame 26 is positioned or seated, the seating detecting pins 46, 47 are pushed upwardly by the top surfaces of the positioning pin 11 and the FS 8 on the core plate 3 (see FIG. 1). Such motions of the seating detecting pins 46, 47 are transmitted respectively via the cam mechanisms 48 to the limit switches 49, whereby the limit switches 49 can be operated. Then, when both limit switches 49 provided to the first positioning state detecting mechanism 63 and the second positioning state detecting mechanism 64 are operated, the seating state confirming indicator lamp 50 (see FIG. 7) provided on the top surface of the main body frame 26 can be illuminated. In this fashion, based on the illuminated state of the seating state confirming indicator lamp 50, the operator can visually check that the main body frame 26 has been seated on the position to properly clamp and unclamp the CR 7, the ES 8, and the CRGT 6. The motion limiting mechanism 23 has a cam 51 which is operated simultaneously with motions of the seating detecting pins 46, 47, and two valve switches 52, 52 whose on/off is switched by a vertical motion of the cam 51. Then, during the hoisting operation of both the FS 8 and the CRGT 6 or only the FS 8, motions of the orifice engaging hook driving cylinders 19 are limited by an air circuit (not shown) which is connected to the valve switches 52, 52. In other words, except the case where the main body frame 26 is seated on the proper position or where no load is applied the FS/CRGT grapple 17, i.e., the FS/CRGT grapple 17 is holding nothing, an actuating fluid is not supplied to the orifice engaging hook driving cylinders 19 so as to disable the orifice engaging hooks 28 of the FS/CRGT clamping mechanism 17. Accordingly, even when either an operating pressure of the orifice engaging hook driving cylinders 19 is lost or the operator performs the wrong operation in the course of the hoisting operation of both the FS 8 and the CRGT 6 or only the FS 8, the reactor-internal equipment handling apparatus 18 never releases the FS 8 and the CRGT 6. As shown in FIG. 8, the motion limiting mechanism locking device 24 is composed of a ball lock pin 54 and a stepped hole (not shown). The ball lock pin 54 is fitted to the upper portion in the main body frame 26. The stepped hole is formed over the motion limiting mechanism cam 51. The motion limiting mechanism locking device 24 is employed to use the Pin-FS 22 (see FIG. 13). More specifically, the Pin-FSs 22 which support a mimic fuel assembly (not shown) are provided to the peripheral portions of the core of the BWR. When the Pin-FSs 22 are employed, the main body frame 26 is rotated by 90 degree rightward or leftward rather than a normal orientation to avoid interference with the pin 53 (see FIG. 13), and then seated. The pin 53 is provided in the Pin-FS 22 to indicate the position of the mimic fuel assembly. In this case, detection of the seating state and restriction of the motion by virtue of the positioning pin 11 provided on the core plate 3 cannot be achieved. Therefore, the motion limiting mechanism 23 is operated and thus the motion of the orifice engaging hook 28 of the FS/CRGT grapple 17 is inhibited. As a result, the Pin-FS 22 and the CRGT 6 cannot be removed. For this reason, in such case, the ball lock pin 54 is inserted into a hole formed on the upper area of the cam 51 of the motion limiting mechanism 23 to lock the cam 51. At that time, since the seating detecting pin 46 of the first positioning state detecting mechanism 63 has already been in a seated condition, the seating on the normal position can be detected only by detecting the top surface of the Pin-FS 22 by the seating detecting pin 47 of the second positioning state detecting mechanism 64. As a result, clamping/unclamping of the Pin-FS 22 and the CRGT 6 can be achieved. In case the Pin-FS 22 is handled as described above, a function of the motion limiting mechanism 23 is lost. In this case, since the mechanical lock which has already been mentioned can be operated, the hoisted substance is in no way released even if, for example, the operating pressure of the orifice engaging hook driving cylinder 19 is lost or the operator performs the wrong operation. As described above, according to the reactor-internal equipment handling apparatus 18 of the embodiment of the present invention, in the event that the removing operation or the installing operation of the CR 7, the FS 8, and the CRGT 6 must be performed in the periodical inspection or the preventive maintenance work, all of the CR 7, the FS 8, and the CRGT 6 can be loaded/unloaded into/from the reactor simultaneously by the CR grapple 16 and the FS/CRGT grapple 17. Therefore, the number of steps can be reduced to half based on a simple calculation rather than the case where the CR 7 and the FS 8 are handled separately from the CRGT 6 in the related art, so that a term of work can be shortened considerably. In the reactor-internal equipment handling apparatus 18 according to the embodiment of the present invention, an operability can be assured to the same extent as the CR and FS grapple in the related art or more. In addition, if the CR and ES grapple and the CRGT grapple in the related art are employed, the CR 7, the ES 8 and the CRGT 6 must be stored separately based on the installing order in the reactor. Therefore, the wide storage space is needed as the fuel pool serving as the storage area. On the contrary, according to the reactor-internal equipment handling apparatus 18 of the embodiment of the present invention, since the CR 7, the ES 8, and the CRGT can be handled together, they can be stored collectively. Therefore, based on a simple calculation, the storage space can be reduced to half of the storage space needed in the related art. As described above, according to the reactor-internal equipment handling apparatus and method of the present invention, since all of the control rod, the fuel support, and the control rod guide tube can be loaded/unloaded into/from the reactor simultaneously, both reduction in the term of work and reduction in their storage spaces can be achieved.
	description	The present invention provides an ion beam generator capable of high average power and repetitive operation over an extended number of operating cycles for thermally treating large surface areas of materials at commercially attractive costs. In particular, the ion beam generator of the present invention can produce high average power (1-1000 kW) pulsed ion beams at 0.25-2.5 MeV energies and pulse durations or lengths of 30-500 nanoseconds (ns). The ion beam generator can directly deposit energy in the top 1-50 micrometers (xcexcm) of the surface of any material. The depth of thermal treatment can be controlled by varying the ion energy and species as well as the pulse length. Irradiating a material with ion beams in accordance with the present invention is a thermal process that does not significantly change the atomic composition of the material. Instead it thermally heats the near surface using typically 3xc3x971013 ions/cm2 per pulse, only approximately 10xe2x88x925-10xe2x88x923 atomic percent of the sample density. Deposition of ion beam energy in a thin near surface layer allows melting of the layer with relatively small energies (typically 1-10 J/cm2) and allows rapid cooling of the melted layer by thermal diffusion into the underlying substrate as depicted in FIG. 1. The relatively small energy densities needed for treatment together with the high instantaneous powers available using the present invention allows large surfaces areas (up to 1000 cm2) to be treated with a single ion beam pulse greatly reducing or eliminating edge effects at the transition between treated and untreated areas. The relatively short ion beam pulse lengths, preferably xe2x89xa6200 ns, developed by the ion beam generator limit the depth of thermal diffusion, thus allowing the treated/melted region to be localized to a selected depth. Typical cooling rates of the present invention (108-1010 K/sec) are sufficient to cause amorphous layer formation in some materials, fine grain structures in some materials, the production of non-equilibrium microstructures (nano-crystalline and metastable phases), and the formation of new alloys by rapid quenching and/or liquid phase mixing of layers of different materials. Such rapid thermal quenching ( greater than 108 K/sec) can significantly improve corrosion, wear and hardness properties of the treated near surface layer. Other applications of the present invention can include etching and cross linking of polymers, surface glazing and sealing of ceramic surfaces for reduced porosity, and a cost-effective, solvent-free process for surface deburring, polishing, cleaning, and oxide layer removal. The unique energy deposition as a function of depth, as depicted in FIG. 2, also allows ion beams produced according to the present invention to be used for bonding films to substrates and mixing interface materials by liquid phase mixing. Surface modification by rapid, localized melting, with or without vaporization of the surface can be used for annealing of surfaces and modifying surface microstructure, including metals, semiconductors, and polymers. The microstructures produced will be determined by the choice of material, the beam intensity, energy, composition, and the number of pulses used to treat an area. The ion beam generator can also produce pulsed ion beams with intensities sufficient to vaporize surface layers. Applications of this capability include the production of high energy vapor for depositing films. The ion beam generator can also be used for ion implantation at relatively high voltages (0.25-2.5 MV), with greatly reduced cost, greater depths, and larger treatment area due to the higher energy per pulse, higher voltage, and low cost, repetitive capability of this new technology. At higher intensities, the ion beam generator can also be used to produce surface ablation that creates shock waves that propagate into materials, causing dislocations and increasing the hardness of the material. The ion beam generator of the present invention is composed of two major components: a high energy, pulsed power system and an ion beam source both capable of high repetition rates and both having extended operating lives. The first of these components is a compact, electrically efficient, repetitively pulsed, magnetically switched, pulsed power system capable of 109 pulse operating cycles as generally described by H. C. Harjes, et al, Pro 8th IEEE Int. Pulsed Power Conference (1991) all of which is incorporated by reference herein. The power system can operate continuously at a pulse repetition rate of 120 Hz delivering up to 2.5 kJ of energy per pulse in 60 ns pulses. The power system can deliver pulsed power signals of 30-500 ns duration with ion beam energies of 0.25-2.5 MeV. The power system can operate at 50% electrical efficiency from the wall plug to energy delivered to a matched load. The power system uses low pulse compression stages incorporating, for example, low loss magnetic material and solid state components, to convert AC power to short, high voltage pulses. High electrical efficiency is important in reducing the cooling requirements and the capital and operating costs of the power system. The ability to produce voltages from 250 kV to several MV by stacking voltage using inductive adders incorporating low loss magnetic material is also an important feature when high voltages are needed although it is also possible to use a single stage, eliminating the need for the adder. The power system can operate at relatively low impedances ( less than less than 100 xcexa9) which also sets it apart from many other repetitive, power supply technologies such as transformer-based systems. This feature is necessary to allow high treatment rates and the treatment of large areas (up to 1000 cm2) with a single pulse so as to reduce edge effects occurring at the transition between treated and untreated areas. A block diagram of the power system P is shown in FIG. 3. From the prime power input, several stages of magnetic pulse compression and voltage addition are used to deliver a pulsed power signal of up to 2.5 MV, 60 ns FWHM, 2.9 kJ pulses at a rate of 120 Hz to an ion beam source. The power system P converts AC power from the local power grid into a form that can be used by the ion beam source. In one embodiment of the invention, the power system P comprises a motor driven 600 kW, 120 Hz alternator 10. In the unipolar mode it provides 210 A rms at a voltage of 3200 V rms with a power factor of 0.88 to a pulse compressor system 15. In an alternative embodiment, the alternator 10 can be replaced by a variable frequency modulator (not shown). Such modulator can deliver a 5 kJ, (1xe2x88x92cos xcfx89t) voltage pulse to the pulse compressor system 15 with a time to peak that permits the elimination of the first 2 or 3 stages in the pulse compressor system 15. The modulator can have a variable pulse repetition frequency (prf) and consequently will be a more flexible power source for the ion beam source (i.e. the input power can be adjusted by simply changing the prf). The pulse compression system 15 can provide unipolar, 250 kV, 15 ns rise, 60 ns FWHM, 4 kJ pulses, at a rate of 120 Hz, to a linear inductive voltage adder (LIVA) 20. In one embodiment, the pulse compression system 15 is a common magnetic pulse compressor composed of a plurality of magnetic switches (i.e. saturable reactors) the operation of which is well known to those skilled in the art. In order to satisfy the systems performance requirements, the pulse compression system 15 should have an efficiency 80% and be composed of high reliability components with very long lifetimes (xcx9c109-1010 pulses). Magnetic switches are preferably used in all of the pulse compression stages because they can handle very high peak powers (i.e. high voltages and currents), and because they are basically solid state devices with the potential to satisfy the lifetime requirement. The LIVA 20 is a liquid dielectric insulated voltage adder and can deliver nominal 2.5 MV, 2.9 kJ, pulses at a rate of 120 Hz to the ion beam source 25. The nominal output pulse of the LIVA 20 is trapezoidal with 15 ns rise and fall times and 60 ns FWHM. The second component of the present invention is an ion beam source or accelerator capable of operating repetitively and efficiently to transform the pulsed power signal from the power system efficiently into an ion beam as described generally by J. B. Greenly et al, xe2x80x9cPlasma Anode Ion Diode Research at Cornell: Repetitive Pulse and 0.1 TW Single Pulse Experimentsxe2x80x9d, Proceedings of 8th Intl. Conf. on High Power Particle Beams (1990) all of which is incorporated by reference herein. The ion beam source is capable of operating at repetitive pulse rates of 100 Hz continuously with long component lifetimes greater than 106. Preferably, the ion beam source comprises a magnetically-confined anode plasma (MAP) source, which draws ions from a plasma anode rather than a solid dielectric surface flashover anode used in present single pulse ion beam sources. The MAP source can provide pure beams of different ion species without employing components or techniques that have intrinsically short lifetimes. The ion beam generates is pulsed rather than continuous, allowing more compact, less expensive equipment to achieve high ion energies. Any ion can be used to deposit the energy. Protons have the greatest penetration depth of any ion and thus provide the greatest treatment depth while higher mass ions can deposit their energy at lesser depths. The MAP source M is shown in FIG. 4. In particular, FIG. 4 is a partially cross-sectional view of one symmetric side of the MAP source M. The MAP source M produces an annular ion beam K which can be brought to a broad focus symmetric about the axis Xxe2x80x94X shown. In the cathode electrode assembly 30 slow (1 xcexcs rise time) magnetic field coils 32 produce magnetic flux S which provides the magnetic insulation of the accelerating gap between the cathode 32 and the anode electrode assembly 34 and which is connected to the output of the pulsed power system (not shown). The anode electrodes 34 also act as magnetic flux shapers. The MAP M source operates in the following fashion: a fast gas valve 36 on the axis of anode assembly 35 produces a rapid (200 xcexcs) gas puff which is delivered through a supersonic nozzle 38 to produce a highly localized volume of gas directly in front of the surface of a fast driving coil 40 located in an insulating structure 42. After pre ionizing the gas with a 1 xcexcs induced electric field, the fast driving coil 40 is energized, inducing a loop voltage of 20 kV on the gas volume, driving a breakdown to full ionization, and moving the resulting plasma toward the flux shaping electrodes 34 in about 1.5 xcexcs, to form a thin magnetically-confined plasma layer. The pulsed power signal from the power system is then applied to the anode assembly 35, accelerating ions from the plasma to form the ion beam K. The slow (S) and fast (F) magnetic flux surfaces, at time of ion beam extraction, are also shown. The plasma can be formed using a variety of gases. This ion source system can use any gas (including hydrogen, nitrogen and argon) or high vapor pressure liquid or metal to produce a pure source of ions without consuming or damaging any component other than the gas supplied to the source. The ion beam K propagates 20-30 cm in vacuum (xcx9c10xe2x88x923) to a broad focal area (up to 1000 cm2) at the target plane where material samples are placed for treatment and can thermally alter areas from 50 cm2 to 1000 cm2. Operation of the MAP source depends in part on the ability to provide ions at current densities exceeding by factors of several those available using standard space charge limited flow defined by the geometric gap and voltage. This is done by forming a virtual cathode consisting of electrons emitted from the cathode near the ion source on the anode. The present invention provides a system and a process for generating high voltage ion beams repetitively over an extended number of operating cycles in a manner that satisfies several criteria and constraints imposed by the use of this technology for the efficient treatment of surfaces in commercial applications. In particular, the present invention comprises a pulsed power system and an ion beam source that are designed and combined in such a way as to satisfy the following criteria. The operation of high current, pulsed ion beam source depends in part on the ability to provide ions at current densities exceeding by factors of several those available using standard space charge limited flow defined by the geometric gap and voltage. This is done by forming a virtual cathode consisting of electrons emitted from the cathode near the ion source on the anode. A carefully optimized magnetic field topology is required to confine these electrons to form a sheath extending from the cathode to the anode in the ion emitting region of the anode. In the simplest model these electrons fill the gap in a few nanoseconds, providing a space charge profile in the gap that allows ion current densities to be much larger than those available without the virtual cathode formed by the electron sheath filling the anode-cathode gap. Experiments show that the delay in obtaining enhanced ion current density is 10-20 ns. This places a lower limit on the pulse length required for reasonable efficiency of the system. Based on this information the pulse length above half voltage points should be at least 30 ns. The voltage, current and impedance of the pulsed power system required for effective, repetitive ion beam surface treatment is constrained by several factors. These include the need to generate large enough voltages to obtain the needed depth of penetration of ions into materials. This voltage is at least 250 kV for multi-micron penetration depth and extends to several MV. A typical operating voltage is 0.75 MV. The energy deposition depth of protons in steel is approximately 5 microns at this voltage. The current density needed to provide 5 J/cm2, a typical level needed to melt surfaces to this depth, is approximately 100 A/cm2 for a 60 ns pulse. In order to treat a 100 cm2 area (large enough to minimize edge effects) a total ion current of 10,000 Amperes is needed. A 67% ion efficiency gives total current and impedance requirements of 15,000 Amperes and 50 xcexa9 respectively. In generally these considerations lead to an ion source impedance significantly less than 100 xcexa9, at voltages of at least 250 kV. The ion beam source must satisfy several criteria that are unique to its combination in repetitively pulsed, long lifetime, high voltage systems for pulsed ion beam treatment of surfaces. The ion beam source must produce only low levels of gas during operation to prevent vacuum degradation beyond the 1xc3x9710xe2x88x923 Torr level between pulses. The ion beam generated must be extractable through any insulating magnetic field with little or no rotation to allow propagation of the ion beam in a field-free region and focusing of the beam at intensities up to several tens of Joules/square centimeter on a material surface at least 20 centimeters away from the ion beam source. Many ion beam systems, including previous Magnetically confined Anode Plasma systems were fundamentally incapable of this because of incompatible magnetic field configurations in the diode (ion beam generating section). In particular, previous MAP sources produced an ion beam which rotated. Such rotation can result in the rapid dispersal of the ion beam as it propagates in free space. In the present invention it is necessary for the ion beam to travel up to 20-30 cm to a material surface and as such rotation of the ion beam is unacceptable. Applications of present invention include production of low cost materials with treated surfaces for handling corrosive environments and treatment of large area metal sheets and critical components used in manufacturing and other areas where hardness, toughness and corrosion or wear resistance are important. It can also be used to produce smooth, crack-resistant ceramic surfaces by melting and re solidification using pulsed ion beams. By varying ion beam deposition levels and pulse durations it is also possible to use this technology for either surface cleaning or annealing. At high deposition levels (xcx9c30 J/cm2) it is possible to do shock hardening of materials. Polymer processing and ion implantation can be done at lower deposition levels. Examples of thermal surface treatment using this process are shown in FIGS. 5, 6, 7a, 7b and 8a, 8b. Surface cleaning or preparation without the use of organic solvents, can be accomplished by choosing higher mass ions (e.g. nitrogen, carbon) to deposit all of the beam energy in an approximately 1 xcexcm thin region at the surface of the material, producing high surface temperatures which vaporize this contaminant-containing layer without significantly disturbing the underlying material. In particular, FIG. 5 depicts a stainless steel 304 sample, which was initially coated with machining oil, producing a 100 nm thick coating of hydrocarbons, after treatment with 1-2 Joules per square centimeter of 0.75 MeV carbon ions the hydrocarbon contamination layer was removed leaving only normal atmospheric contamination FIG. 6 is a cross sectional view of an 0-1 tool steel sample showing the effects of rapid surface melting and cooling after exposure by a 60 ns, 10 J/cm2, 1 MeV mixed proton and carbon beam which resulted in increased hardness as well as a mixture of amorphous and retained metastable austenite with dissolved carbide precipitates that are retained in solution. FIGS. 7a depicts a sample of Ti-6A1-4V alloy after treatment with a 300 keV, 2-4 J/cm2 beam of mixed protons and carbon resulting in polishing of pre-existing machining marks with 10 xcexcm roughness to less than 1 xcexcm depicted in FIG. 7b. By inducing surface melting, porous dielectric films can be sealed or metal surfaces xe2x80x9cpolishedxe2x80x9d by material reflow. FIG. 8 depicts a sample of polished alumina which was subjected to a 60 ns, 10 J/cm2, 1 MeV beam of mixed protons and carbon resulting in surface melting and re-solidification. Unlike lasers which deposit energy to metals in the near surface (≈30 nm) only, an ion beam deposits its energy throughout its penetration depth extending to several microns. Additionally ion beams have the unique capability to deposit energy preferentially near the end of their penetration depth, in the interior of the sample, due to the inverse dependence of ion stopping power and energy. In addition to improving adhesion between the coating and substrate over large areas, the thermal fatigue resistance can also be enhanced. In a ceramic/metal system thermal-expansion-generated stresses at the boundary can be distributed over a relatively thick boundary layer following pulsed melting rather than being concentrated at the original atomically sharp interface. The flexibility of choosing the ion beam species and energy density provide substantial control over the degree of mixing, especially for dissimilar materials. Another related application of this technology is in the use of its rapid quenching and surface smoothing capability to produce thin layers of amorphous material with much higher quench rates and at much lower cost than is possible using existing splat-quenching techniques. The use of this technology to form very fine grain materials is also valuable in advanced battery applications. The capability of the present invention for producing high purity, high average power ion beams results in the potential for a new, low cost, compact surface treatment technology capable of high volume commercial applications and new treatment techniques not possible with existing systems. Having thus described the present invention with the aid of specific examples, those skilled in the art will appreciate that other similar combinations of the capabilities of this technology are also possible without departing from the scope of the claims attached herewith.
039740293	description	The reactor comprises a core 1 contained within a chamber 2 through which a gas such as helium is arranged to be circulated in normal use of the reactor, the gas passing around a closed circulatory system comprising a plurality of parallel loops 3, each incorporating a gas turbine, one of which is represented at 4, coupled to a main alternator 5. Each turbine unit 4 includes, in the usual way, a power turbine 6 coupled to the alternator 5, and a high pressure turbine 7 for driving low pressure and high pressure compressors 8 and 9, a precooler 10 and intercooler 11 being provided for cooling the gas passing to the low pressure and high pressure compressors respectively, and a recuperator 12 effecting an exchange of heat between the gas leaving the power turbine 6 and that fed from the high pressure compressor 9 to the reactor chamber 2. In accordance with the invention there is also provided an auxiliary circulatory system comprising a plurality of auxiliary cooling loops, one of which is represented at 13, connected to the chamber 2 in parallel with the main circulatory system. Each of these auxiliary cooling loops incorporates a motor driven circulator 15 producing a flow of gas around the loop, and a boiler 14 in which the hot gas leaving the reactor core 1 provides the heat for generating steam for an auxiliary steam-turbine 20 driving an auxiliary generator 16. The auxiliary boiler/steam turbine system is arranged to be run continuously and provides sufficient cooling to cover reactor start-up, shut-down and emergency conditions with the main gas turbine system inoperative. Power produced by the auxiliary generator 16 can be used to supply power for reactor services and ancillary equipment. The arrangement thus enables the reactor to be started-up separate from the gas turbine cycle start-up, and in addition long term heat removal, either after a reactor trip or under normal reactor shut-down is facilitated by shutting down the gas turbine units and removing heat from the core by the auxiliary circulatory system, as previously explained. The temperature of the gas fed to the auxiliary boilers 14 from the reactor core 1 is preferably reduced, by mixing the gas emerging from the core with gas at a lower temperature. For this purpose each cooling loop 13 may be associated with a by-pass pipe 17 which allows some of the cooled gas leaving the respective boiler 14 to by-pass the reactor core 1 as shown. However other means of cooling the gas between the reactor core and the boilers can alternatively be provided. Thus the gas leaving the reactor core may be mixed with cooler gas from another source, for example an appropriate part of the gas turbine system. Typically four gas turbine units and four auxiliary boilers will be associated with the reactor and these are conveniently located in chambers, commonly termed pods, in a concrete pressure vessel wall which surrounds and provides the chamber 2 for the reactor core 1. FIG. 2 shows one such arrangement, each gas turbine unit 4, including associated compressors, recuperator and the like, being housed in an adjacent pair of chambers or pods 21, 22 formed within and spaced around a pressure vessel wall 18 which surrounds the reactor core, each auxiliary boiler 14 being located in a pod 23 separating adjacent pairs of gas turbine pods as shown. In a modification, not illustrated, the reactor core is housed in a primary pressure vessel, surrounded by a secondary pressure vessel, the gas turbine units and auxiliary boilers being located in the space between the primary and secondary pressure vessels, for example in a manner similar to that described with reference to FIGS. 5 and 6 of co-pending U.S. patent application Ser. No. 222947. Although the invention is particularly applicable to nuclear reactors of the kind incorporating closed cycle gas turbines, it will be appreciated that it has application to reactors in which alternative forms of energy convertors are employed. Thus the invention may be used to advantage in gas cooled reactor systems in which the major portion of the reactor heat is used for process heating.
	summary
	abstract	There is provided an illumination system. the illumination system includes (a) a source of light having a wavelength of less than or equal to 193 nm, and (b) an optical element in a path of the light, having a first raster element, a second raster element, a third raster element and a fourth raster element situated thereon. The second raster element is adjacent to the first raster element, and located a first distance from the first raster element. The fourth raster element is adjacent to the third raster element, and located a second distance from the third raster element. The second distance is different from the first distance.
048195243	claims	1. An apparatus for remotely delivering and for aligning a pattern of screws over the thimble screw holes of a bottom nozzle of a fuel assembly, comprising a plate having a pattern of bores registrable with said screw holes, a sleeve of resilient material mounted over each bore for receiving, retaining and aligning the threaded end of a screw with one of said screw holes, means for mounting each resilient sleeve over its respective screw hole including a tubular member having an annular recess for receiving said sleeve of resilient material, alignment plate means registrable with pre-existing bores in the nozzle for aligning the bores in the plate with the screw holes in the nozzle, and a collar means connected to said plate for facilitating the remote handling and positioning of the plate over the nozzle of a fuel assembly. 2. An apparatus as defined in claim 1, wherein said sleeve is formed from urethane plastic having a hardness of between 60A and 75A durometers.
050248073	summary	FIELD OF THE INVENTION This invention relates to nuclear reactor fuel assemblies and in particular those assemblies which include spaced fuel rod support grids mounted in a reactor core as a unit. The fuel rods are held between an upper end fitting and a lower end fitting by means of spacer grids. The reactor coolant flows upwardly from holes in the lower end fitting along the fuel rod lower end caps and upwardly along the fuel rod cladding and through the spacer grids of the fuel assembly. BACKGROUND OF THE INVENTION Metallic debris in the coolant which collects or is trapped in fuel rod spacer grids adjacent the fuel containing cladding of the active region is believed responsible for a significant percentage of known fuel rod failures. Laboratory and in-reactor experience indicate that fuel rod cladding failures can be caused by debris trapped in a grid region which reacts against the fuel rod cladding in a vibratory fashion resulting in rapid wear of the cladding. The size and shape of the debris capable of causing severe damage is quite variable. In fact, metal fragments which can only be picked up with tweezers have been known to "drill" a hole in fuel rod cladding adjacent to a grid in less than 1,000 hours in a test simulating reactor operation. Since most cladding failures in reactors due to debris have occurred either within or below the lowermost spacer grid, a conventional grid apparently provides a rather effective screen for collecting debris. In order to prevent damage in the active area of the reactor, applicant set out to design an improved spacer grid structure for straining debris which: has a good probability of filtering out particles that could cause cladding damage; does not significantly increase fuel assembly uplift; will not jeopardize fuel rod support; will not hinder fuel assembly reconstitutability; will not significantly compromise fuel performance; has high mechanical integrity; is cost effective considering the risk/benefit; will not significantly infringe fuel rod plenum volume; and, does not require unplanned out-of-reactor flow testing. An earlier debris catching grid of Combustion Engineering, Inc., the assignee of the instant invention, was issued as U.S. Pat. No. 4,781,884 on Nov. 1, 1988. That invention was a separate debris catching strainer grid with no rod support function. A more traditional and prior art Combustion Engineering, Inc. design of fuel assembly (FIG. 1) has sustained a known distribution of debris-induced failures which shows clearly that the lowest spacer grid represents a very effective filter for debris. Unfortunately, the short lower end cap on the fuel rod of that fuel assembly ensures that the hollow cladding tube is adjacent to the trapped debris, and that any flow-induced motion of the debris can wear through the thin wall of the tube and create a failed rod. Based on available knowledge, conventional fuel from all vendors has experienced about the same distribution relationship between failures near the bottom of the assembly and those higher up. In the traditional or prior art fuel assembly, the lowest spacer grid is 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 needs to be located at an elevation where it will always laterally capture a "lifted" rod. Rods could potentially lift in response to coolant flow (FIG. 2) during abnormal conditions. Taking into account the known distribution of debris-induced failures mentioned above, one choice for a debris-resistant fuel assembly design is one in which the solid end cap is merely lengthened such that it extends up through the bottom spacer grid. This simplistic solution, however, has several drawbacks, as follows: a. zirconium alloy bar stock used for end caps is very expensive and, therefore, there is a strong incentive to minimize end cap length; and b. void volume within the fuel rod and/or the active fuel length will be affected negatively. SUMMARY OF THE INVENTION The present invention is an "egg crate" type spacer grid which is the lowermost spacer grid adjacent the lower end fitting. The grid is made up of "wavy" strips which provide bends or arched portions to cooperate with springs to provide support and positive capture for the fuel rods with a minimum of pressure drop across the grid. The solution provided by the "spring detent spacer grid" of the instant invention is to move the grid down, to thereby reduce the solid zirconium alloy material length required. To preclude the condition depicted in FIG. 2, the grid includes the provision of a fuel rod capturing spring detent device. This spring device engages a circumferential groove with tapered sides in the fuel rod end cap which creates 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, to further reduce the amount of potentially harmful debris passing by the first spacer grid (which typically has flow areas the same size as those of the spacer grids at higher elevations), integral leaves substantially symmetrically arranged on either side of the strip intersections have been added to greatly increase the likelihood that debris that passes the novel first or bottom "spring detent spacer grid" is too small to become trapped at a higher grid where it can damage the cladding of the active fuel region. The particular advantages of this spacer grid over other debris-catching concepts are: a. Compared to a separate screen-type grid: there is no cost of an extra component and there is a much lower incremental increase in pressure drop; b. Compared to a lower end fitting type filter (e.g., small holes in casting or integral screen): tests showed that with the spacer grid-type device there is a large amount of retained debris (probably because of tapered surfaces into which debris becomes "wedged") when the coolant flow is turned off, instead of with the other devices allowing debris to drop back into the region below the core for multiple attacks and chances to get through to the cladding. Tests also showed that the turbulence present in the region just below the lower end fitting caused the debris in many cases to "bounce" along the lower surface of the lower end fitting type filters and to eventually move over to the edge where it could pass up the gap between the fuel assemblies to get caught in a higher grid. With the "spring detent spacer grid", the debris passes through large lower end fitting holes, hits the grid edges, and even if it does not immediately become wedged, it is blocked from escaping into the assembly-to-assembly gap by the perimeter strip. c. Compared to a long end cap and conventional grid: there is less cost and less negative impact on stack length or rod void volume.
	claims	1. An electron beam apparatus having an electron analyzer, comprising:an illumination optical system consisting of lenses and deflection means for illuminating electrons at a specimen, the electrons being produced and accelerated from an electron gun;an imaging optical system for directing electrons transmitted through the specimen positioned within a magnetic field of an objective lens; andsaid electron analyzer having a detection system for detecting the imaged electrons and energy selection means for energy-dispersing the detected electrons and selecting electrons having a certain energy,wherein an accelerating voltage of the electron gun is varied to shift the detected energy of electrons and signals supplied to the lenses and deflection means of the imaging optical system are corrected for focus and position using amounts of correction each obtained by multiplying an energy shift value corresponding to a variation in the accelerating voltage by a corrective coefficient. 2. An electron beam apparatus having an electron analyzer as set forth in claim 1, wherein the corrective coefficients can be calibrated. 3. An electron beam apparatus having an electron analyzer as set forth in claim 1, wherein corrective coefficient KI of the lenses and corrective coefficients KDx and KDy of the deflection means are calculated based on equationsKI=(I2−I1)/(E2−E1)KDx=(IX2−IX1)/(E2−E1)KDy=(IY2−IY1)/(E2−E1)where I1 is the value of the current through the corrective lens and IX1, IY1 are the values of the current through the corrective deflection means when the energy shift is a first energy shift value of E1, I2 is the value of the current through the corrective lens, and IX2, IY2 are the values of the current through the corrective deflection means when the energy shift is a second energy shift value of E2. 4. An electron beam apparatus having an electron analyzer as set forth in claim 3, wherein lens-correcting value I when the energy shift assumes a value of E and correcting values IX and IY for the deflection means are found using equationsI=KI EIX=KDx E IY=KDy E. 5. An electron beam apparatus having an electron analyzer as set forth in claim 1, wherein the energy selection means for selecting electrons having a certain energy is an analyzer for energy-dispersing electrons by the use of a magnetic field. 6. An electron beam apparatus having an electron analyzer as set forth in claim 1, wherein the energy selection means for selecting electrons having a certain energy is an analyzer for energy-dispersing electrons by the use of an electric field. 7. A method of controlling lenses in an electron beam apparatus having an illumination optical system consisting of lenses and deflection means for illuminating electrons at a specimen, the electrons being produced and accelerated from an electron gun, an imaging optical system for imaging electrons transmitted through the specimen positioned within a magnetic field of an objective lens, and the electron analyzer having a detection system for detecting the imaged electrons and energy selection means for energy-dispersing the detected electrons and selecting electrons having a certain energy, said method comprising the steps of:varying an accelerating voltage of the electron gun to shift the detected energy of electrons; andcorrecting signals supplied to the lenses and deflection means of the illumination optical system using amounts of correction for focus and position each obtained by multiplying an energy shift value corresponding to a variation in the accelerating voltage by a corrective coefficient.
044406738	summary	FIELD OF THE INVENTION Our present invention relates to a method of treatment of radioactive waste water of the type which must be removed or discharged from time to time from nuclear electricity-generating power plants. BACKGROUND OF THE INVENTION Nuclear electricity-generating power plants of practically all types from time to time must dispose of radioactive waste water which can be derived from secondary or tertiary coolant cycles, from water in contact with contaminated materials or zones, or from the steam-generating system. In general the radioactive waste water which must be disposed of often contains solids, especially boric acid, which are in dissolved form. A conventional disposal technique is to store the radioactive waste water for a period sufficient to allow decay of some of the radioactive substances therein and then subject the stored water (with reduced radioactive level) to waste-water processing by any of a number of techniques including chemical precipitation or biological treatment. A disadvantage of this approach is the need to store relatively large quantities of water for long periods of time. It has also been proposed to concentrate the waste water and thereby reduce the volume of this substance which must be handled. In this conventional process, the waste water is concentrated by evaporation and the evaporation is carried out until the solids concentration in the water is at a level less than that which would represent a saturated solution at room temperature. The water is then stored for decay of radiation, e.g. for one half to three quarters of a year and then packaged, e.g. by incorporation in a solid mass, for permanent disposal and transportation. The permanent disposal may involve mixing the concentrated water with cement, (e.g. hydraulic cement) or incorporating the water in a hardenable bitumen or in a synthetic resin mass. In all cases the hardened material constitutes a leach-resistant body which can be sealed in a container, canister or drum with or without significant radiation-shielding capacity, the resulting package being given subterranean storage or being otherwise disposed of by techniques conventional in this art. The incorporation of the radioactive waste, whose activity has been reduced by long-term storage, in a hardenable mass prevents contamination of the environment in a particularly effective manner and the concentration step reduces significantly the volume of the material which must be handled in this manner. However, the degree of concentration is limited in the prior art process by the need to prevent the concentration of solids, during evaporation, from reaching the saturation concentration at room temperature, thereby ensuring that no solids will precipitate from the water and deposit in the system. The storage vessels which are commonly used for the radioactive decay process may have volumes of about 60 m.sup.3 and consequently, the cost of a storage facility for the interim storage of the waste water can be considerable and the operating cost of the power plant correspondingly high. OBJECTS OF THE INVENTION It is the principal object of the present invention to provide an improved method of treating radioactive waste water from a nuclear power plant whereby the disadvantages of earlier systems mentioned can be avoided. Another object of this invention is to provide an improved method of operating a nuclear power plant to minimize the cost and inconvenience heretofore encountered with long term large volume waste water storage. Yet another object is to provide a method of treatment of radioactive waste water from a nuclear power plant whereby the ratio of storage capacity to processed water can be reduced and the time between successive discharge of such water can be increased thereby improving the efficiency of the nuclear power plant. SUMMARY OF THE INVENTION These objects are attained, in accordance with the present invention, which provides a method in which the radioactive waste water of the nuclear electricity-generating power plant, containing soluble solids and especially boric acid, is concentrated by evaporation to a solids concentration above that which prevails in a saturated solution at room temperature and thereupon introducing the concentrate into a storage vessel in which the concentrate is cooled to cause precipitation of the solids. The solids are permitted to sediment (settle) from the liquid and liquid from which the solids have settled is decanted and recycled to the evaporator for further concentration. Eventually the sludge or slurry in the storage tank can be withdrawn and processed in the manner known in the art for the stored waste water although because of the repeated and cyclical concentration, is of smaller volume for a given amount of the starting material. The invention thus provides a significantly higher degree of concentration than has been contemplated heretofore and especially utilizes sludge or sediment formation to allow a smaller storage capacity to accommodate the material for long term storage in a power plant of a given output and/or allows a given storage capacity to process far more of the radioactive waste water originating in a plant than heretofore, thereby increasing the periods between discharges of the respective tanks. It is important to note that the contents of a given tank as part of the long-term storage or prior to the commencement thereof, consists of a waste water which has been concentrated several or many times by evaporation without any danger that there will be a deposit in the evaporator since each concentration is effected to a subsaturation level at the temperature in the concentrator (evaporator) or the lines thereof leading to the tank, but to a concentration above the saturation level at room temperature whereby precipitation of some solids from the concentrated liquid of each recycling is ensured. When attempts in the past have been made to increase the concentration of the liquid in the evaporator, deposits invariably formed within the evaporator or in the pipes leading therefrom. According to the invention, the process can be carried out as frequently as is necessary with recycling and until the entire storage vessel is filled with the sediment sludge up to the point at which the decantate is drawn off. According to a feature of the invention, the waste water concentrate is held at a temperature of at least 50.degree. C. from the point at which it enters an evaporator to the point at which it is discharged into the storage stage or vessel. It has been found to be advantageous, moreover, to introduce into the storage tank radioactive solids, especially diatomaceous earth which may be recovered from filters in which this filter aid is trapped, such solids being recovered from further treatment of the waste water or from treatments of the sump water of nuclear power plants. In general, such solids or even the sump waters themselves containing entrained solids, are introduced into the system separately from the waste water to be concentrated, e.g. directly in the storage stage with the decantate being concentrated in the manner described. This results in a more efficient utilization of the storage capacity. Surprisingly, these solids have the tendency to loosen the sedimented sludge and to keep the latter more flowable and lighter so that the sludge may be more readily handled. In addition, when diatomaceous earth, for example, is added to the sludge, the hydraulic cement serving as a hardening agent forms a mass which has greater stability than otherwise is the case and allows less of the portland cement to be used so that the disposal system is more cost efficient as well. The apparatus for carrying out the process of the present invention can comprise a storage vessel which is surmounted by an evaporator in which the concentration takes place and which opens downwardly into the storage vessel. At a point above the bottom of this vessel, a pipe opens into the storage vessel for recycling the decantate to the evaporator and this pipe may be united with a feed pipe through which the waste water is withdrawn from the nuclear reactor. A further pipe may open directly into the storage vessel for delivering the contaminated diatomaceous earth thereto. The collected sedimented sludge may be discharged from the storage tank by an immersion pump, a swirl lance or the like. Experiments have shown that the system of the present invention can, for tanks of the given storage capacity and for the same nuclear power plant, delay the need to discharge each tank and process the contents thereof to a period twice as long with the present invention than with the prior art system described in which only a single concentration to a point above the saturation level at room temperature is carried out. Since the residence time in the storage vessel can be increased, e.g. to two or more times the residence time heretofore, the discharge and further treatment costs can also be reduced.
063226103	abstract	Device and relative method to inject technological gases and solid material in powder form, in a furnace used in steel and metallurgical melting processes, comprising an emission element including an outlet mouth arranged at a certain distance from the liquid bath, said emission element being innerly shaped as a double nozzle, respectively first and second, the first nozzle having a convergent/divergent conformation defined by a neck and being predisposed for the emission of a jet of gas at supersonic speed, said second nozzle having at the end a conformation convergent towards the axis of the emission element and being pre-disposed, according to the step of the melting process in progress, for the emission of a jet at subsonic speed either of technological gases alone, such as for example oxygen, or of solid material in powder form on a gassy vehicle, or of liquid in atomized form.
059638842	summary	BACKGROUND OF THE INVENTION 1. Technical Field The present invention generally relates to a predictive maintenance system and, more particularly, to a predictive maintenance system using a local area network (LAN) to control, to acquire, and to analyze data from a plurality of remote data acquisition nodes. 2. Description of Related Art Various systems are known for analyzing vibration signatures of machines. One such prior art system uses a portable data collector having sensors which are connectable by a technician to a machine at one or more points to collect vibration data. The vibration signatures thus collected are downloaded to a personal computer, for example, which includes software for analyzing the vibration data. The software uses artificial intelligence, for example, to diagnose the risk of machine breakdown by first comparing the vibration signatures to earlier baseline signatures and then comparing any changes to a database of machine health-related characteristics and a knowledge base containing fault related scenarios. Since tiny changes in vibration data can indicate a machine fault, the software can predict the risk of breakdown. Such predictions can be used to shut down and/or effect repairs to machines before a catastrophic breakdown occurs, possibly resulting in reduced plant operations or, in some instances, plant shutdown. Also known are predictive maintenance systems which do not require a technician to walk through a plant or facility to collect data. Such systems use permanently mounted sensors which are connected to a system controller for collecting vibration data. However, while such so-called "on-line" data collection systems are available, they generally function to collect data only and cannot perform automatic real-time expert analysis as part of the system's functionality. In addition, such systems often have complicated architectures. Still further, these systems generally have a text and menu driven interface, making it difficult for a user to configure and use the system. SUMMARY OF THE INVENTION The present invention provides a predictive maintenance system for informing a user of the operating characteristics of machines. The system utilizes a local area network to control, to acquire, and to analyze data from remote data acquisition nodes applied to predictive maintenance. The local area network, as applied, conforms to industry standard and local area network protocols. The present invention provides an on-line data acquisition system which has greater responsiveness and speed than prior art on-line systems and provides for an almost unlimited size of highway, running at the speed of the data acquisition process for the slowest of the remote data stations. The network of the present invention also provides for parallel data acquisition on an almost unlimited number of data acquisition devices providing for a plurality (e.g., sixteen) sensor inputs from each device. The system of the present invention also includes a graphical interface that manages not only the navigation, but also the storing, retrieval, and analysis of machine information. This interface is constructed using graphical images obtained, for example, using digital cameras, photographic scanners, and/or software applications for creating graphical images. The interface uses these intuitive graphical images to help a user navigate through a complex system. The process of constructing the interface is performed entirely within the functionality of the system without the need to use external devices, external applications, or tools. A user may configure his/her own graphical interface by interfacing with a digital camera, scanner, or graphics software package. On-screen tools are used to customize the user's interface. At the time of constructing the interface, the data acquisition, storage, and analysis structure of the system are also created. The system of the present invention utilizes an expert system to interpret machinery data as it is collected and presents conclusions to the user based on the application of a knowledge base in a real-time manner without affecting the data acquisition cycle time. Part of the analysis is to perform data quality assessment and to return conclusions of both the data quality and data analysis processes. The features and advantages of the present invention will be better understood from a reading of the following detailed description in conjunction with the accompanying drawings.
054815754	summary	TECHNICAL FIELD The present invention relates to a method and a device for detecting oscillations in the core of a boiling-water nuclear reactor (BWR) comprising a plurality of neutron detectors, and wherein instability is detected on the basis of oscillations in the output signals of the neutron detectors. BACKGROUND ART A core in a nuclear reactor comprises a plurality of fuel assemblies. These are arranged vertically in the core in spaced relationship to each other. A fuel assembly comprises a plurality of vertical fuel rods, each of which contains a stack of pellets of a nuclear fuel, arranged in a cladding tube. The core is surrounded by water which serves both as coolant and as neutron moderator. The reactor core also comprises a plurality of control rods which, by being inserted into and withdrawn from the core, control the reactivity of the core and hence its output power. Thermohydraulic core instability is a well-known problem which may arise in a boiling-water nuclear reactor which is run with a high power and a low coolant flow. When cooling water is pumped upwards through the reactor core, steam is generated. The steam bubbles are in motion all the time, which produces variations in the water/steam ratio in the core. If these variations in the water flow are not damped in a natural way by friction, they may grow into sustained oscillations, that is, the reactor core has become unstable. Since water is a good moderator whereas steam is a poor moderator, these flow oscillations will also induce oscillations in the neutron flux. The oscillations may be of varying types and have different appearances and propagation. Certain oscillations start with a low amplitude and slowly grow larger and larger, others may be initiated by a temporary event, for example by a cooling pump stopping, and may then almost immediately assume its maximum amplitude and then continue to oscillate with a constant amplitude. The propagation of the oscillation may vary from global in-phase oscillations to local counter-phase and phase-shifted oscillations when some part of the core oscillates out of phase to the rest of the core. One form of intermittent oscillations have also been detected, that is, oscillations which move around in the core with a certain frequency. Oscillations in the neutron flux and the coolant flow may result in fixed margins for the fuel being exceeded, which in turn may lead to fuel damage. Instability in the reactor core must, therefore, be detected and acted upon. For measuring the neutron flux in boiling-water nuclear reactors (BWRs), neutron-sensitive detectors, so-called LPRM (Local Power Range Monitor) detectors, are used. A plurality of neutron detector tubes are arranged in spaced relationship to each other in the core, each of these tubes including four LPRM detectors located at four levels separated from each other in the vertical direction. The detectors form a regular lattice in the core. If the core becomes unstable and the neutron flux starts oscillating, this means that the output signals of the LPRM detectors start oscillating. U.S. Pat. No. 5,174,946 discloses a device and a method for determining whether the core is unstable, starting from the output signals from the LPRM detectors. The method and device described in the US patent specification determine whether instability exists starting from a combination of the output signals from a plurality of detectors. A disadvantage of using a combination of a plurality of output signals is that it may be difficult to detect local oscillations where only few (1-3) detectors oscillate. If only few detectors oscillate and these detectors are combined with other detectors which do not oscillate, the risk is great that the resultant signal does not oscillate to a sufficient extent to reveal the instability. The method and device described in the US patent specification detect only oscillations whose amplitude is growing, which means that an instability which causes an oscillation with a constant or slightly decreasing amplitude cannot be detected. The described method and device do not comprise any special detection of intermittent oscillations. SUMMARY OF THE INVENTION The invention aims to provide a device and a method for detecting oscillations in a reactor core. What characterizes the method and the device according to the invention will become clear from the accompanying claims.
	summary
	description	This invention was made with Government support under Contract No. DE-NE0000633 awarded by the Department of Energy. The Government has certain rights in this invention. In known pressurized water reactors (PWR) and boiling water reactors (BWR), a reactor core may contain a large number of fuel rods that are several meters in height. The reactor core may be surrounded by water contained within a reactor vessel. Additionally, the reactor may contain in-core instrumentation including a number of instrument assemblies located in the reactor core. During maintenance or refueling operations, in which some or all of the fuel rods in the reactor core may be inspected or replaced, respectively, the reactor vessel must be at least partially disassembled or removed in order to gain access to the reactor core. Prior to disassembling the reactor vessel, the in-core instrumentation may be disconnected and physically removed from the reactor core by opening the reactor vessel penetrations and pulling the in-core instrumentation out of the reactor core. However, in order to remove the in-core instrumentation, an operator and/or tool is typically introduced into the containment vessel in order to access the in-core instrumentation. For example, the containment structure may comprise a man-way that is large enough for an operator to enter a containment region located above the reactor pressure vessel. Work conditions and precautionary measures may be established to allow operators to position themselves on top of the reactor pressure vessel head to withdraw the in-core instruments. To withdraw an instrument, the operator may loosen a Swagelok fitting for each in-core instrument and physically grasp the external end of the in-core instrument, which may comprise a forty to eighty foot long tube or cable. The operator then pulls about fifteen feet of the in-core instrument through the reactor pressure vessel such that the lower end of the in-core instrument is withdrawn from the reactor core. Withdrawing the in-core instrumentation via known refueling operations may therefore not only require providing access to the inside of containment, but the refueling tool or operator may also need to be placed in close physical proximity to the reactor core in order to loosen or open the Swagelok fitting located on top of the reactor pressure vessel. Accordingly, two of the primary means of reducing potential radiation exposure, namely providing shielding from and maintaining distance to a radioactive source, may be compromised in known refueling operations. Alternatively, if the in-core instrumentation and reactor core are first allowed to cool down and/or become less radioactive before the operator or tool is used, then a significant amount of time may transpire in which the reactor module is taken off-line and is unable to generate electricity. This application addresses these and other problems. An in-core instrumentation system for a reactor module is disclosed herein. The in-core instrumentation system may comprise a plurality of in-core instruments connected to a containment vessel and a reactor pressure vessel at least partially located within the containment vessel. A reactor core may be housed within a lower head that is removably attached to the reactor pressure vessel, and lower ends of the in-core instruments may be located within the reactor core. The in-core instruments are configured such that the lower ends may be concurrently removed from the reactor core as a result of removing the lower head from the reactor pressure vessel. A method for withdrawing in-core instrumentation from a reactor module is disclosed herein. The method may comprise initiating a shut-down procedure for a reactor core located within a reactor pressure vessel. A sealed reactor module may be transported to a refueling pool. The sealed reactor module may comprise the reactor pressure vessel housed within a containment vessel, and in-core instrumentation may be at least partially located within the reactor core while the sealed reactor module is being transported. A lower containment head of the containment vessel may be removed in the refueling pool. Additionally, a lower head of the reactor pressure vessel may be removed in the refueling pool. In response to removing the lower head from the reactor pressure vessel, the method may comprise withdrawing the in-core instrumentation from the reactor core. A system for withdrawing in-core instrumentation from a reactor module is disclosed herein. The system may comprise means for performing a method similar to that described above. Various examples disclosed and/or referred to herein may be operated consistent with, or in conjunction with, one or more features found in U.S. Pat. No. 8,588,360, entitled Evacuated Containment Vessel for a Nuclear Reactor, U.S. Pat. No. 8,687,759, entitled Internal Dry Containment Vessel for a Nuclear Reactor, U.S. patent application Ser. No. 14/814,904, entitled Control Rod Position Indicator, and U.S. patent application Ser. No. 14/923,277, entitled Passive Cooling to Cold Shut-Down, the contents of which are incorporated by reference herein. FIG. 1 illustrates an example nuclear reactor module 100 with a dry and/or evacuated containment region 14. The nuclear reactor module 100 may comprise a reactor core 6 surrounded by a reactor pressure vessel 52. Primary coolant 10 in the reactor pressure vessel 52 surrounds the reactor core 6. Reactor pressure vessel 52 may be surrounded by a containment vessel 54. In some examples, containment vessel 54 may be located in a reactor pool 150. The reactor pool 150 may contain borated water stored below ground level. Containment vessel 54 may be at least partially submerged in the reactor pool 150. In some examples, at least a portion of the upper head of containment vessel 54 may be located above a surface 155 of the reactor pool 150 in order to keep any electrical connections and/or penetrations through the upper head dry. Additionally, containment vessel 54 may be configured to prohibit the release of any primary coolant 10 associated with reactor pressure vessel 52 to escape outside of containment vessel 54 into the reactor pool 150 and/or into the surrounding environment. Containment vessel 54 may be approximately cylindrical in shape. In some examples, containment vessel 54 may have one or more ellipsoidal, domed, or spherical ends, forming a capsule shaped containment. Containment vessel 54 may be welded or otherwise sealed to the environment, such that liquids and/or gases are not allowed to escape from, or enter into, containment vessel 54 during normal operation of reactor module 100. In various examples, reactor pressure vessel 52 and/or containment vessel 54 may be bottom supported, top supported, supported about its center, or any combination thereof. In some examples and/or modes of operation, an inner surface of reactor pressure vessel 52 may be exposed to a wet environment comprising the primary coolant 10 and/or vapor, and an outer surface of reactor pressure vessel 52 may be exposed to a substantially dry environment. The reactor pressure vessel 52 may comprise and/or be made of stainless steel, carbon steel, other types of materials or composites, or any combination thereof. The containment region formed within containment vessel 54 may substantially surround the reactor pressure vessel 52. Containment region 14 may comprise a dry, voided, evacuated, and/or gaseous environment in some examples and/or modes of operation. Containment region 14 may comprise an amount of air, a noble gas such as Argon, other types of gases, or any combination thereof. Additionally, the surfaces of one or both of reactor pressure vessel 52 and containment vessel 54 that bound containment region 14 may be exposed to water during certain modes of operation such as refueling, shutdown, or transport within the reactor pool 150. Containment region 14 may be maintained at or below atmospheric pressure, including a partial vacuum of approximately 300 mmHG absolute (5.8 psia) or less. In some examples, containment region 14 may be maintained at approximately 50 mmHG absolute (1 psia). In still other examples, containment region 14 may be maintained at a substantially complete vacuum. Any gas or gasses in containment vessel 54 may be evacuated and/or removed prior to operation of reactor module 100. During normal operation of reactor module 100, containment region 14 may be kept dry and/or evacuated of any water or liquid. Similarly, containment region 14 may be kept at least partially evacuated of any air or gases. A heat exchanger may be configured to circulate feedwater and/or steam in a secondary cooling system in order to generate electricity. In some examples, the feedwater passes through the heat exchanger and may become super-heated steam. The feedwater and/or steam in the secondary cooling system are kept isolated from the primary coolant 10 in the reactor pressure vessel 52, such that they are not allowed to mix or come into direct (e.g., fluid) contact with each other. The heat exchanger and/or associated piping of the secondary cooling system may be configured to penetrate through reactor pressure vessel 52 at one or more plenum 30. Additionally, the secondary piping may be routed to the upper region of containment above the level of the reactor pool 150, where the piping penetrates through containment vessel 54. By exiting containment above the reactor pool 150, the high temperature steam and feedwater lines do not loose heat to the reactor pool water 150. FIG. 2 illustrates the example nuclear reactor module 100 of FIG. 1, with a flooded or at least partially flooded containment region 14. During a normal, non-emergency shutdown, one or more steam generators may be configured to release steam and cool down the reactor module 100 from normal operating temperatures down to about 250° F. (121° C.). However, as the process of releasing steam may become somewhat ineffective at 250° F., the temperature of the reactor module may become essentially static or fixed the closer that it gets to the boiling temperature of the secondary coolant. The cool-down process may be augmented by at least partially flooding the containment region 14 of the example reactor module 100. In some examples, the containment region 14 may be flooded with borated water from the reactor pool 150 until the level of the water is at or above the height of a pressurizer baffle plate located within the reactor pressure vessel 52. During the cool-down process, water that enters containment region 14 is kept outside of reactor pressure vessel 52 and, similarly, all of the primary coolant 10 is kept within reactor pressure vessel 52. The upper head of the reactor pressure vessel 52 may be kept above the level of the water to avoid any connections that may pass through the upper head from being submerged in or otherwise exposed to water. In some examples, the predetermined level of the water within the containment region 14 may be associated with flooding the containment region 14 so that the majority of the reactor pressure vessel 52 is surrounded by the water. In other examples, the entire reactor pressure vessel 52 may be surrounded or submerged in the water that floods the containment region 14. The containment region 14 may be at least partially filled with water to initiate a passive cool-down process to a cold shutdown state, e.g., a shutdown state associated with primary coolant temperatures of less than 200° F. (93° C.). Once the containment region 14 is flooded above a predetermined level, no further action may be required, and the passive cool-down of the operating temperature to less than 200° F. may occur primarily as a function of natural circulation of the primary coolant 10 within the reactor pressure vessel 52, the shutdown reactor's decay heat, the transfer of heat from the primary coolant 10 to the water in the containment region 14, and the temperature of the reactor pool 150. During the cool-down process, an upper portion 16 of the containment region 14 may remain substantially dry and/or above the surface of the water contained therein. The pressure within upper portion 16 may be equalized to approximate atmospheric conditions as the reactor module reaches the shutdown state. A manway and/or release valve may be provided in the upper portion 16 of the containment region 14 to vent gases to atmosphere. In some examples, the manway and/or one or more valves may be configured to provide access to the containment region 14 for purposes of adding water. The pressure in the upper portion 16 may be controlled in order to maintain the level of water within the containment region 14 to a predetermined height within containment vessel 54. In examples where the reactor module 100 is configured to operate without any conventional thermal insulation being applied to the exterior of the reactor pressure vessel 52, heat may be readily transferred through the reactor vessel wall to the surrounding water in the containment region 14 during the cool-down process. FIG. 3 illustrates an example nuclear reactor module 300 comprising a reactor pressure vessel 320 housed within a partially disassembled containment vessel 340. A lower containment head 345 is shown removed from containment vessel 340. The removal of lower containment head 345 may be performed during refueling, maintenance, inspection, or other non-operational processes of reactor module 300. Containment vessel 340 may be removably attached to lower containment head 345 via an upper containment flange 342 and a lower containment flange 344. For example, a plurality of bolts may pass through and/or connect upper containment flange 342 to lower containment flange 344. Similarly, the bolts may be loosened and/or removed prior to removing lower containment head 345 from containment vessel 340. In-core instrumentation 330 is shown as being at least partially inserted into a reactor core 360 contained within reactor pressure vessel 320. In some examples, in-core instrumentation 330 may comprise twelve or more in-core instrument assemblies. Each in-core assembly may comprise a monitor, a sensor, a measuring device, a detector, other types of instruments, or any combination thereof. Additionally, the in-core assemblies may be attached to a number of wires or cables. The wires or cables associated with in-core instrumentation 330 may extend from an upper containment head 355 of containment vessel 340 down to reactor core 360. Upper containment head 355 may comprise one or more penetrations that are configured to allow in-core instrumentation 330 to be electrically coupled to wiring located outside of containment vessel 340. Lower containment head 345 may remain completely submerged below the surface 155 of a reactor pool, such as reactor pool 150 (FIG. 1) during the disassembly of containment vessel 340. While reactor pressure vessel 320 may remain intact and/or sealed during the disassembly of containment vessel 340, at least the lower portion of reactor pressure vessel 320 may also be surrounded by the reactor pool. FIG. 4 illustrates the example nuclear reactor module 300 of FIG. 3 comprising a partially disassembled reactor pressure vessel 320. A lower vessel head 325 is shown having been removed from the reactor pressure vessel 320, such as during refueling, maintenance, inspection, or other non-operational processes of reactor module 300 Reactor pressure vessel 320 may be removably attached to lower vessel head 325 via an upper vessel flange 322 and a lower vessel flange 324. For example, a plurality of bolts may pass through and/or connect upper vessel flange 322 to lower vessel flange 324. Similarly, the bolts may be loosened and/or removed prior to removing lower vessel head 325 from reactor pressure vessel 320. As a result of removing lower vessel head 325 from reactor pressure vessel 320, the in-core instrumentation 330 may be effectively withdrawn from the reactor core 360 as the lower vessel head 325 is being separated. Where in-core instrumentation 330 comprises multiple in-core instrument assemblies, all of the in-core instrument assemblies may be withdrawn from reactor core 360 substantially at the same time. In-core instrumentation 330 is shown as being at least partially protruding from or extending below the partially disassembled reactor pressure vessel 320 following the removal of lower vessel head 325. During a non-operational process, such as refueling, a visual inspection of the exterior of the reactor pressure vessel 320 and containment vessel 340 may be performed. Following the removal of lower containment head 345 and/or lower vessel head 325, remote inspection of the flanges and internal surfaces of the vessels may be performed while the vessels and/or lower heads are supported in one or more stands. In some examples, the remote inspections may comprise ultrasonic testing of key welds and full visual inspection of the internal surfaces. Additionally, some or all of the inspection may occur underneath the surface 155 of a reactor pool. In-core instrumentation 330 may remain connected to the top of containment vessel 340, and sealed by one or more pressurizer penetrations, as the reactor flanges are separated and lower vessel head 325 is removed from reactor pressure vessel 320. Each instrument assembly associated with in-core instrumentation 330 may be configured to slide out of their respective guide tubes in response to separating lower vessel head 325 from reactor pressure vessel 320. The withdrawal of in-core instrumentation 330 from the reactor core 360 and guide tubes may be accomplished without breaking the water-tight seal formed between containment vessel 340 and the surrounding pool of water. For example, the upper head of containment vessel 340 located at least partially above the surface 155 of the reactor pool may remain completely sealed to the surrounding environment during the disassembly of both the reactor pressure vessel 320 and the containment vessel 340, such that withdrawal of in-core instrumentation 330 from the guide tubes may be accomplished without providing any external access through the upper head of containment vessel 340. The guide tubes may be located in reactor core 360 and in some examples may extend up into a lower riser assembly 365 located above reactor core 360. In some examples, the in-core instrumentation 330 may be configured such that the lower ends are concurrently removed from both the lower riser assembly 365 and the reactor core 360 as a result of removing the lower head from the reactor pressure vessel 320. When in-core instrumentation 330 is clear of lower riser assembly 365, containment vessel 340 may be moved to a maintenance facility. On the other hand, lower vessel head 325 may be moved to a refueling bay, or remain behind without being moved, such that multiple operations may be performed on separated components of reactor module 300. During disassembly and transport of reactor module 300 and/or containment vessel 340, the lower ends of in-core instrumentation 330 may remain submerged in and surrounded by the reactor pool water at all times. The reactor pool water may operate to both reduce the temperature of in-core instrumentation 330 and provide a shield for any radiation which may be emitted from the lower ends. FIG. 5 illustrates a partial view of a nuclear reactor building 500 comprising equipment for assembling and/or disassembling a reactor module, such as reactor module 300 (FIG. 3). The equipment may comprise one or more stands located at the bottom of a containment pool or refueling bay. A first stand 510 may be configured to assemble and/or disassemble a containment vessel, such as containment vessel 340 (FIG. 3), after the reactor module has been shut down. During disassembly of the reactor module, a lower containment head 545 of the containment vessel may be placed in first stand 510. For example, a crane may be configured to transport the entire reactor module from a reactor bay and then lower the reactor module into first stand 510. After being placed in first stand 510, a containment flange associated with the lower containment head 545 may be de-tensioned by a containment tool 550, such as by loosening and/or removing a number of bolts. With lower containment head 545 decoupled from the containment vessel, the reactor module may be lifted from first stand 510 by the crane and placed in a second stand 520. With lower containment head 545 remaining behind in first stand 510, a lower vessel head 525 associated with a reactor pressure vessel may be placed in second stand 520. After being placed in second stand 520, a reactor vessel flange associated with lower vessel head 525 may be de-tensioned by a reactor pressure vessel tool 560, such as by loosening and/or removing a number of bolts. One or both of reactor pressure vessel tool 560 and containment tool 550 may be operated remotely. With lower vessel head 525 decoupled from the reactor pressure vessel, the reactor module may be lifted from second stand 520 by the crane and moved to a maintenance facility. Additionally, the lower vessel head 525 may be moved separately from the reactor module, or lower vessel head 525 may be refueled and/or maintenance work performed while being held in second stand 520. In some examples, the refueling bay containing reactor pressure vessel tool 560 and containment tool 550 may comprise a rectangular area approximately sixty feet long by thirty feet wide. The floor of the refueling bay may be at elevation twenty feet, and covered by seventy five feet of water. In some examples, the refueling bay floor may be approximately six feet below the bottom of pool for the balance of the facility. An inspection of the inner and outer surfaces of lower vessel head 525 and lower containment vessel 545 may be performed following the partial disassembly of the reactor module. Additionally, the exposed core support assembly and lower riser assembly may also be inspected. The inspection of the vessel features may include visual, volumetric, ultrasonic, and/or other inspection techniques. The inspections may be performed during the refueling process of the reactor module. A visual examination may be conducted to detect discontinuities and imperfections on the surface of components, including such conditions as cracks, wear, corrosion, or erosion. Additionally, the visual examination may be conducted to determine the general mechanical and structural condition of components and their supports by verifying parameters such as clearances, settings, and physical displacements, and to detect discontinuities and imperfections, such as loss of integrity at bolted or welded connections, loose or missing parts, debris, corrosion, wear, or erosion. A volumetric examination may indicate the presence of discontinuities throughout the volume of material and may be conducted from either the inside or outside surface of a component. The volumetric examination may comprise remotely deployed ultrasonic devices for examination of code identified vessel welds. A lower vessel inspection tree (LVIT) may comprise operating control console and cabling used to perform visual and ultrasonic testing of surfaces and features within lower vessel head 525 and lower containment vessel 545. An LVIT may be installed at or near the lower vessel sections in the bottom of the refueling pool. One or more LVITs may be used to locate, monitor, and report the position of inspection elements, and to acquire data that is transmitted back to the control console. The installation of the LVIT on the lower vessels may be performed remotely using a reactor building crane with the wet hoist attached. Additionally, in-pool cameras may be used by the crane operator to control crane motion and load placement. FIG. 6 illustrates a nuclear power building 600 comprising a plurality of reactor modules, such as a reactor module 610 and an additional reactor module 620. Nuclear power building 600 is shown as including twelve reactor modules by way of example only, and fewer or more reactor modules per nuclear power building are contemplated herein. Nuclear power building 600 may comprise an overhead crane 655 configured to move or transport the plurality of reactor modules. In the illustrated example, reactor module 610 has been removed from a reactor bay 630 and is in the process of being transported through a shared reactor building passageway 650. The passageway 650 may be fluidly connected to each of the reactor bays, such as reactor bay 630, allowing reactor module 610 to be transported by crane 655 while being at least partially submerged under water. Passageway 650 may fluidly connect reactor bay 630 to a spent fuel pool 680 and/or to a dry dock 690. Additionally, the passageway 650 may fluidly connect reactor bay 630 to a refueling bay 665 containing a containment vessel stand 660 and a reactor pressure vessel stand 670. In some examples, containment vessel stand 660 and reactor pressure vessel stand 670 may be configured similarly as first stand 510 and second stand 520 illustrated in FIG. 5, and may include a containment assembly/disassembly tool and a reactor pressure vessel assembly/disassembly tool, respectively. By including a plurality of reactor modules, reactor module 610 may be taken off-line for purposes of refueling and/or maintenance while the remaining reactor modules continue to operate and produce power. In a nuclear power facility comprising twelve reactor modules, where each reactor module has a designed fuel life of two years, a different reactor module may be refueled every two months as part of a continuous refueling cycle. For reactor modules having longer designed fuel lives, the reactor modules may be refueled less frequently. An LVIT 640 may be configured to enter nuclear power building 600 through an opening or door for purposes of conducting visual and/or ultrasonic inspections of the reactor modules. In some examples, LVIT 640 may be moved within nuclear power building 600 by crane 655. After the LVIT 640 has been placed by or near the vessel to be inspected, crane 655 may be disengaged from the LVIT 640, freeing crane 655 to perform other operations in support of the refueling outage while the inspections are conducted. Once the inspection is completed, crane 655 may be used to remove the LVIT 640 from the vessel that was inspected. LVIT 640 may be configured to inspect one or both of a reactor pressure vessel and a containment vessel. In some examples, two or more LVITs may operate concurrently to inspect the reactor pressure vessel and the containment vessel, providing the ability to perform multiple inspections at the same time. Providing duplicate and/or redundant inspection devices may reduce the amount of equipment necessary to complete the reactor module inspections, allow concurrent inspections of multiple reactor modules, and/or provide the ability to use either inspection device as a spare in the event of equipment failure. FIG. 7 illustrates an example nuclear reactor module 700 with at least a portion of the in-core instrumentation 730 withdrawn into a containment vessel 740. A lower containment head has been removed from containment vessel 740, such that a reactor pressure vessel 720 at least partially housed within containment vessel 740 may be accessed below a lower containment flange 742. Similarly, a lower vessel head has been removed from reactor pressure vessel 720 in the illustrated example. With both of the lower heads removed from reactor pressure vessel 720 and containment vessel 740, a lower reactor pressure vessel flange 722 may be located beneath a surface 755 of a pool of water. In other examples, both the lower reactor pressure vessel flange 722 and the lower containment flange 742 may be located beneath the surface 755. The surface 755 associated with the pool of water may be located within a dry dock, such as dry dock 690 (FIG. 6). In some examples, the level of surface 755 may be adjusted when the reactor module 700 is located at the dry dock in order to provide access to one or more components, such as a steam generator. In still other examples, the position of reactor module 700 may be lowered or raised to adjust the relative level of surface 755. In-core instrumentation 730 may be electrically coupled to an upper containment head 745 of containment vessel 740. A connection device 780 may provide a sealed penetration through upper containment head 745. External wiring 785 may be operably coupled to in-core instrumentation 730 via the connection device 780. In some examples, connection device 780 may comprise a two-part connector configured to attach to both in-core instrumentation 730 and external wiring 785. Additionally, in-core instrumentation 730 may be routed through a sealed penetration 760 of an upper head 725 of reactor pressurizer vessel 720. The withdrawal of in-core instrumentation 730 through sealed penetration 760 and into containment vessel 740 may operate to withdraw a lower end 735 of in-core instrumentation 730 into reactor pressure vessel 720. The withdrawal process may be initiated after the temperature of the reactor coolant and/or the reactor pressure vessel have decreased to a threshold cooling temperature, e.g., by the transfer of heat to the surrounding pool water. While the reactor coolant and/or the reactor pressure vessel are being cooled down, refueling and other maintenance operations may be performed on other components, such as the reactor core which has been separated from reactor pressure vessel 720. In some examples, such as where in-core instrumentation 730 comprises a plurality of in-core instrument assemblies, all of the lower ends 735 of the instrument assemblies may be withdrawn into reactor pressure vessel 720 at the same time. In other examples, each of the instrument assemblies may be separately withdrawn into reactor pressure vessel 720. An access portal 770 may be provided in upper containment head 745. Access portal 770 may be configured to provide access for an operator and/or a tool to enter containment vessel 740 for purposes of withdrawing in-core instrumentation 730. For example, the tool may comprise a pole and grasping device configured to attach to a portion of in-core instrumentation 730 located near or some distance above sealed penetration 760, in order to pull the portion of in-core instrumentation 730 into upper containment head 745. In other examples, the upper portion of in-core instrumentation 730 may be pulled up through a containment penetration provided at or near connection device 780, so that the upper portion of in-core instrumentation 730 may be pulled outside of containment vessel 740 without providing access through access portal 770. Sealed penetration 760 may comprise a Swagelok fitting. The fitting may be configured to retain the position of in-core instrumentation 730 at a fixed position. In some examples, sealed penetration 760 may be loosened to allow the withdrawal of the upper portion of in-core instrumentation 730 into containment vessel 740. Once the in-core instrumentation 730 has been withdrawn, sealed penetration 760 may be tightened to again fix the position of lower ends 735 within reactor pressure vessel 725. A Swagelok tool may be inserted into the upper containment head 745 of containment vessel 740 through access portal 770. In other examples, the sealed penetration 760 may be automatically loosened and tightened during different stages of the disassembly operation, without requiring access into the containment vessel 740. FIG. 8 illustrates an example an in-core instrumentation system 800 for a nuclear reactor. After performing a refueling operation and/or other maintenance activity, the reactor module may be prepared for reassembly of the reactor pressure vessel and containment vessel so that the reactor module may be placed back on line. For example, a lower head containing a reactor core with new fuel rods may be reattached to a reactor pressure vessel 820. Following the reattachment of the lower head to reactor pressure vessel 820, in-core instrumentation 830 may be reinserted into the replenished reactor core and/or into the corresponding guide tubes. In-core instrumentation 830 may comprise relatively flexible cabling suspended from a containment vessel connection 880. The instrumentation cabling may be routed through relatively rigid instrumentation sheathing 855. An upper end of instrumentation sheathing 855 may be supported by a bracket 850. Bracket 850 may be configured to stabilize the upper ends of sheathing 855 and, in some examples, may provide a means of simultaneously withdrawing in-core instrumentation 830 through an instrumentation position control device 860. Instrumentation position control device 860 may provide for a sealed penetration through an upper head of reactor vessel 820. Similar to the example nuclear reactor module 700 illustrated in FIG. 7, an upper portion of in-core instrumentation 830 may have been withdrawn into the containment vessel. The withdrawal of in-core instrumentation 830 may be accomplished by raising bracket 850. In some examples, the insertion of in-core instrumentation 830 into the reactor core and/or guide tubes may be accomplished by lowering bracket 850. In some examples, bracket 850 and/or instrumentation sheathing 855 may provide a guide by which in-core instrumentation 830 may be threaded or inserted down into the reactor core. Instrumentation sheathing 855 may comprise a threaded portion 835. The threaded portion 835 may be substantially the same length as the length of in-core instrumentation 830 that is withdrawn from reactor pressure vessel 820 into the containment vessel. Instrumentation position control device 860 may comprise one or more threaded gears and/or motors that may be configured to raise or lower in-core instrumentation 830 via a threaded engagement with the threaded portion 835 of instrumentation sheathing 855. Instrumentation position control device 860 may be remotely actuated to control the position of in-core instrumentation 830. Additionally, instrumentation position control device 860 may be remotely sealed or unsealed. FIG. 9 illustrates an example system 900 associated with withdrawing and/or inserting in-core instrumentation into a reactor core. A plurality of in-core instruments 930 may be connected to a containment vessel 940. In some examples, in-core instruments 930 may be suspended from an upper head of the containment vessel 940 at instrument connection 980. A reactor pressure vessel 920 may be at least partially located within the containment vessel 940. During full power operation of the reactor module, reactor pressure vessel 920 may be entirely housed in a sealed containment region within containment vessel 940. During the initial stages of a refueling operation, in which a lower head of containment vessel 940 may be removed in order to access the internal components of the reactor module including the reactor core 990, a portion of reactor pressure vessel 920 may be partially located outside of containment vessel 940. For example, a lower head of the reactor pressure vessel 920 may be exposed to a surrounding pool of water below the containment vessel 940. The reactor core 990 may be housed within a lower vessel head that is removably attached to the reactor pressure vessel 920. With the lower vessel head attached to the reactor pressure vessel 920, the lower ends of in-core instruments 930 may be located within the reactor core 990 which is housed in the lower vessel head of reactor pressure vessel 920. Additionally, in-core instruments 930 may pass through a vessel penetration 960 located in an upper vessel head of reactor pressure vessel 920. The in-core instruments 930 may be configured such that the lower ends are concurrently removed from the reactor core 990 as a result of removing the lower vessel head from the reactor pressure vessel 920. The lower ends of in-core instruments 930 may be removed from the reactor core 990 without unsealing the vessel penetration 960. Additionally, in examples in which the upper vessel head of the containment vessel 940 is environmentally sealed, the lower ends of in-core instruments 930 may be removed from the reactor core 930 without unsealing the upper vessel head of containment vessel 940. The containment vessel 940 may be at least partially submerged in a surrounding pool of water. As a result of removing the lower vessel head from the reactor pressure vessel 920, the lower ends of the in-core instruments 930 may be exposed to the pool of water. In examples where the lower vessel head is removably attached to the reactor pressure vessel 920 at a vessel flange, the exposed lower ends of the in-core instruments 930 may extend several meters below the vessel flange in the pool of water. The reactor module may be transported to a maintenance bay and/r refueling bay while the exposed lower ends of the in-core instruments 930 extend below the vessel flange into the pool of water. Additionally, an upper portion of the in-core instruments 930 may be withdrawn from the reactor pressure vessel 920 into the containment vessel 940 while the reactor pressure flange remains submerged in the pool of water. A reactor controller 970 may be configured to monitor the temperature of the in-core instruments 930. Reactor controller 970 may comprise a sensor, a gauge, a thermometer, a thermocouple, other means of monitoring temperature, or any combination thereof. Additionally, reactor controller 970 may be configured to monitor, measure, detect, read, sense, estimate, or otherwise determine the temperature associated with the reactor pressure vessel 920. Reactor controller 970 may be configured to raise and/or lower in-core instruments 930 in response to determining that the temperature associated with the in-core instruments 930 has reached a threshold cooling temperature. FIG. 10 illustrates an example process 1000 of refueling a nuclear reactor module. The reactor module may comprise a reactor vessel housed within a containment vessel. The containment vessel may at least partially surround the reactor pressure vessel by a containment region. The containment region may be evacuated of liquid and/or air during normal operation of the reactor module. Additionally, the containment vessel may be at least partially submerged in a reactor pool. At operation 1010, a reactor shut-down or other type of maintenance activity may be initiated. For example, a plurality of control rods may be inserted into the reactor core. At operation 1020, the sealed reactor module may be transported to a refueling pool. The reactor module may comprise a reactor pressure vessel housed within a containment vessel. In-core instrumentation may be at least partially located within the reactor core while the sealed reactor module is being transported. At operation 1030, a lower containment head of the containment vessel may be removed in the refueling pool. The lower containment head may be removed by placing the reactor module in a first stand and then loosening a plurality of bolts connecting the lower containment head to the containment vessel. The containment vessel may then be lifted off of the lower containment head while the lower containment head remains fixed in the first stand. At operation 1040, a lower head of the reactor pressure vessel may be removed in the refueling pool. The lower head of the reactor pressure vessel may be removed by placing the reactor module in a second stand and then loosening a plurality of bolts connecting the lower head of the reactor pressure vessel to the reactor pressure. The reactor pressure vessel may then be removed from the lower head while the lower head remains fixed in the second stand. The in-core instrumentation may be withdrawn from the reactor core together with, or as a result of removing, the lower vessel head from the reactor pressure vessel. In some examples, the in-core instrumentation is withdrawn from the reactor core after the lower vessel head is disconnected from a reactor pressure vessel flange. The in-core instrumentation may extend below the reactor pressure vessel flange in the refueling pool after the lower vessel head has been removed. At operation 1050, the temperature of the reactor coolant and/or the reactor pressure vessel may be allowed to cool down. During the cool down period, the reactor core may be separately processed for refueling. At operation 1060, at least a portion of the in-core instrumentation may be withdrawn from the reactor pressure vessel into the containment vessel after the lower head has been removed from the reactor pressure vessel. Operation 1060 may be performed in a maintenance facility, such as a maintenance bay. The maintenance bay may be fluidly connected to a refueling bay, such as by a shared waterway of a reactor building At operation 1070, the reactor core may be refueled. In some examples, the reactor core may be refueled in the refueling bay while the portion of the in-core instrumentation is withdrawn from the reactor pressure vessel. Additionally, the reactor core may be refueled while the in reactor coolant and the reactor pressure vessel are allowed to cool down at operation 1050. At operation 1080, the lower head of the reactor pressure vessel may be reattached to the reactor module after the reactor core has been refueled. At operation 1090, the in-core instrumentation may be inserted into the replenished reactor core while returning the portion of the in-core instrumentation from the containment vessel back into the reactor pressure vessel. At operation 1110, the reactor module, with the in-core instruments having been inserted into the reactor core, may be transported to a reactor bay after the reactor module has been environmentally sealed by reattaching the lower containment head to the containment vessel. In other examples, the insertion of the in-core instrumentation at operation 1090 may occur after transporting the reactor module to the reactor bay at operation 1110. One or more example systems described herein may comprise various nuclear reactor technologies, and may comprise and/or be used in conjunction with nuclear reactors that employ uranium oxides, uranium hydrides, uranium nitrides, uranium carbides, mixed oxides, and/or other types of fuel. Although the examples provided herein have primarily described a pressurized water reactor and/or a light water reactor, it should be apparent to one skilled in the art that the examples may be applied to other types of power systems. For example, the examples or variations thereof may also be made operable with a boiling water reactor, sodium liquid metal reactor, gas cooled reactor, pebble-bed reactor, and/or other types of reactor designs. Additionally, the examples illustrated herein are not necessarily limited to any particular type of reactor cooling mechanism, nor to any particular type of fuel employed to produce heat within or associated with a nuclear reaction. Any rates and values described herein are provided by way of example only. Other rates and values may be determined through experimentation such as by construction of full scale or scaled models of a nuclear reactor system. Having described and illustrated various examples herein, it should be apparent that other examples may be modified in arrangement and detail. We claim all modifications and variations coming within the spirit and scope of the following claims.
	abstract	An electro-optic window is provided, together with a method of manufacturing the window. The window (3) is made of a material substantially transparent to at least one of infra-red, visible and UV radiation and treated to have reduced RF/MICROWAVE transmission characteristics by the provision of a grid (1) set into at least one surface (2) thereof. The grid (1) is formed of a material selected to be either reflective or absorptive to RF/MICROWAVE radiation.
	description	This is a Continuation-In-Part Application of application Ser. No. 16/456,587 filed Jun. 28, 2019, entitled EMERGENCY COOLING WATER SYSTEM FOR A FLOATING NUCLEAR REACTOR. The invention of Applicant's earlier patent application Ser. No. 15/807,182 entitled FLOATING NUCLEAR REACTOR PROTECTION SYSTEM relates to a floating nuclear power reactor. More particularly, the invention of the '182 application relates to a floating nuclear power reactor including a barge which is floatably positioned in the interior of a large water-filled tank or body of water and wherein the nuclear power reactor is positioned on the barge. Even more particularly, the invention of the '182 application relates to a protection system for a floating nuclear power reactor to protect the nuclear reactor from an aircraft strike or a missile strike. Additionally, the protection system of the invention of the '182 application includes structure to reduce the impact forces of an aircraft strike or a missile strike. The instant invention relates to structure to maintain the temperature and water level of the water-filled tank in which the barge of the floating nuclear power reactor is floating. Applicant has received U.S. Pat. Nos. 9,378,855; 9,396,823; and 9,502,143 relating to nuclear reactors positioned in a body of water to be able to flood and cool the nuclear reactor in the event of overheating or over pressurization of the nuclear reactor. In Applicant's latest invention shown and described in the co-pending application Ser. No. 15/807,049 filed Nov. 8, 2017, a suspension system is described for suspending and stabilizing a barge which is floating in a large water tank. That system is incorporated herein which further enhances the protection of the nuclear reactor in the event of an aircraft strike or a missile strike. The invention of Applicant's earlier '182 application provides a protection system for the nuclear power reactor of the co-pending application and to provide protection to other exposed nuclear power reactors of different designs. The prior application and patents of Applicant do not have any means to maintain the temperature and water level of the water-filled tank. This Summary is provided to introduce a selection of concepts in a simplified form that are further described below in the Detailed Description. This Summary is not intended to identify key aspects or essential aspects of the claimed subject matter. Moreover, this Summary is not intended for use as an aid in determining the scope of the claimed subject matter. The structure of the floating nuclear reactor of the '182 application will first be disclosed. The floating nuclear reactor of the '182 patent application includes a tank, which may be rectangular, having a bottom wall, an upstanding first end wall, an upstanding second end wall, an upstanding first side wall and an upstanding second side wall. Each of the first end wall, the second end wall, the first side wall and the second side wall of the tank have an outer side, an inner side, a lower end and an upper end. The tank is partially or fully buried in the ground with the tank having water therein. A barge is floatably positioned in the tank with the barge having a bottom wall, a first end wall, a first side wall, a second side wall and an open second end. A nuclear reactor is positioned on the barge. At least one suspension assembly, and preferably a plurality of suspension assemblies, connect the first end wall of the barge to the first end wall of the tank. At least one suspension assembly, and preferably a plurality of suspension assemblies, connect the first side wall of the tank to the first side wall of the barge. At least one suspension assembly, and preferably a plurality of suspension assemblies, connect the second side wall of the tank to the second side wall of the barge. At least one suspension assembly, and preferably a plurality of suspension assemblies, connect the second end wall of the barge to the second end wall of the tank. The suspension assemblies permit the barge to move upwardly and downwardly with respect to the tank while maintaining the barge in a level condition. The suspension assemblies permit the barge to move downwardly if struck by a missile or aircraft to lessen the impact thereof. The nuclear reactor of the '182 patent application is positioned in the tank so as to close the open second end of the barge. The nuclear reactor includes a first containment member which has a cylindrical body portion, a hemi-spherical upper end and a hemi-spherical lower end. The first containment member is comprised of stainless steel or other suitable material. The first containment member is positioned at the open end of the barge with the sides of the containment member being in engagement with the ends of the sidewalls of the barge so as to close the open end of the barge. The positioning of the first containment member causes the outer side of the first containment member to be in contact with the water in the tank. The first containment member defines a sealed interior compartment. The first containment member has a hatch or door mounted thereon at the lower end thereof which selectively closes an opening in the first containment member. The first containment member also has a pipe extending from the lower end thereof which is in fluid communication with the interior compartment thereof. A normally closed one-way valve is imposed in the pipe. A reactor vessel is positioned in the interior compartment of the first containment member. The nuclear reactor of the co-pending application has a unique cooling system for the nuclear reactor which does not form a part of this invention. A heat exchanger is positioned adjacent the first containment member and includes a body section, an upper section and a lower section. The heat exchanger includes an outer wall member or second containment member which is comprised of metal. A vessel is positioned within the second containment member of the heat exchanger. The vessel has an interior compartment which is filled with fluid. The heat exchanger is connected to a turbine or other device. The suspension assemblies also permit the barge to move downwardly in the tank in the event of an aircraft strike, a missile strike or an earthquake to reduce the impact forces on the barge and nuclear reactor. A hollow steel conical-shaped member is mounted on the upper end of the first containment member of the nuclear reactor. A hollow steel conical-shaped member is also mounted on the upper end of the heat exchanger. If an aircraft or a missile should strike either of the conical-shaped members, the conical-shaped members would cause the disintegration of the aircraft or missile and would deflect the same. A roof is positioned over the upper end of the barge which hides the location of the nuclear reactor and heat exchanger from view so that an aircraft attempting to strike either the nuclear reactor or the heat exchanger will not know the precise position of those structures on the barge. In the preferred embodiment of the invention of the '182 application, the interior of the conical-shaped members on the upper ends of the nuclear reactor and the heat exchanger will be filled with a material which acts as an impact absorbing member. The instant invention provides means for maintaining the temperature and water level of the water-filled tank in which the barge floats. It is therefore a principal object of the invention to provide a floating nuclear reactor. A further object of the invention is to provide structure for maintaining the temperature and water level of the water-filled tank in which the barge of the nuclear reactor floats. A further object of the invention is to provide cooling water to the condenser and to drain hot water from the condenser. These and other objects will be apparent to those skilled in the art. Embodiments are described more fully below with reference to the accompanying figures, which form a part hereof and show, by way of illustration, specific exemplary embodiments. These embodiments are disclosed in sufficient detail to enable those skilled in the art to practice the invention. However, embodiments may be implemented in many different forms and should not be construed as being limited to the embodiments set forth herein. The following detailed description is, therefore, not to be taken in a limiting sense in that the scope of the present invention is defined only by the appended claims. Applicant has previously received U.S. Pat. Nos. 9,378,855; 9,396,823; and 9,502,143 relating to floating nuclear power reactors. Applicant incorporates the disclosure of the above identified patents in their entirety by reference thereto to complete this disclosure if necessary. Applicant also incorporates the disclosure of application Ser. No. 15/807,182 filed Nov. 8, 2017 in its entirety by reference thereto to complete this disclosure if necessary. The floating nuclear reactor of the invention of the '182 application, as seen in FIGS. 1-7, is referred to generally by the reference numeral 10. The nuclear reactor 10 floats in a concrete tank 12 having a bottom wall 14, a first end wall 16, a second end wall 18, a first side wall 20, a second side wall 22 and an open upper end 24. Tank 12 is buried in the ground 26 as seen in FIG. 1 so that the open upper end 24 of tank 12 is at or above ground level 28. The tank 12 is partially filled with water 30 from a source of water. Preferably the water 30 is gravity fed to the tank 12. The tank 12 may be completely buried in the ground. The numeral 32 refers to a barge-like vessel which floats in the tank 12. Barge 32 includes a bottom wall 34, a first side wall 36, a second side wall 38, a semi-circular end wall 40 and an open end 41 at the ends 42 and 43 of side walls 36 and 38 respectively. Barge 32 is comprised of a metal material such as stainless steel, steel, iron, aluminum or other suitable material. Barge 32 is supported in tank 12 by a plurality of upper suspension assemblies 44, 46, 48, 50, 52, 54, 56 and 58 which extend between the barge 32 and the tank 12 as will be described in detail hereinafter. Barge 32 is also supported in tank 12 by a plurality of lower suspension assemblies, identical to suspension assemblies 44, 46, 48, 50, 52, 54, 56 and 58, which are positioned below suspension assemblies 44, 46, 48, 50, 52, 54, 56 and 58. The numeral 59 refers to a nuclear reactor which is positioned in barge 32 so as to close the open end 41 of barge 32 as will be explained in detail hereinafter. Reactor 59 includes an upstanding containment member 60 which has a cylindrical body portion 62, a hemi-spherical upper end 64 and a hemi-spherical lower end 66. Containment member 60 is comprised of stainless steel or other suitable material. Containment member 60 is positioned at the open end 41 of barge 32 with the sides of containment member 60 being in engagement with the ends 42 and 43 of side walls 36 and 38 respectively of barge 32 and being secured thereto by welding or the like to close the open end 41 of barge 32. The positioning of the containment member 60 as just described causes the outer side of containment member 60 to be in contact with the water 30 in tank 12. Containment member 60 defines a sealed interior compartment 68. Containment member 60 has a hatch 70 mounted therein as seen in FIG. 3. Containment member 60 also has a pipe 72 extending from the lower end thereof which is in fluid communication with the interior compartment 68. A normally closed one-way valve 74 is imposed in pipe 72. A reactor vessel 75 is positioned in compartment 68 and has an interior compartment 76. Vessel 75 is supported in compartment 68 by braces 77 which extend between the exterior of reactor vessel 75 and the interior side of containment member 60 as seen in FIG. 3. The numeral 80 refers to an upstanding heat exchanger which is positioned adjacent containment member 60 as seen in the drawings. Heat exchanger 80 includes a body section 82, an upper section 84 and a lower section 86. Heat exchanger 80 is comprised of a metal material such as stainless steel or other suitable material. A vessel 88 is positioned within heat exchanger 80 and is supported therein by braces 90 extending therebetween. Vessel 88 defines an interior compartment 92. A tube 93 interconnects the reactor vessel 75 and the vessel 88 of heat exchanger 80 as seen in the drawings. The heat exchanger 80 is connected to a turbine 96, or other device, by tube 94, which is connected to a generator 98 or other structure. A hollow metal cone 100 is mounted on the hemi-spherical upper end 64 of containment member 60. Cone 100 is comprised of stainless steel, steel or other suitable material. Cone 100 has an interior compartment 102 which is preferably filled with a filter material 104 which not only may serve as a filtration bed but serves as an impact absorber should the cone 100 be struck by an aircraft or a missile. The cone 100, if struck by an aircraft or missile, will disintegrate or tear apart the aircraft or missile and deflect the aircraft or missile away from the cone 100. An outlet pipe 106 may be provided in the upper end of containment member 60 to permit steam or the like to pass upwardly therethrough onto the filtration material 104. The cone 100 may also have a discharge tube assembly 108 extending upwardly from pipe 106 and which has discharge tubes 110 extending downwardly and outwardly from the upper end of tube 108 as seen in FIG. 5. A metal cone 112 extends upwardly from the upper end of heat exchanger 80 and is filled with an impact absorbing material 114. Cone 112, if struck by an aircraft or missile, will disintegrate the aircraft or missile in the same manner as the cone 100. A roof 116 extends over the cones 100, 112 and the barge 32 to hide the reactor 59 and the heat exchanger 80 from view. Thus, if an aircraft is attempting to strike the reactor 59, the pilot of the aircraft will not be able to determine the exact location of the reactor 59. A pair of vertically disposed guide tracks or channels 120 and 122 are secured to the inner side of end wall 18. A pair of vertically disposed guide tracks or channels 124 and 126 are secured to the inner side of side wall 20. A pair of vertically disposed guide tracks or channels 128 and 130 are secured to the inner side of end wall 16. A pair of vertically disposed guide tracks or channels 132 and 134 are secured to the inner side of side wall 22. Each of the guide tracks 120, 122, 124, 126, 128, 130, 132 and 134 have an upper wheel and a lower wheel vertically movable therein. The guide tracks 120, 122, 124, 126, 128, 130, 132 and 134 form a part of the suspension assemblies 46, 48, 50, 52, 54, 56, 58 and 44 respectively. Inasmuch as the suspension assemblies 44, 46, 48, 50, 52, 54, 56 and 58 are identical except for length, only suspension assembly 48 will be described in detail. Suspension assembly 48 includes an upper chain member 136, a lower chain member 138 and an intermediate chain member 140. The outer ends of chain members 136, 138 and 140 are secured to the upper wheel in guide track 122. The inner ends of chain members 136, 138 and 140 are secured to the barge 32. As seen, upper chain member 136 extends upwardly and inwardly from guide track 122 to barge 32. As also seen, lower chain member 138 extends downwardly and inwardly from guide track 122 to barge 32. Further, as seen, intermediate chain member 140 extends horizontally inwardly from guide track 122 to barge 32. The suspension assembly below suspension assembly 46 would be similarly attached to the lower wheel in guide track 122 and the barge 32. The other suspension assemblies would be attached to the guide tracks 124, 126, 128, 130, 132 and 134 and the barge 32. The suspension assemblies 44, 50, 56 and 58 are identical. The suspension assemblies 46, 48, 54 and 56 are identical. The only difference between the suspension assemblies 44, 50, 56, 58 and the suspension assemblies 46, 48, 54 and 56 is that the suspension assemblies 46, 48, 54 and 56 are somewhat longer than the suspension assemblies 44, 50, 56 and 58. As stated in the co-pending patent application, the guide tracks or channels could be secured to the barge rather than being secured to the walls of the tank. In that embodiment, the ends of the chains of the suspension assemblies would be secured to the tank. Although it is preferred that each of the suspension assemblies have a horizontally disposed intermediate chain member 140, the intermediate chain member 140 may be omitted in some situations. If an aircraft or a missile should strike either of the cones 100 or 112, the cones will disintegrate and deflect the aircraft or the missile to prevent damage to the nuclear reactor. Additionally, the impact absorbing material in the cones 100 and 112 will lessen the damage to the nuclear reactor. Further, if the barge 32 or the cones 100 and 112 are struck by an aircraft or missile, the suspension systems will permit the barge 32 to move downwardly in the tank 12 to lessen or absorb the impact forces of the strike. The instant invention will now be described. The numeral 142 refers to a water inlet pipe or tube to bring water into the water tank 12 from a large source of water preferably by gravity. Inlet tube 142 includes a valve 144 to control the rate of flow of the water therethrough. An optional pump 146 in inlet tube 142 would help to accelerate the flow of water through inlet tube 142 if necessary. This helps to maintain the temperature and water level of the tank 16 at an optimal range. Valve 144 and pump 146 may be remotely controlled. The numeral 148 refers to a flexible water inlet water pipe or tube for bringing water into the interior compartment 68 of containment member 60 from a large source of water preferably by gravity. A valve 150 is provided in inlet tube 148 to control the rate of flow of water therethrough. An optional pump 152 in inlet tube 148 helps to accelerate the flow of water through inlet tube 148 if necessary. Valve 150 and pump 152 may be remotely controlled. Inlet tube 148 has a flexible, slack and folded tubular section 154 at its inner end. The numeral 156 refers to a flexible water inlet pipe or tube for bringing water into the interior compartment 68 of containment member 60 from a large source of water preferably by gravity. A valve 158 is provided in inlet tube 156 to control the rate of flow of water therethrough. An optional pump 160 in inlet tube 156 helps to accelerate the flow of water through inlet tube 156 if necessary, Valve 158 and pump 160 may be remotely controlled. Inlet tube 156 has a flexible, slack and folded tubular section 162 at its inner end. The inlet tubes 148 and 156 deliver cold water directly into the interior compartment 68 of containment member 60 during emergency cooling to cool the cooling loops and the reactor wall thereof. By adjusting the flow of water using the valves 150 and 158 and optional pumps 152 and 160, the temperature of water in the interior compartment 68 of confinement member 60 is optimized during emergency conditions in the containment compartment 68 during emergency cooling. The numeral 164 refers to a gate which is situated in the upper part of containment member 60 just below water level. Gate 164 is opened during emergency cooling of the reactor. The outflow of water through gate 164 permits additional cold water to enter the interior compartment 68 of containment member 60 via flexible tubes 148 and 156 which helps to maintain maximum cooling conditions in the interior compartment 68 of containment member 60 during emergency cooling. The numeral 166 refers to a conventional condenser. A pipe or tube 168 connects turbine 96 and condenser 166. An inlet pipe or tube 170 extends from a large source of water preferably fed by gravity to the condenser 166. Inlet tube 170 has a flexible, slack and folded tubular section 172 imposed therein. Valves 174 and 176 are imposed in tube 170. An optional pump 178 may be imposed in tube 170. A water inlet pipe or tube 184 extends inwardly from the tank water to the condenser 166. A valve 186 is imposed in tube 184. The purpose of tube 184 is to fill water from the tank 16 into the condenser 166 by gravity. Tube 184 functions as a back-up in case of a break in tube 170. An outlet pipe or tube 188 extends from condenser 166 to a location outwardly of tank 16 by gravity. Tube 188 drains hot water from the condenser 166. Valves 190 and 192 are imposed in tube 188. Optional pump 194 may also be imposed in tube 188. Tube 188 includes a flexible, slack and folded tubular section 196 positioned in tank 12 as seen in FIG. 8. The valves 174 and 176 in tube 170 adjust the flow of inlet water into the condenser 166 with the optional pump 178 accelerating the inlet flow of water if necessary. The valves 190 and 192 in tube 188 adjust the outlet flow of water from the condenser 166 with the pump 194 accelerating the outlet of water if necessary. The valves 174, 176, 190 and 192 maintain optimal cooling temperature of the cooling water in condenser 166. The valves 174 and 190 serve as backups to valves 176 and 192 respectively. As seen, the valves 176 and 192 are located on land while valves 174 and 190 are located inside the barge 16. The numeral 198 refers to an outlet pipe or tube for draining the water from tank 16 preferably by gravity. A valve 200 is imposed in tube 198 for controlling the flow of water from tank 16. The outlet tube 198 with the valve 200 imposed therein helps to maintain the water level and temperature of the water in tank 12 at an optimal range. An optional pump 202 may be provided in tube 198 to accelerate the flow of water from tank 16 if needed. A pipe or tube 204 extends from condenser 166 to interior compartment 92. The flexible tubing sections 154, 162, 172 and 196, with slack, allow the barge 16 to move while maintaining the integrity of the tubes or pipes 148, 156, 170 and 188 respectively. The functioning of the instant invention will now be summarized. The inlet pipe 142 brings water from a large source of water preferably by gravity to the water tank 12. The valve 144 in the pipe 142 controls the rate of flow of water to help maintain the temperature and water level of the tank at optimal range. An optional pump 146 is imposed in the inlet pipe 142 to accelerate the water flow if needed. Water inlet pipes 148 and 156 bring water from a large source of water preferably by gravity to the interior compartment 68 of containment member 60. Both of the tubes 148 and 156 deliver cold water directly to the interior compartment 68 of containment member 60 by utilizing the valves and optional pumps. The system helps to deliver cold water directly into the interior compartment 68 of containment member 60 during emergency cooling to cool the cooling loops and the reactor wall. By adjusting the flow of water using the valves and optional pumps, the temperature of water inside the interior compartment 68 of containment member 60 is optimized during emergency cooling. The valves associated with the pipes 148 and 156 could be controlled from a location farther away from the nuclear reactor. The gate 164 is open during emergency cooling of the reactor. Water would flow out of the interior compartment 68 of containment member 60 via the gate 164. The outflow from the interior compartment 68 of containment member 60 helps to maintain maximum cooling conditions in the interior compartments 68 during emergency cooling. Inlet pipe 170 brings water from a large source of water to feed cooling water to the condenser 166 preferably by gravity. Outlet pipe 188 drains hot water from the condenser. The valves in pipes 170 and 188 would adjust the flow of inlet and outlet water and the optional pumps in pipes 170 and 188 would accelerate the flow if needed. Together, they maintain optimal cooling temperature of the water in the condenser 166 to enable the condenser 166 to function with maximal efficiency. The valve 174 in pipe 170 controls the inlet flow of water to the condenser 166 and the valve 190 in pipe 188 controls the outlet flow of water from the condenser to optimize the temperature of the cooling water in the condenser 166. The valves in pipes 170 and 188 adjust the inflow and outflow of water to and from the condenser 166 to optimize the temperature of the cooling water in the condenser 166. The pipe 198 functions to drain the tank preferably by gravity. The valve 200 in pipe 198 and the optional pump 202 would accelerate the drainage if needed. This helps to maintain the water level and temperature of the water in the tank at optimal range. The flexible, slack and folded tube sections 154, 162, 172 and 196 have slack to allow the barge to move while maintaining the integrity of the tubes. The pipe 184, having valve 186 imposed therein, permits water from the tank to flow into the condenser 166. Thus it can be seen that the invention accomplishes at least all of its stated objectives. Although the invention has been described in language that is specific to certain structures and methodological steps, it is to be understood that the invention defined in the appended claims is not necessarily limited to the specific structures and/or steps described. Rather, the specific aspects and steps are described as forms of implementing the claimed invention. Since many embodiments of the invention can be practiced without departing from the spirit and scope of the invention, the invention resides in the claims hereinafter appended.
	description	This application claims the benefit of U.S. Provisional Application No. 61/794,206 filed Mar. 15, 2013 and titled “PASSIVE TECHNIQUES FOR LONG-TERM REACTOR COOLING”. U.S. Provisional Application No. 61/794,206 filed Mar. 15, 2013 and titled “PASSIVE TECHNIQUES FOR LONG-TERM REACTOR COOLING” is hereby incorporated by reference in its entirety into the specification of this application. This invention was made with Government support under Contract No. DE-NE0000583 awarded by the Department of Energy. The Government has certain rights in this invention. The following relates to the nuclear power generation arts, nuclear reactor safety arts, nuclear reactor emergency core cooling (ECC) arts, and related arts. In a loss of coolant accident (LOCA), the nuclear reactor core is to be kept immersed in water so as to provide for removal of decay heat and to prevent exposure of the fuel rods to air which can lead to chemical reactions and release of airborne radioactivity. The system which provides this water injection is referred to as the emergency core cooling (ECC) system. In a typical arrangement, a refueling water storage tank (RWST) is located with the nuclear reactor inside radiological containment to provide water for use during reactor refueling, and this RWST also serves as a water source for the ECC system. The RWST is located above the reactor core so that the passive ECC system can operate by gravity-driven water flow. Water injected into the depressurized pressure vessel by the ECC system is converted to steam by decay heat from the nuclear reactor core. Preferably, this steam is recaptured by condensing it into the RWST so as to form a closed-loop recirculating heat exchange system. In practice, some steam is lost from the break that caused the LOCA. This lost steam condenses inside the surrounding radiological containment, thereby contributing to heat transfer from the reactor core although not in a recirculating fashion. In some embodiments, the water collects in a containment sump, and a sump pump is provided to recirculate the water back into the RWST. However, this approach is susceptible to failure if the diesel generators or other power source driving the sump pump fail, and moreover there is the potential to transfer contamination into the RWST that can interfere with operation of the ECC system. In one disclosed aspect, an apparatus comprises: a pressurized water reactor (PWR) comprising a pressure vessel containing a nuclear reactor core comprising fissile material; a radiological containment structure inside of which the PWR is disposed; an emergency core cooling system configured to respond to a vessel penetration break at the top of the pressure vessel that results in depressurization of the pressure vessel by draining water from a body of water through an injection line into the pressure vessel; and a barrier configured to operate concurrently with the emergency core cooling system to suppress flow of liquid water from the pressure vessel out the vessel penetration break at the top of the pressure vessel. The barrier may comprise one or more of: (1) an extension of the injection line disposed inside the pressure vessel and passing through the central riser to drain water from the body of water into the central riser of the pressure vessel; (2) openings in a lower portion of a central riser arranged to shunt a portion of the upward flow in the central riser into a lower portion of the downcomer annulus; and (3) a surge line configured to provide fluid communication between a pressurizer volume at the top of the pressure vessel and the remainder of the pressure vessel, the surge line configured to direct water outboard toward a downcomer annulus. In another disclosed aspect, a method comprises operating a pressurized water reactor (PWR) comprising a pressure vessel containing a nuclear reactor core comprising fissile material, and responding to a vessel penetration break at the top of the pressure vessel that results in depressurization of the pressure vessel by operations including: draining water from a body of water through an injection line into the pressure vessel; and during the draining, suppressing flow of liquid water from the pressure vessel out the vessel penetration break. The suppressing may include generating a counterflow in the pressure vessel during the draining in a direction opposite a flow of coolant water in the pressure vessel during the operating, for example by injecting the water from the body of water into the central riser. The suppressing additionally or alternatively may comprise shunting a portion of the upward flow of coolant water in the central riser through holes in the central riser and into a lower portion of the downcomer annulus without the shunted water reaching a top of the central riser. The suppressing additionally or alternatively may comprise directing surge flow between a pressurizer volume and the remainder volume of the pressure vessel outboard toward a downcomer annulus. With reference to FIG. 1, a cutaway perspective view is shown of an illustrative small modular reactor (SMR) 10 and an illustrative refueling water storage tank (RWST) 12 (typically, two or more RWSTs are provided for redundancy). The SMR unit 10 is of the pressurized water reactor (PWR) variety, and includes a pressure vessel 14 and one or more integral steam generators 16 disposed inside the pressure vessel 14 (that is, the illustrative SMR 10 is an integral PWR 10). Alternatively, an external steam generator may be employed. The SMR 10 also includes an integral pressurizer 18 defining an integral pressurizer volume 19 at the top of the pressure vessel 14; alternatively, an external pressurizer may be employed that is connected at the top of the SMR 10 by suitable piping. The pressure vessel 14 contains a nuclear reactor core 20 comprising fissile material such as 235U (typically in an alloy, composite, mixture, or other form) immersed in (primary) coolant water (more generally herein, simply “coolant” or “coolant water”). With the reactor core 20 immersed in coolant water, and when control rod drive mechanisms (CRDMs) 22 at least partially withdraw control rods made of neutron-absorbing material, a nuclear chain reaction is initiated in the nuclear reactor core 20 which heats the (primary) coolant water. The illustrative CRDMs 22 are internal CRDMs, in which the CRDM unit including its motor 22m including both rotor and stator are disposed inside the pressure vessel 14, and guide frame supports 23 guide the portions of the control rods located above the core; in other embodiments, external CRDM units may be employed. In the illustrative integral PWR 10, a separate water flow (secondary coolant) enters and exits the steam generators 16 via feedwater inlet 24 and steam outlets 26, respectively. The secondary coolant flows through secondary coolant channels of the steam generator or generators 16, and is converted to steam by heat from the reactor core carried by the (primary) coolant water. Alternatively, if an external steam generator is employed then large-diameter closed-loop piping feeds (primary) coolant water from the pressure vessel to the external steam generator where heat from the primary coolant converts secondary coolant flow in the external steam generator to steam. The pressure vessel 14 of the illustrative integral PWR 10 includes a lower portion 30 housing the nuclear reactor core 20 and an upper portion 32 housing the steam generators 16, with a mid-flange 34 connecting the upper and lower portions of the pressure vessel; however, the pressure vessel may be otherwise constructed or otherwise configured. A primary coolant flow circuit F inside the pressure vessel 14 is defined by a cylindrical central riser 36 extending upward above the reactor core 20 and a downcomer annulus 38 defined between the central cylindrical riser 36 and the pressure vessel 14. The flow F may be driven by natural circulation (i.e. by primary coolant heated by the reactor core 20 rising through the central cylindrical riser 36, discharging at the top and flowing downward through the downcomer annulus 38), or may be assisted or driven by reactor coolant pumps (RCPs), such as illustrative RCPs including RCP casings 40 containing impellers driven by RCP motors 42. The RCPs may alternatively be located elsewhere along the primary coolant path, or omitted entirely in a natural circulation reactor. It is again noted that the illustrative SMR 10 is merely an illustrative example, and the disclosed ECC techniques are suitably employed with substantially any type of light water nuclear reactor. With continuing reference to FIG. 1, a diagrammatic sectional view is shown of the SMR 10 disposed in a radiological containment structure 50 (also referred to herein as “radiological containment” or simply “containment”) along with the refueling water storage tank (RWST) 12. While a single RWST 12 is illustrated, it is to be understood that two or more RWSTs may be disposed inside containment to provide redundancy and/or to provide a larger total volume of water. The RWST 12 serves multiple purposes. As the name implies, is provides water for use during routine refueling (that is, removal of spent fuel comprising the nuclear reactor core and its replacement with fresh fuel). The RWST 12 also serves as a water reserve for use during certain accident scenarios, such as a loss of heat sinking event in which the heat sinking via the steam generators 16 or other heat sinking pathway is interrupted causing the pressure and temperature in the reactor pressure vessel 14 to rise; or a loss of coolant accident (LOCA) in which a break occurs in a (relatively large-diameter) pipe or vessel penetration connected with the pressure vessel 14. FIG. 1 diagrammatically illustrates the response to a LOCA comprising a break from which steam 52 (possibly in the form of a two-phase steam/water mixture 52) escapes. In FIG. 1 such a LOCA is diagrammatically indicated as originating in the proximity of the integral pressurizer 18 at the top of the pressure vessel 14. In some embodiments the SMR 10 is designed to eliminate the possibility of a LOCA break occurring at an elevation equal to or lower than the top of the reactor core 20. This can be done by designing the pressure vessel 14 with all large-diameter vessel penetrations located above the top of the reactor core 20 (e.g., the steam generator couplings 24, 26 are so located in the embodiment of FIG. 1). As used herein, “large diameter” vessel penetrations are defined as vessel penetrations of diameter 1.8-inch or larger. Additionally or alternatively, passive integral isolation valves may be employed for large-diameter vessel penetrations, so that any pipe breakage at the vessel penetration is immediately and passively sealed by the integral isolation valve. For example, in the case of a make-up line or other water input line, the passive integral isolation valve may be constructed as a check valve built into the mounting flange (rather than in or connected by external piping that is susceptible to breakage) that operates passively to prevent outflow of coolant from the flange having the integral valve. In the case of a letdown line, the passive integral isolation valve can be constructed with a spring bias that maintains the valve in the open position against the pressure of fluid flowing out via the letdown line, with the spring bias chosen such that an increase in (differential) outward pressure above a threshold value overcomes the spring bias to passively close the valve. Again, the integral isolation valve is preferably built into the mounting flange. With such measures, it can be ensured that any LOCA break occurs at an elevation well above the top of the reactor core 20. In the illustrative pressure vessel 14, the only large-diameter vessel penetrations susceptible to a break constituting a LOCA are located at the integral pressurizer 18 at the top of the pressure vessel 14. In such a LOCA, the steam/water 52 that escapes from the integral pressurizer 18 of the pressure vessel 14 is contained by the radiological containment 50, and the released energy is ejected to an ultimate heat sink (UHS) 54 via a suitable transfer mechanism. In illustrative FIG. 1, this heat transfer is achieved (at least in part) by direct thermal contact between the UHS 54 which comprises a large body of water located on top of and in thermal contact with the top of the containment 50. Additionally, a passive emergency core cooling (ECC) is activated, which depressurizes the reactor 10 using valves connected to the pressurizer 18 (in the illustrative example of FIG. 1, or elsewhere in other reactor designs) to vent the pressure vessel 14 to the RWST 12. This operation is diagrammatically indicated by steam path 60 carrying steam (or two-phase steam/water mixture) from the pressurizer 18 to be re-condensed in the RWST 12. Any excess pressure in the RWST 12 resulting from the venting of the pressure vessel to the RWST escapes via a steam vent 62 from the RWST. While depressurizing the reactor, water is initially injected into the reactor vessel from two (for redundancy, or more than two for further redundancy) nitrogen pressurized intermediate pressure injection tanks (IPIT, of which one illustrative IPIT 64 is shown in FIG. 1) to assure the reactor core 20 remains immersed in coolant water during the depressurization. The water in the IPIT 64 optionally includes boron or another neutron poison to facilitate rapid shutdown of the nuclear chain reaction. Once the reactor 10 is depressurized, water in the RWST 12 (or RWSTs, if two or more redundant RWST units are provided inside containment) drains into the reactor vessel 14 via an injection line 66 running from the RWST 12 to the reactor pressure vessel, thus refilling the vessel 14. (Note that in illustrative FIG. 1, a downstream portion of the injection line 66 also provides the input path for water from the IPIT 64, in which case suitable valving is provided to valve off the IPIT 64 after initial depressurization is complete. The valving is optionally passive, e.g. automatically closing when the pressure in the pressure vessel 14 falls below a setpoint. It is also contemplated to connect the IPIT with the reactor pressure vessel via a separate line from the injection line 66.) The water in the RWST(s) 12 provides long-term cooling for the reactor core 20. The RWST 12 is a large body of water conveniently located inside the radiological containment structure 50 and hence is an attractive body of water for use by the ECC system; however, it is alternatively contemplated to connect the injection line 66 to another suitably large body of water that is located at an elevated position respective to the reactor core 20 so as to be drained into the pressure vessel 14 so as to provide emergency core cooling (ECC). During the depressurization, it is expected that substantial primary coolant in the form of steam will exit the pressure vessel 14 via the break that caused the LOCA. After startup of the ECC system, it is expected that steam will continue to exit the pressure vessel 14 via the break, albeit at a lower mass flow rate than during the initial depressurization. In some embodiments the volume capacity of the RWST(s) 12 is designed to be sufficient to remove decay heat for a design time interval, e.g. 72 hours in some embodiments, or 14 days in other embodiments, without the need to recirculate water from a containment sump using sump pumps. This avoids the potential for transferring contaminants from the sump into the RWST. Because the ECC system relies upon gravity feed of water from the RWST 12 into the pressure vessel 14, it is necessary for the water level in the RWST 12 to be higher than the water level in the pressure vessel 14 in order for the ECC to operate. In some embodiments, the initial water level in the RWST 12 is higher than the top of the pressure vessel 14—in such embodiments, it is expected that the water level in the reactor vessel 14 will rise to the top of the pressurizer 18 and liquid water will flow out through the LOCA break. However, once the water level in the RWST 12 drops below the top of the pressurizer 18, it might be expected that the flow out of the break would transition from mostly water to essentially all steam. This transition allows efficient utilization of the RWST water inventory. Since the heat capacity of the water then includes the latent heat for converting the water to steam. However, RELAP (Reactor Excursion and Leak Analysis Program) analysis of long-term cooling indicates that this is not necessarily the case; rather, a two-phase steam/water mixture with substantial water content continues to leave the LOCA break even after the water level in the RWST 12 has drained below the level of the LOCA break. Without being limited to any particular theory of operation, it is believed that this effect is caused as follows. Decay heat from the reactor core 20 generates steam that reduces the density of the water above the reactor core 20. This effect prevents an equilibrium from being established between the water/steam column in the reactor vessel 14 and the water column in the RWST 12. The higher RWST driving head therefore continues to force water out of the break. The magnitude of the problem is illustrated by a simple calculation, performed for a nuclear island design substantially similar to that shown in FIG. 1, in which the RWST (or plurality of RWSTs) has a capacity of about 350,000 gallons, the initial water level in the RWST is at an elevation of 95 feet, and the LOCA break is at a point 10 feet lower in elevation, i.e. at 85 feet. At 120° F., the water in the RWST has a density of 61.7 lb/ft3 (pounds/cubit foot). At 15 psia, saturated water has a density of 59.8 lb/ft3 and steam has a density of 0.038 lb/ft3. If the ECC inlet to the vessel (that is, the inlet of the injection line 66 to the pressure vessel 14 in illustrative FIG. 1) has an elevation of 31 feet and an average quality of 1% (that is, the water flowing in from the RWST is almost purely water with little or no steam content), then the density inside the reactor would be 3.58 lb/ft3. In this case, the water level in the RWST would need to drop to an elevation of about 34 ft (which is 7 feet below the bottom of the RWST in some contemplated embodiments) in order to reach an equilibrium static head. To compensate for this effect it is disclosed herein that the total quality in the central riser 36 (or other upward flow path of the circulating primary coolant) is reduced, or additional pressure drop is incorporated into the ECC injection system. Toward this end, a barrier mechanism, diagrammatically indicated in FIG. 1, is implemented to suppress the flow of liquid water in the central riser 36 (or other upward flow path of the circulating primary coolant) from passing upward to the LOCA break. The barrier may take the various forms, as described in the following. In some embodiments (described herein with reference to FIG. 2), the barrier comprises a modification of the pathways connecting the volume contained by the central riser 36 with the internal pressurizer volume 19. This approach forms the barrier as a direct physical barrier, i.e. a baffle or tortuous path that limits the flow of liquid water from the central riser 36 into the internal pressurizer 18. In some embodiments (described herein with reference to FIG. 3), the barrier comprises modifying the ECC system so that it injects water from the RWST 12 into the central riser 36 in a manner that tends to drive circulation in a direction opposing the primary coolant flow circuit F inside the pressure vessel 14. This forms the barrier indirectly, by slowing or even reversing the velocity of the primary coolant flow circuit F so as to limit the flow of liquid water from the central riser 36 into the internal pressurizer 18. In some embodiments (described herein with reference to FIG. 4), the barrier comprises providing bypass valves that divert a portion of the upward flow leg of the primary coolant flow circuit F from the central riser 36 into the downcomer annulus 38. This again forms the barrier indirectly, by reducing the volume of upward flow in the central riser 36 so as to limit the flow of liquid water from the central riser 36 into the internal pressurizer 18. It will be appreciated that these mechanisms are not mutually exclusive, and the barrier may comprise a combination of two or more of these mechanisms or variants thereof. In general, the amount of steam generated in the reactor vessel 14 after a LOCA is determined by the core decay heat. This cannot be altered by the designer without changing the power level of the plant. However, the quality in the riser 36 can be improved by increasing the flow of water within the riser, by constructing the pressure vessel 14 to be configured to entrain water with the steam. Toward this end, a flow path is provided with the barrier so as to separate the steam and water at the top of the reactor vessel 14 allowing the water to flow to the bottom of the pressure vessel 14 where it can be entrained with steam in the core again. The high quality natural circulation path should interface with the pressurizer 18 in a way that allows the excess water to be separated and directed back to the bottom of the pressure vessel 14. However, this is difficult to achieve in the context of an integral pressurizer, because flow paths are designed to permit relatively free fluid communication between the volume contained in the central riser 36 and the volume 19 of the integral pressurizer 18. With reference to FIG. 2, and in particular the inset of FIG. 2, when the RWST 12 has sufficient driving head to fill the reactor vessel 14, two-phase flow rises and enters the pressurizer 18 through stand pipes 80 that extend through a pump support plate 82. (More generally, the stand pipes 80 pass through a plate separating the pressurizer space 19 from the remainder of the pressure vessel volume. Note that in the inset of FIG. 2 the illustrative RCPs 40, 42 are removed for clarity leaving mounting openings 84 in their place in the pump support plate 82. More generally, the RCPs may be located elsewhere, or may be omitted entirely in a natural circulation reactor.) The surge pipes 80 provide steam venting into the pressurizer space 19 during reactor depressurization. During normal operation, surge lines 86 are provided via which water passes, in a constricted manner e.g. by baffles or the like, to allow pressure in the pressurizer 18 and remainder of the pressure vessel 14 to reach an equilibrium. During normal operation, pressure control elements 88, e.g. resistive heaters, spargers, or the like, are operable to raise or lower the pressure in the pressurizer volume 19, with the surge lines 86 allowing these changes to transfer to the lower operational portion of the pressure vessel 14. During depressurization, however, the surge lines 86 allow water collected in the pressurizer to drain out through the surge lines 86. This flow is directed into the rising two-phase steam/water flow rising up in the central riser 36. This prohibits a natural flow of the water, increasing the average quality within the riser. In the embodiment of the barrier of FIG. 2, the pressurizer surge line 86 is modified to discharge along paths 90 that direct the water through the reactor coolant pumps and then down the tubes of the steam generators 16. (More generally, the modified surge lines 90 direct water outboard toward the downcomer annulus, and are also suitably employed in embodiments that do not employ RCPs or that locate RCPs elsewhere along the primary coolant flow circuit.) These modified paths 90 can be used during normal reactor operation as the surge lines, or can be opened by passive valves in response to an overpressure condition. In another alternative embodiment, the paths 90 are omitted and instead passive overpressure shutoff valves are installed on the surge lines 86 to close these lines off during ECC operation so that only the stand pipes 80 provide steam transport pathways into the pressurizer volume 90. With reference to FIG. 3, in another embodiment of the barrier, the inlet of the injection line 66 to the pressure vessel 14, which conventionally feeds into the downcomer annulus 38, is modified by adding an extension pipe 100 so as to feed into the central riser 36. Optionally, the extension pipe 100 has a downwardly oriented outlet spigot 102 so as to direct the injected coolant from the RWST 12 downward. As diagrammatically indicated in FIG. 3, this tends to produce a coolant circulation flow -F oriented opposite to the direction from the primary coolant flow circuit F inside the pressure vessel 14 that is driven by the decay heat from the reactor core 20. In some embodiments, the magnitude of the counterflow -F is sufficient to actually reverse the direction of circulation in the pressure vessel 14, while in other embodiments the magnitude of the counterflow -F is less than that of the flow F, but is sufficient to slow the velocity of the flow F. The counterflow -F aligns with the discharge of water through the pressurizer, and thereby has the effect of reducing the flow of water into the pressurizer volume 19 driven by the upward current of the flow F in the central riser 36. The counterflow -F can be interrupted when ECC flow is sufficiently low, or if some heat removal is available through a remedial operational mode of the steam generator (in embodiments that include the internal steam generator 16). With reference to FIG. 4, in another embodiment of the barrier, a circulation pattern 108 is created using openings 110 in the core barrel (or other lower portion of the vessel central riser 36) so that a portion of the upward flow in the central riser 36 is shunted into the lower portion of the downcomer annulus 38 without passing upward into proximity with the pressurizer 18. The openings 110 can be holes, holes with flow diodes (i.e. check valves) or passively opened bypass valves to minimize normal bypass flow. This allows natural circulation flow 108 in the lower vessel. The flow 108 can be either in the normal direction (as illustrated) or in the reverse direction. It is to be appreciated that the disclosed mechanisms for implementing the barrier 70 described with reference to FIGS. 2-4 are merely illustrative, and may be combined in various ways. As another illustrative example, if the RCPs are located in a lower portion of the pressure vessel such that they are submerged during the ECC operation, and if electrical drive power is available, then it is contemplated to implement the barrier at least in part by operating the RCPs in retrograde so as to provide the counterflow -F (see FIG. 3) in an active fashion. (Although such operation may be relatively inefficient since the impeller blades are not designed for retrograde operation, the RCPs are nonetheless expected to be capable of generating counterflow -F sufficient to usefully reduce flow of water out the LOCA break.) The disclosed barrier is effective for a pressurized water reactor (PWR) in the case of a LOCA break occurring at the top of the pressure vessel, e.g. in a vessel penetration into an integral pressurizer (as illustrated in FIGS. 1-4) or at piping between the top of the pressure vessel and an external pressurizer or at piping connecting at the top of such an externally pressurized vessel (variants not illustrated). As used herein, phraseology such as “top of the pressure vessel” is intended to encompass any break in a vessel penetration into the integral pressurizer 18 that is large enough to constitute a LOCA (that is, any break in a pipe of diameter greater than 1.8-inch). In the case of an externally pressurized vessel (that is, a pressure vessel that is pressurized using an external pressurizer connected via piping), “top of the pressure vessel” is intended to encompass any break large enough to constitute a LOCA in a vessel penetration at an elevation high enough to be located above the primary coolant circuit in the pressure vessel. Still further, while integral PWR systems in which steam generators 16 are disposed inside the pressure vessel 14 are illustrated, it is contemplated to employ the disclosed embodiments of the barrier in PWR systems that utilize external steam generators. The preferred embodiments have been illustrated and described. Obviously, modifications and alterations will occur to others upon reading and understanding the preceding detailed description. It is intended that the invention be construed as including all such modifications and alterations insofar as they come within the scope of the appended claims or the equivalents thereof.
063079186	summary	BACKGROUND OF THE INVENTION This invention relates generally to computed tomography (CT) imaging and more particularly, to filtration of an x-ray beam in an imaging system. In at least one known CT system configuration, an x-ray source projects a fan-shaped beam which is collimated to lie within an X-Y plane of a Cartesian coordinate system and generally referred to as the "imaging plane". The x-ray beam passes through the object being imaged, such as a patient. The beam, after being attenuated by the object, impinges upon an array of radiation detectors. The intensity of the attenuated beam radiation received at the detector array is dependent upon the attenuation of the x-ray beam by the object. Each detector element of the array produces a separate electrical signal that is a measurement of the beam attenuation at the detector location. The attenuation measurements from all the detectors are acquired separately to produce a transmission profile. In known third generation CT systems, the x-ray source and the detector array are rotated with a gantry within the imaging plane and around the object to be imaged so that the angle at which the x-ray beam intersects the object constantly changes. A group of x-ray attenuation measurements, i.e., projection data, from the detector array at one gantry angle is referred to as a "view". A "scan" of the object comprises a set of views made at different gantry angles during one revolution of the x-ray source and detector. In an axial scan, the projection data is processed to construct an image that corresponds to a two dimensional slice taken through the object. One method for reconstructing an image from a set of projection data is referred to in the art as the filtered back projection technique. This process converts that attenuation measurements from a scan into integers called "CT numbers" or "Hounsfield units", which are used to control the brightness of a corresponding pixel on a cathode ray tube display. To reduce the total scan time, a "helical" scan may be performed. To perform a "helical" scan, the patient is moved while the data for the prescribed number of slices is acquired. Such a system generates a single helix from a one fan beam helical scan. The helix mapped out by the fan beam yields projection data from which images in each prescribed slice may be reconstructed. The x-ray source is typically comprised of an evacuated glass x-ray tube containing an anode and a cathode. X-rays are produced when electrons from the cathode are accelerated against a focal spot on the anode by means of a high voltage across the anode and cathode. The spectrum of the x-rays produced encompasses a band of radiation of differential frequencies having different energies. The short wavelength radiation of higher energy is referred to as "hard" x-ray radiation and the longer wavelength radiation of lower radiation is referred to as "soft" x-ray radiation. The very lowest energy x-rays are almost entirely absorbed by the body and therefore provide little contribution to the x-ray image. Nevertheless, these soft x-rays contribute to the total exposure of the patient to harmful ionizing radiation. In at least one known CT system, a filter is used to remove or reduce the amount of "soft" x-rays. Filters are typically of a "fixed" type or a "shaped" type. The fixed filters are used to improve beam quality by removing soft x-rays which contribute to patient dose but do not contribute to image data measurement. Shaped filters are used to modify the x-ray intensity as a function of fan angle to obtain a more uniform x-ray intensity when a patient is present. The shaped filters are used to reduce x-ray intensity toward a patient extremity where less x-ray beam penetration is required. However, as a result of the different types and areas of the body to be scanned, selection of the ideal filter is difficult, if not impossible. As a result, the selected filter typically compromises either patient dose or beam quality. Accordingly, it would be desirable to provide a filter which allows selection of filtration characteristics depending upon the scan to be completed. More specifically, the filter may be selectably configured to provide proper filtration for suitable x-ray beam quality and intensity for various types of scans. BRIEF SUMMARY OF THE INVENTION These and other objects may be attained in an imaging system which, in one embodiment, utilizes a filter assembly for altering the x-ray beam. Specifically, in one embodiment, the filter assembly includes a movable filter having a plurality of filter portions for altering the quality and intensity of the x-ray beam. Particularly, the filter material and physical shape of each filter portion is configured so that a different quality and intensity x-ray beam is generated from the filtered x-ray beam radiated from an x-ray source. In operation, by positioning the movable filter so that the x-ray beam is filtered by the first portion of the filter assembly, the amount of soft x-rays and the intensity of the x-ray beam is altered to perform a selected type of scan, i.e., a body portion scan. By positioning the filter assembly to the second portion, the shape of the x-ray beam is altered to perform a different type of scan, i.e., a head scan. By using the above described imaging system the x-ray beam quality and shape is alterable depending upon the scan to be completed. More specifically, the filtration characteristics of the imaging system may be selected to provide proper filtration for suitable x-ray beam quality and intensity for various types of scans.
	description	The United States Government has rights in this invention pursuant to Contract No. DE-AC07-05ID14517, between the U.S. Department of Energy (DOE) and Battelle Energy Alliance, representing Idaho National Laboratory (INL). The present disclosure relates generally to nuclear fission fuel compositions, and more particularly to metallic additives for use in nuclear fuel mixtures, and methods for selecting the additive species and amounts thereof. Metals and alloys thereof are used in nuclear fuels to increase burn up (fuel utilization) and to gain beneficial chemical, thermal, and mechanical properties over pure or combined actinide fuels, such as uranium (U) and plutonium (Pu). For example, an additive is often mixed with uranium to form a pseudo-binary mixture “U-xM,” where “M” is a metallic additive, and “x” conveys relative amount information. The metallic additive can be an elemental metal, an alloy of metals, or a mixture of metals. Typically, “x” refers to the weight of the additive as a percentage of the whole U-xM mixture. More generically, U-M refers to a mixture having uranium and a metallic additive without indication of relative amounts or whether the metallic additive is a single metal or an alloy or mixture of metals. Zirconium (Zr) in particular is added to actinide metal fuels to produce mixtures having relatively high solidus temperatures, below which only the solid phase is present. The liquid phase is prohibited in nuclear fuel materials by, for example, regulations of the Nuclear Regulatory Commission (NRC) of the United States. Zirconium additives are also believed to reduce the chemical reactivity of a fuel mixture with materials such as iron (Fe), which is typically present in the steel cladding of a fuel rod. For example, a cylindrical slug of U—Pu-10Zr has been encased in steel cladding to produce a fuel rod. Despite that the chemical components of a nuclear fuel slug are typically uniformly mixed prior to service in an induced fission environment, thermodynamic processes ultimately govern the spatial arrangement of the components, which may migrate even when only the solid phase is macroscopically present. Zirconium, for example, can migrate along temperature gradients to the hottest zone, if hot enough to produce a phase with high solubility for Zr, typically the central core of a cylinder, within an in-use fuel slug such that a central zirconium-rich region is formed and a corresponding radial zirconium-depleted region appears over time. Whatever the geometry, any redistribution of the chemical components of a nuclear fuel slug inhibits accurate modeling and complicates efforts toward designing and implementing a nuclear fuel rod and predicting its performance over time. In various exemplary embodiments, a nuclear fission fuel mixture includes at least one naturally fissioning actinide, and a ternary metallic additive that includes a metal first component as a first percentage of the additive total weight, a metal second component as a second percentage of the additive total weight, and a metal third component as a third percentage of the additive total weight, wherein the first percentage, second percentage, and third percentage sum to about one hundred percent. The first percentage, second percentage, and third percentage are selected in a solid phase region of an isothermal ternary phase diagram of the ternary metallic additive taken at a temperature between about 450 Celsius degrees and about 700 Celsius degrees. In another exemplary embodiment, a nuclear fission fuel mixture is provided for use in a fission reactor in which the nuclear fission fuel mixture stays below an upper temperature limit. The nuclear fission fuel mixture includes at least one naturally fissioning actinide as a first percentage of the total weight of the nuclear fission fuel mixture, molybdenum as a second percentage of the total weight of the nuclear fission fuel mixture, and one or more metals other than molybdenum as a third percentage of the total weight of the nuclear fission fuel mixture. The first percentage, second percentage, and third percentage are selected such that the nuclear fission fuel mixture exhibits a solidus temperature above the upper temperature limit. In yet another exemplary embodiment, a method of making nuclear fission fuel includes providing at least one naturally fissioning actinide, providing molybdenum, providing a metal(s) other than molybdenum, and preparing a total weight of a fuel mixture by mixing the at least one naturally fissioning actinide, the molybdenum, and the metal. The fuel mixture includes the at least one naturally fissioning actinide as a first percentage of the total weight, the molybdenum as a second percentage of the total weight, and the metal as a third percentage of the total weight. The first percentage, second percentage, and third percentage are selected in a body-centered cubic solid phase region of a phase diagram of the fuel mixture. In various exemplary embodiments, nuclear fuel alloys or mixtures and methods of making nuclear fuel mixtures are provided. Pseudo-binary actinide-M fuel mixtures described herein are expected to form alloys and exhibit: body-centered cubic solid phases at advantageously low temperatures; high solidus temperatures; and/or minimal or no reaction or inter-diffusion with steel and other cladding materials. Methods described herein through metallurgical and thermodynamics advancements guide the selection of amounts of fuel mixture components by use of phase diagrams. For example, weight percentages for components of a metallic additive to an actinide fuel are selected in a solid phase region of an isothermal phase diagram, such as shown in FIG. 7, taken at a temperature below an upper temperature limit for the resulting fuel mixture in reactor use. The upper temperature limit can be the melting point of solid uranium (1135° C.), or above 1200° C., or preferably above 1250° C. Advantageous exemplary U-M fuel mixtures and metallic additive M mixtures are described herein. Methods of selecting and making such mixtures, with particular regard to selecting relative component amounts, are also described. Several embodiments of particular mixture types are described herein, some of which at least were arrived upon by way of the above considerations and others. Uranium-molybdenum-tungsten (U—Mo—W) mixtures are detailed in the following descriptions, in which reference is made to various mixtures collectively as the U—Mo—W system. For example, the U—Mo—W system in the range of 90U-10Mo-0W (wt. %) to 80U-10Mo-10W (wt. %) is described, wherein the latter refers to a mixture composed of 80% U by weight, 10% Mo by weight, and 10% W by weight. Mixtures in the uranium-molybdenum-tantalum (U—Mo—Ta) system are also described. For example, the U—Mo—Ta system in the range of 90U-10Mo-0Ta (wt. %) to 80U-10Mo-10Ta (wt. %) is described. The Mo—Ti—Zr system is described for use in U-xM (wt. %) fuel mixtures, in which, for example 5≦x≦20. Furthermore, a region of the U—Mo—Ti system having corners at approximately 98U-2Ti, 90U-10Mo, 87U-11Mo-2Ti, and 90U-9Mo-1Ti is described. U—Pu-M systems should be considered within the scope of these descriptions wherever U-M systems and actinides are expressly recited. Note, however, that U—Pu-xM and U-xM refer to mixtures having the same metallic additive (M) content by weight, but different amounts of U according to Pu content, of which Pu is typically the lesser ingredient (wt. %). It is believed that plutonium has a greater adverse effect than uranium with regard to compromising the advantageously raised solidus temperatures gained by supplementing fissile material with metallic additives. An understanding of various embodiments described herein and the benefits thereof may be improved by consideration of FIG. 1, which is a plot 100 that provides approximate solidus temperatures for U-xZr and U-xMo mixtures from pure U (100 wt. % uranium) to x=10 (90 wt. % uranium) as represented between a U-wt. % axis 102 and a temperature axis 104. As represented by a dashed line 106, the U—Zr system exhibits raised solidus temperatures relative to pure uranium, particularly as Zr is increased (% wt.). As represented by a dashed line 108, the U—Mo system exhibits little or no change in solidus temperatures relative to uranium in the range shown. Furthermore, while U—Mo fuels react more readily with Fe and steel than do U—Zr fuels, U—Mo fuels do not typically undergo migration. U—Mo alloy exhibits low melting temperature particularly when Pu is included in the alloy. The U—Mo system, however, advantageously exhibits bcc-onset temperatures (dashed line 110) below those of uranium and the U—Zr system (dashed line 112), particularly as Mo is increased (wt. %) up to 10 wt %. Note that the solidus and bcc-phase onset temperatures for 100% U are indicated respectively by the convergence of corresponding dashed lines (106, 108) and (110,112). Embodiments set forth herein specify metallic additives (M=?) and analyses for selecting such metallic additives, which are expected to exhibit advantageously high solidus temperatures (solid line 114) and low bcc-phase onset temperatures (solid line 116), both of which are desirable in metallic nuclear fuel performance. For example, novel pseudo-binary and ternary U—Mo fuel systems described herein are expected to exhibit lower onset temperatures for bcc phase than simple U—Zr fuels, and higher solidus temperatures than simple U—Mo fuels. New methods of analysis and selection determine possible new alloy systems and composition ranges expected to show desired improvements. Previous selection methods followed a more Edisonian approach. A metallic nuclear fuel according to at least one exemplary embodiment described herein does not undergo constituent migration, has a high solidus temperature, is minimally reactive toward steel cladding, is safer overall, more predictable, and potentially exhibits higher ultimate burn up. Such fuels may be advantageous, for example, in fast nuclear reactors. The stability of a nuclear fuel mixture can be enhanced by increasing the melting temperature and lowering the bcc-phase onset temperature. Melting temperatures are an indication of bond strength, and melting typically proceeds from the body-centered cubic (bcc) phase. Thus, the stability of a U-M or U—Pu-M fuel mixture can be inferred from the solidus temperature of the fuel mixture itself, or even that of M, which may be a single metal or a metallic mixture that is binary, ternary, or higher. The lowering of γ-U (gamma-U, the bcc-phase) onset temperatures for U-M fuel mixtures is believed to be advantageous with particular regard to minimizing the redistribution of mixture components by migration at reactor temperatures and interdiffusion between fuel alloy and steels. If a cubic phase of a fuel alloy is achieved, thermal expansion is expected to be isotropic and migration is expected to be minimized or prevented. According to various embodiments described herein, a preferred species of U-M, U—Pu-M or M (FIG. 1) exhibits a widened temperature range between its bcc-phase onset temperature (line 116) and its solidus temperature (line 114). That is, a preferred species has a wide bcc-phase temperature range in which component migration is inhibited and melting is avoided. Fuel properties, such as thermal conductivity, will be more uniform throughout the fuel body when the fuel body is single phase and cubic. The solidus temperatures for species of the U—Pu—Mo and U—Pu—Zr systems can be compared to assess the improved stability of these systems relative to U—Pu only. For example, the solidus of U-19Pu-10Mo (wt. %), corresponding to U62.1Pu16.2Mo21.7 (at. %), is 1000 degrees Celsius. The solidus of U-19Pu-10Zr (U61.4Pu16Zr22.6 in at. %) is 1150 degrees Celsius. For this ratio of U to Pu (U-21Pu, in wt % or at. % approximately), the solidus is approximately 950 degrees Celsius. That the transition at melting is from the γ-U (bcc) phase in all three cases indicates that the presence of molybdenum stabilizes the γ-U phase a little against melting (approx. 50 degrees Celsius), while the presence of zirconium stabilizes it quite a bit more (approx. 200 degree Celsius). The stabilization of the γ-U phase toward melting is chemical (metallurgical) in nature, so that the propensity of reaction or interdiffusion with iron (steel) can also be ranked by the solidus temperature, provided the dissolved Mo and Zr in the alloys are not reacting differentially and more strongly toward the steel than U and Pu. U—Mo based fuel alloys according to embodiments herein offer a solidus comparable to or better than U—Zr. The net result is not only a fuel alloy that is safer with regard to melting temperature in the off-normal event, but is also less reactive toward steel. Compositions and alloy systems described herein were determined in view of these considerations, which considerations correspond to novel analyses and methods. FIG. 2 is a phase diagram of the U—Mo system across a range of U content (at. %). As shown, liquid phase (L) develops in U—Mo above 1135 degrees Celsius (solidus), with modest increases shown with decreasing U. A high-temperature two-phase region, (U)ht2, is the γ-U or bcc-phase region having an onset temperature of 550 degrees Celsius near 80% uranium. At lower temperatures an alpha (a) phase is present. A high solidus temperature and a low γ-U onset are desired of candidate alloys and mixtures. In Table 1, several U, Pu, Mo and Zr systems are considered. Of which, U-10Zr (wt. %) exhibits the highest solidus temperature. TABLE 1Solidus Alloy(degrees Celsius)U—10Zr1350U—10Mo1150U—19Pu—10Zr1150U—19Pu—10Mo1000U—21Pu 950 Stabilization of fuel is approached, according to embodiments herein, by developing alloys with high solidus temperatures. Melting temperature is a direct indication of bond strength in extended solids, for example tungsten (W) and carbon (C). Such direct indication is expected among different fuel alloys if all of the alloys to be compared enter liquid phase from γ-U. FIG. 3 is a phase diagram of the U—Zr system, in which multiple phases are exhibited at typical reactor temperatures (approximately 430-730 degrees Celsius). Between 435 and 617 degrees, α-U and δ-UZr2 (delta-UZr2) are present. Between 617 and 656, α-U and γ-2 are present. Between 656 and 692, β-U (beta-U) and γ-2 are present. Between 692 and 737 degrees Celsius, γ-1 and γ-2 are present. Owing to the radial temperature gradient within U—Zr and U—Pu—Zr alloy fuel slugs while under irradiation, and the ensuing thermochemical gradient, constituent migration occurs within the fuel. As a consequence, the fuel properties (such as melting point, density, thermal conductivity, etc.) can vary markedly between the center and periphery of the fuel slug. Zirconium is used to raise the solidus of the fuel composition and guard against melting, and additional benefit is realized because zirconium migrates toward the center (hottest) part of the fuel when the hottest part is in the γ phase region. It is now recognized that the presence of the bcc-phase (γ-U) excludes anisotropic fuel expansion, and for this reason it is desirable to lower the onset temperature of the bcc-phase through alloying and stabilize it as much as possible to lower temperatures. The bcc-phase is also desirable in the case of an off-normal, rapid, high-temperature excursion if a Zr-depleting zone exists along the radius of the fuel slug. If, for example, the bcc phase is stable over a larger range of fuel temperatures, then Zr migration would be less and perhaps, due to the lack of mobility of Zr at low temperatures, may not occur. The solidus temperature would be only minimally affected, and only in the cooler operating areas of the fuel and the high temperature off-normal event becomes less a concern. In the case of eutectoid U-10Mo (wt. %), the γ-U forms at 550 degrees Celsius. In U-10Zr, the γ phase is fully formed (γ-1 plus γ-2) at 680 degrees Celsius (the eutectoid composition and temperature are approximately U-65Zr and 610 degrees Celsius). However, the U-10Mo system also offers a lower solidus temperature compared to U—Zr, which is less desirable for fuel performance, and the same is true for the comparable ternary U—Pu-M systems (M=Mo or Zr). Moreover, the U—Pu—Mo system appears to react and interdiffuse with iron and steels more aggressively, and perhaps at lower temperatures, as compared to the corresponding U—Pu—Zr alloys (at least for some compositions if not uniformly for all). As a consequence, it is desirable to develop a fuel alloy formulation that 1) stabilizes the bcc γ-U phase at as low a temperature as possible and that 2) offers a solidus temperature that is high, but not too high for casting the alloy and fabricating the fuel and 3) introduces properties to inhibit fuel/cladding interdiffusion. U-10Mo is a basis for a new ternary alloy U—Mo-M, in which an alloy with M is to be prepared that increases the solidus temperature. Analysis of existing phase diagrams shows that W and Ta alloy additions are interesting for three reasons: 1) W and Ta are bcc metals at room temperature, which should favor a lower onset temperature of γ-U in the ternary system; 2) the solidus properties for binary U—W and U—Ta systems (FIG. 4) are improved as compared to binary U—Mo (FIG. 5); and 3) the ternary systems (U—Mo—Ta and U—Mo—W) had been unexplored. Mo—Ta and Mo—W binary systems are expected to exhibit favorable solidus/liquidus behavior. The U—Mo solidus is depressed in the U-rich region of the phase diagram (FIG. 4); the solidus temperature rises to 1284 degrees Celsius only after the Mo fraction reaches 37 wt. % (59 at. %). At approximately 16 wt. % Mo (32 at. %), the liquidus rises almost ideally from 1284 degrees Celsius to the melting point of Mo. This indicates a relative decrease in the cohesive energy of the alloy as compared to ideal behavior over the whole composition range. While the solidus temperature for U—Ta (FIG. 5) is also slightly depressed for high uranium fractions, the solidus increases up to 1160 degrees Celsius with only 2.3 wt % Ta and a sharp increase in liquidus ensues. This may indicate a stronger cohesive energy for the U—Ta system as compared to the U—Mo system. The U—W system gives a solidus at 1135 degrees Celsius, and the liquidus rises very rapidly beginning at 0.8 wt. % W (1 at. %) (FIG. 6). Attention is given when casting U—Mo—Ta and U—Mo—Ta systems to the rapidly rising liquidus temperatures for the binaries. Higher systems with U and Pu derive significant benefits, as Pu tends to lower the solidus and liquidus. In at least one embodiment of a ternary alloy or composition for use in nuclear fuel, the U—Mo—W system is utilized in the range of 90U-10Mo-0W (wt. %) to 80U-10Mo-10W (wt. %). In at least one embodiment, 90U-9Mo-1W (wt. %) is utilized. In at least one embodiment of a ternary alloy or composition for use in nuclear fuel, the U—Mo—Ta system is utilized in the range of 90U-10Mo-0Ta (wt. %) to 80U-10Mo-10Ta (wt. %). In at least one embodiment, 87.3U-9.7Mo-3Ta (wt. %) is utilized. A second candidate system involves preparing a bcc alloy of Mo—Ti—Zr that exhibits a low onset temperature for the bcc phase. This occurs below 600 degrees Celsius for the approximate composition A=50Mo-43Ti-7Zr (wt. %), or approximately equivalently Mo35Ti60Zr5 (at. %). The bcc alloy (A) may then be combined with U up to approximately U-10A (wt. %) in order to achieve a uranium alloy with low bcc onset temperature and with an expected high solidus temperature. Though appearing complex, this system is a pseudo-binary system. The rationale is straightforward: 1) use Zr and Ti which are expected to raise the solidus in combination with Mo at the composition offering the lowest onset for bcc phase of the ternary alloy; and 2) combine with U to lower the bcc temperature even further for the quarternary system while keeping the Mo—Ti—Zr ratio fixed (otherwise the system becomes rapidly complex). The desired outcome is likely the single, simple bcc phase at reactor temperatures, and a pseudo binary alloy of U-xA is therefore an advantageous choice by analysis and design. When alloyed with uranium, the U—Mo—Ti—Zr total system is pseudo binary (U-M), and is expected to give a low bcc-phase onset temperature. A large region of the ternary Mo—Ti—Zr system exhibits a single-phase bcc onset temperature below 600 degrees Celsius, in fact more than 25% of the ternary diagram. A bcc ternary alloy M (for example, M=50Mo-43Ti-7Zr in wt. %) may therefore be combined with uranium up to approximately U-10M (wt. %) in order to achieve a uranium alloy with the desired low bcc onset temperature and higher solidus temperature. FIG. 7 is an isothermal ternary phase diagram of the Mo—Ti—Zr system taken at about 600 degrees Celsius. The diagram has three axes representing respective atomic percentages (at. %) for molybdenum (Mo), titanium (Ti) and zirconium (Zr). Any ternary mixture of Mo, Ti and Zr can be represented as a point within a triangular zone between the three axes. Pure Mo is represented as a corner of the triangular zone at 100 atomic percent. Similarly, pure Zr and pure Ti is represented respective 100 at. % corners. Any ternary mixture of Mo, Ti and Zr can be indicated by specifying respective atomic percentages for Mo, Ti and Zr, such that the sum of the percentages is one hundred percent. For example, a mixture having Ti and Zr in equal atom counts, and twice as many atoms of Mo as either Ti or Zr, can be indicated as Mo50Ti25Zr25 (at. %). Weight percentages (wt. %) corresponding to atomic percentages (at. %) can be calculated by considering atomic masses. A first exemplary alloy or mixture, quantified in Table 2 and indicated in FIG. 7 as composition A, can be quantified as either Mo35Ti60Zr5 (at. %) or 50Mo-43Ti-7Zr (wt. %) without ambiguity. A second exemplary alloy or mixture, quantified in Table 2 and indicated in FIG. 7 as composition B, can be quantified as either Mo15Ti70Zr15 (at. %) or 23.4Mo-54.4Ti-22.2Zr (wt. %). These exemplary compositions may be used in U-M alloys, for example, in the 5-10 wt. % M range for lowered γ-U phase onset. Other compositions within the single-phase bcc region of FIG. 7 may be used also. For example, higher Zr compositions, two of which are particularly plotted as additional points within the single-phase bcc region of FIG. 7, may be used in U-M alloys for lowered γ-U phase onset. COMP. A and COMP. B of Table 2 are particular embodiments of additives (M) in U-xM, U—Pu-xM and Actinide-xM systems, where x ranges from small values to approximately 10 (wt. %). For example, these systems at 3%, 6% and 10% M content by weight are particularly considered embodiments. Further compositions are warranted in systems where eutectoid temperature is indicated in 0≦x≦10. Even further compositions are warranted in systems up to x=20 (wt. %) where a eutectoid minimum is indicated but not found at 10 wt. % W, and solidus/liquidus temperatures are not too high (1200/1350 degrees Celsius, for example). TABLE 2SPECIESCOMP. ACOMP. ACOMP. BCOMP. BMo35 at. %50 wt %15 at. %23.4 wt.%Ti60 at. %43 wt %70 at. %54.4 wt. %Zr 5 at. % 7 wt %15 at. %22.2 wt. % Furthermore, exemplary Actinide-(Mo—Ti—Zr), Actinide-(Mo—W), Actinide-(Mo—Ta), and Actinide-(Mo—Ti) systems according to embodiments herein include, but are not limited to, compositions: preferably at least approximately 80% Actinide content by weight; more preferably at least approximately 85% Actinide content by weight; even more preferably at least approximately 90% Actinide content by weight; and no greater than approximately 97% Actinide content by weight. Moreover, niobium (Nb) is included in fuel-metal and metal compositions as an alloying agent in some embodiments. For example, Nb can be added in small quantities, such as approximately 0.5% and 2.0% by weight of the metal or fuel-metal composition. Lanthanide fission products in metallic fuels interact with steel claddings, and the resulting wastage factors into the limitation on burn up. Even though high burn ups are achieved with metallic fuels, the chemical stabilization of the lanthanides against fuel/cladding chemical interaction (FCCI) is expected to enhance fuel performance to higher burnup. Lanthanides can migrate extensively to the fuel slug periphery whereupon fuel expansion permits formation of low melting eutectic systems, such as Ce—Ni and Ce—Fe. To address lanthanide-based FCCI, niobium can be added to cladding materials, cladding coatings, liners between fuel and cladding, and fuel additives to stabilize the lanthanides as intermetallic compounds. As intermetallic compounds, the lanthanides may be immobilized in the fuel matrix, or, if not immobilized, may show lesser reactivity with cladding. Attention is to be directed toward the bcc-phase onset temperature and solidus temperature with inclusion of additives such as platinum, niobium, or chromium, to reduce deformations upon thermal cycling at the α-β transformation temperature. It is worth noting that palladium is chemically similar to platinum, with comparable atomic radius, and may work as platinum does on mitigating deformation, and it does mitigate lanthanide-based FCCI. Platinum, however, has much higher solubility in α and β uranium than palladium does, which may be the reason for the influence of platinum on curbing deformation during repeated α-β phase transformations. A brief summary of electrorefining is given here to help provide further discussion of the impacts on fuel performance and processing of the new alloy developments. Chopped segments of spent fuel are loaded into anode baskets that are then immersed into molten LiCl—KCl (eutectic ratio) at 500 degrees Celsius. A steel mandrel is used as the cathode during electrorefining to collect the purified uranium metal dendrites. At the anode, uranium in the spent fuel is oxidized into the molten salt as U+3 and, while at the cathode, is reduced to uranium metal. Active metal fission products (such as lanthanides) accumulate in the molten LiCl—KCl as their respective chlorides (e.g., LaCl). Metal fission products more noble (less electropositive) than uranium (such as Tc, Ru) are either retained in the cladding hull segments or released into the electrorefiner where they adhere to vessel components or dissolve in liquid metals. Alloy components such as zirconium can be partially oxidized to deposit with the purified uranium, or can suffer a similar fate as the more noble metal fission products (such as Tc, Ru). Regarding expected electrorefining behavior for new alloy components, Table 3 lists an abbreviated electromotive force (emf) series for fuel components in LiCl—KO eutectic at 450 degrees Celsius. For this series, neodymium is the most electropositive and ruthenium is the least electropositive. Proceeding down the table, the relative stability of the chlorides is ranked, with NdCl3 being the most stable chloride (with NdCl3 stability comparable to the chlorides of the other lanthanide fission products). On account of the stability of the lanthanide chlorides, these fission products accumulate in the molten LiCl—KCl salt, which contributes to their disposition in the ceramic waste form. The relative abundance of chemical species present in irradiated fuel has been observed, for example, at the Experimental Breeder Reactor—II (EBR-II) of Argonne National Laboratory. The electrorefining behavior of zirconium in the fuel alloy exhibits oxidation, deposition with uranium at the cathode, and spurious deposition of zirconium on the exterior of the cladding. Depending on the process conditions, it is possible to electrochemically oxidize zirconium from the steel cladding and the extent of oxidation can be controlled. Since the stability of the chlorides of titanium is comparable to the chlorides of zirconium in this system, the electrorefining characteristic of titanium is expected to be comparable to zirconium, and likewise controllable. The emf for ruthenium is comparable to those for molybdenum and palladium. Ruthenium can be retained with the cladding, and it can contaminate electrodeposited uranium as a consequence of convection of fine particulates. However, palladium should be chemically bound with the lanthanides, and the lanthanides should be more difficult to oxidize to their respective chlorides, by the lowering of their free energy upon formation of the intermetallic compound. This could benefit the ceramic-salt waste streams, as the lanthanides would not accumulate in the eutectic LiCl—KCl as rapidly as compared to absence of palladium. Molybdenum will not electrochemically oxidize in the electrorefining process, and its presence is expected to aid the retention of zirconium and titanium. Alloy composition can include unirradiated depleted uranium alloys. The presence of molybdenum could slow down and impede the electrochemical oxidation of the last few percentage points of uranium, as zirconium does. TABLE 3OXIDATION/REDUCTIONCOUPLEE0M (Pt)(V)E0m (Ag)(V)Nd (Ill)/Nd (0)−2.819−2.097U (III)/U (0)−2.218−1.496U (IV)/U (0)−1.950−1.230Zr (IV)/Zr (II)−1.864−1.153Zr (IV)/Zr (0)−1.807−1.088Zr (II)/Zr (0)−1.75−1.02Ti (II)/Ti (0)−1.74−1.01Ti (Ill) I Ti (0)−1.60−0.88Ti (IV)/Ti (0)−1.486−0.767Cr (D)/Cr (0)−1.425−0.698Ti (III)/Ti (II)−1.32−0.61In (I)/In (0)−1.210−0.467Fe (II)/Fe (0)−1.172−0.445Mo (Ill)/Mo (0)−0.603+0.119Pd (II)/Pd (0)−0.214+0.513Ru (lll)/Ru (0)−0.107+0.615 Impacts on fuel fabrication are also minimal. Casting should be little affected, based on the experience with the binary systems. It may be necessary to perform some pre-alloying step to produce a more homogeneous alloy composition. Thermal conductivity can be expected to be comparable or better than U-10Zr, because of the presence of molybdenum. Likewise, fuel swelling and constituent redistribution are expected to be improved compared to U-10Zr, because of the sizable increase in the fraction of cubic phase. For TRU burning fuels, the safety margin toward fuel melting will be increased compared to a U—Mo alloy with TRUs. The fast reactor neutronics penalty for molybdenum is substantial in comparison to zirconium (but less than a factor of ten); however, the change in reactivity can be accommodated in a fast reactor with core size and fissile content, for example U-xMo, with x=7-10% by weight. The cross section for titanium in fast reactors is comparable to zirconium, and the titanium cross section is less than zirconium for most of the spectrum between 0.2 and 3 MeV. The impact on the waste streams should be minimal. The waste stream already needs to accommodate Tc-99, because U-10Zr at 8 atom percent burn up contains 0.17 wt. % (0.34 atom %) technetium. The Tc-99 generated from Mo-98 in the fuel alloy therefore does not introduce a new waste stream. Otherwise, the titanium and molybdenum should behave similar to zirconium; electro refining with a U-M alloy is discussed in the following. Reference is now made toward superior uranium alloys with regard to alloy solidus transition temperature and the transition temperature of the body centered cubic γ-phase for improved performance and safety margin. The alloy of interest includes uranium and various concentrations of M, composed as U-xM, where M is 50Mo-43Ti-7Zr (wt. %). Particular examples include x=5, 7.5, 10, 12.5, and 15. Increasing the concentration of M decreases the temperature of the eutectoid and increases the temperature of the solidus onset up to approximately 10 wt. % M. Above x=10, the eutectoid onset, solvus onset, and solidus reach a plateau in temperature. According to collected differential scanning calorimetry (DSC) data, increasing the alloy addition of M in uranium increases the solidus transition temperature and decreases the γ-phase transition temperature with increasing alloy addition up to approximately 10 wt. % M. Environmental scanning electron microscopy (ESEM) images coupled with energy dispersive X-ray (EDX) analysis, indicate that a feature consisting of a proeutectoidal precipitation and divorced eutectoid, primarily of titanium and zirconium, is forming during the cooling through γ-α phase transition. The molybdenum addition is uniformly distributed in the uranium and little if any molybdenum is present in the feature consisting of proeutectoidal precipitation and divorced eutectoid. See table 4 for DSC results. TABLE 4Alloy CompositionSolidus ° C.Solvus ° C.Eutectoid ° C.U-15 M (wt. %)1228.3 ± 9.1682.2 ± 8.9633.0 ± 2.0U-12.5 M (wt. %) 1226.7 ± 7.9694.7 ± 10.1626.0 ± 9.0U-10 M (wt. %)1220.9 ± 16.2681.6 ± 22.0627.7 ± 7.4U-7.5 M (wt. %)1186.7 ± 6.8677.6 ± 2.9631.9 ± 5.3U-5 M (wt. %)1173.6 ± 3.5687.6 ± 6.3641.6 ± 2.9Uranium 1135 ± 0 781 ± 0670.6 ± 0 Coring, segregation, and cooling were found to occur. X-ray diffraction (XRD) analysis was used for qualitative phase identification. The ESEM results illustrate a proeutectoidal precipitation and divorced eutectoid bounding prior γ-phase grains. EDX results indicate proeutectoidal precipitation and divorced eutectoid in a Zr,Ti-rich phase and proeutectoidal precipitation and divorced eutectoid composed of a uranium and molybdenum rich α-phase. In addition, the EDX results suggest that coring exists because of oscillations in the number of counts for the alloy constituents. Optical microscopy was used as an additional verification of phases using both chemical and electrolytic etch techniques. The presence or absence of phases within each micrograph suggest a cooling and solidification history unique to the composition of the casting. The DSC results were used to determine transition temperatures for each alloy. The DSC results verified the expansion of the γ-phase field and gave specific temperatures to suggest reactions using the results from XRD analysis and optical microscopy coupled with transition temperatures. The as-cast alloys may produce a range of composition dependent transition temperatures during phase transformation specifically at the transition temperature of the γ-phase and the transition temperature of the solvus. A range in transition temperatures is illustrated by the error estimations on the DSC analysis results in Table 4. The results of the DSC analysis suggest that with increasing M, the transition temperatures from γ-to-α+U2Ti(Zr) phases decrease from unalloyed uranium transition temperature for the α-to-β at 670 degrees Celsius and plateau at about 640 degrees Celsius. The transition temperature of γ-solidus increases from about 1135 degrees Celsius and plateaus at about 1225 degrees Celsius. The solvus temperature also decreased from 781 to about 680 degrees Celsius. Generally, the γ-phase field broadens with increasing alloy addition of M. A transition temperature found at 670 degrees Celsius may indicate the temperature region where γ-to-U2Ti(Zr)+α occurs for alloys containing 10 wt. % M and above. Lower concentrations of M may transition from the γ-phase by γ-to-U2Ti(Zr)+β. A significance of a γ transition temperature found at 640° C. is that the temperature is similar to that of the two-phase region located in the uranium-molybdenum binary in FIG. 5. The two-phase region consists of the γ+α phases and is not a eutectoid transition. The transition temperature corresponds to a eutectoid transition temperature where the reaction is γ-to-α-U2Mo. The γ-phase transition temperature at 640 degrees Celsius indicates that the eutectoid temperature relative to the uranium-molybdenum system has increased with the addition of titanium and zirconium to the system. The solvus temperature relative to the molybdenum-titanium system has also increased roughly 40 degrees Celsius. The greatest expansion of the γ-field is demonstrated by uranium-10M alloy where M is 50Mo-43Ti-7Zr (wt. %) The results indicate that γ-phase field broadens with the addition of M to uranium as a solute. In addition, the solvus temperature decreased after the addition of 5 wt. % M but transition temperature plateaued with increasing addition of M. The solvus temperature currently has the most variance of the transition temperatures because of peak overlap from the end of the eutectoid heating curve peak and the beginning of the solvus heating curve peak (FIG. 8). The temperatures have the largest variation in transitions temperatures because there are two possible reactions occurring that may obscure the solvus onset temperature ranging between 730 to 668 degrees Celsius. The first possible reaction is the transition from γ-to-β+U2Ti(Zr) first observed at 700 degrees Celsius. The second reaction is γ-to-α+U2Ti(Zr) and is first observed at 662 degrees Celsius. In addition, added degrees of freedom associated with the phase rule for systems with greater than two components indicate a broadening of equilibrium lines. The broadening of equilibrium lines over a range of alloy additions is verified by DSC analysis. U—Mo based fuel alloys are of interest because they have a low onset temperature for the body-centered cubic γ-phase. The bcc phase expands isotropically with increasing temperature. Anisotropic fuel swelling was an early metallic fuel performance issue and was mitigated through alloy additions that stabilized higher symmetry fuel phases. The DSC data suggest that increasing M broadens the transition temperature bounding the γ-phase field. X-ray diffraction data indicates that addition of M greater than 5 wt. % suppresses the transition to the β-phase field. The γ-phase field is broadened by lowering the α-to-γ transformation temperature and increasing the γ-to-solidus transformation temperature. The ternary diagram shown in FIG. 6 indicates that alloying with M would lower the γ-phase transition temperature relative to pure uranium because of the presence of the bcc phase at 600 degrees Celsius. When the data is compared to the uranium-molybdenum binary diagram, the eutectoid temperature relative to FIG. 5 at 550 degrees Celsius has risen to roughly 640 degrees Celsius. The maximum spread of the γ-phase field is 600 degrees Celsius at 12.5 wt. % M. Metastability was not observed based on DSC results but significant γ-phase uranium was identified using X-ray diffraction analysis. A third candidate system is the U—Mo—Ti system. A particular region of interest within the ternary diagram of the U—Mo—Ti system is represented in FIG. 9. Compositions specified as of particular interest according to embodiments described herein fall approximately within a region defined by four compositions C, D, E and F, which are selected as four corners of a region of interest as shown in FIG. 9. Composition C is approximately 98U-2Ti (wt. %). Composition D is approximately 90U-10Mo (wt. %). Composition E is approximately 87U-11Mo-2Ti (wt. %). Composition F is approximately 90U-9Mo-1Ti (wt. %). These alloys were chosen as the corners of a region of interest within the ternary diagram of the U—Mo—Ti system because of their eutectoid compositions that offer the low bcc-phase onset temperatures found in the binary systems. A lower bcc onset temperature may exist within the specified region of interest, with the corners of the phase field posing the “high” temperature boundaries. These U—Mo—Ti species are given in Table 5. TABLE 5SPECIESCOMP. CCOMP. DCOMP. ECOMP. FU98 wt. %90 wt. %87 wt. %90 wt. %(92.5 at. %)(78 at. %)(70.5 at. %)(78 at. %)Mo0 %10 wt. %11 wt. %9 wt. %(22 at. %)(22 at. %)(19.4 at. %)Ti2 wt. %0 %2 wt. %1 wt. %(7.5 at. %)(7.5 at. %)(2.6 at. %) In at least one embodiment of a composition for use in nuclear fuel, the U—Mo—Ti system is utilized approximately as 88.3U-3.5Mo-1.8Ti (wt. %). In at least one other embodiment of a composition for use in nuclear fuel, the U—Mo—Ti system is utilized approximately as 92U-6.9Mo-1.1Ti (wt. %). FIG. 10 summarizes the results for two pseudo-binary systems, U-xMA and U-xMB, where x varies up to 15 weight percent. MA designates composition A (Table 2), the constant weight ratio 50Mo-43Ti-7Zr. MB designates the constant weight ratio 73Mo-27Ti. It may be noted that both systems offer increased melting temperature over the U—Mo system. However, the U-xMA does not appear to offer the desired benefit for lowering the bcc onset temperature substantially in comparison to U—Zr. The U-xMB system appears promising because the minimum onset temperature has not been observed yet, and the curve has not leveled off. Higher concentrations of MB (higher x in U-xMB) likely exhibit lower onset temperatures, the minimum of which may correspond to an advantageous U-xMB composition. Various embodiments of alloys and mixtures described herein can be made by, for example, vacuum induction melting, vacuum arc remelting, and cold crucible melting without limiting processing to these examples. The target alloy composition can be prepared by melting the individual elements together. The target alloy composition can be prepared by making pre-alloys. For example, first a “pre-alloy” of U-10Mo (wt. %) can be prepared, and then, in a separate heating, additional uranium, molybdenum, titanium, zirconium and/or other species described herein can be added to reach the desired composition. More than one pre-alloy may be used to prepare the final alloy. For example, a pre-alloy of U—Mo might be combined with a pre-alloy of Mo—Ti—Zr, with compositions chosen such that the desired final alloy composition is reached during the final melting operation. In vacuum induction melting, material is melted by eddy currents in an evacuated induction furnace. In vacuum arc remelting (VAR), the material to be melted is used as one of two electrodes across which a currents arcs, causing melting of the material within a cooled crucible held under vacuum. In cold-crucible melting, an induction coil surrounds a crucible containing the material to be melted. The crucible has, in its construction, electrically-isolated water-cooled tubing. When an electric current passes through the induction coil, the material contained by the crucible is heated inductively. The water-cooled tubing freezes or maintains a solid outer shell of the material within the crucible, with a liquified melt contained within the cooled shell. Thus, high temperatures can be reached while maintaining the melt in a solid containment shell that isolates the crucible from the melt. In cold-crucible melting, direct contact of current carrying electrodes with the melt is avoided. In at least one embodiment of an alloy with uranium as the only actinide, the uranium composition ranges from 85 to 95 weight percent, with the balance taken by molybdenum, titanium, and zirconium. In another embodiment, molybdenum and titanium are alloyed with uranium (U—Mo—Ti), of which several examples are specified by wt. % herein. Other exemplary embodiments include U—Mo—W and U—Mo—Ta. In other embodiment, a small amount of niobium may be used as a minor alloying agent (e.g. 2 wt. % Nb). In still other embodiments, an alloy contains uranium and/or other actinides, such as plutonium and uranium. In at least one embodiment, an alloy contains uranium, plutonium and the minor actinides neptunium and americium. In some embodiments in which the only actinide present is uranium, the desired solidus temperature is in excess of 1200 degrees Celsius. The desired liquidus temperature (the temperature of complete melting, or temperature above which only liquid alloy is present) is below 1400, or preferably below 1350 degrees Celsius. For at least one solid alloy, at temperatures below the solidus temperature, it is desired that the alloy is single-phase, body-centered cubic (bcc) structure, also known as the γ phase. It is further desired that the single-phase behavior is retained from the solidus temperature down to 600 degrees Celsius, preferably to 550 degrees, and more preferably to 500 degrees Celsius. In other embodiments, other actinides are present in alloys in addition to uranium, such as plutonium or neptunium. In some such embodiments, both the solidus and liquidus temperatures of the alloy are less than temperatures of uranium-based single-actinide alloys. For example, solidus is preferably above 1150 degrees Celsius for the multi-actinide alloys. As with single-actinide alloys, a single-phase bcc structure is desired in multi-actinide alloys. TABLE 6SPECIESIIIIII IVVVIVIIVIIIIXXXIU90.085.088.392.089.067.069.09087.39067.5Mo 5.0 3.5 9.82 6.9 3.55.09.49.0 9.7710Ti 4.3 8.21.8 1.1 8.2 4.30.6Zr 0.7 3.3 3.30.7Nb 1.02.00.5W1.02.52.5Ta3.0Pu19.015.015Np2.5 3.03Am1.51.02 In Table 6, in which wt. % values are given, species I corresponds to composition A of Table 2, combined with uranium in a 90U-10M (wt. %) fuel composition in which M=50Mo-43Ti-7Zr (wt. %). Species II corresponds to composition B of Table 2, combined with uranium in a 90U-10M (wt. %) fuel composition in which approximately M=23.4Mo-54.4Ti-22.2Zr (wt. %). Species III and IV are U—Mo—Ti fuel mixtures as specified. Species V and VI are fuel mixtures of the Mo—Ti—Zr system, in which species V includes uranium as fuel and niobium (Nb) as an alloying agent, and species VI includes uranium and several additional transuranic elements. Species VII is a fuel mixture of the U—Mo—Ti—Zr system, in which niobium and several additional transuranic elements are included. Species VIII and IX and X and XI are fuel alloys within the U—Mo—W and U—Mo—Ta systems. In all these systems given as non-limiting examples, a fuel alloy including uranium and other metals provide a basis for an alloy with elevated solidus and reduced gamma onset temperature, compared to U—Zr alloys, that also tolerates addition of transuranics so that the melting temperature (solidus) is not depressed substantially below 1150 C, and preferably not below 1200 C. The above U-M, U—Pu-M, M systems can be analyzed, for example, by melting and casting the alloys into buttons or slugs and conducting testing to determine such properties as bcc-phase onset, solidus and liquidus temperatures. Exemplary analysis methods include, but are not limited to: differential scanning calorimetry; differential thermal analysis; and/or high-temperature methods such as dilatometry and thermal diffusivity testing. Quenching can be applied to isolate high-temperature phases, and long-term annealing can be applied for grain growth and traditional metallographic examination. Alloys can be characterized for microstructure (e.g., SEM) and for phase content (XRD). According to various embodiments, a method of making nuclear fission fuel is diagrammed as a flowchart in FIG. 11. The method 1000 includes at least the processes illustrated. In step 1002, at least one naturally fissioning actinide is provided. For example, U and U—Pu are provided in respective embodiments. Additional actinides and transuranic elements may be provided as well. For example, see species VI and VII in Table 6. In step 1004, molybdenum is provided. In step 1006, a metal other than molybdenum is provided. For example, various U—Mo—Ta, U—Mo—W, U—Mo—Ti and Mo—Ti—Zr systems are described herein. In step 1008, respective amounts of the naturally fissioning actinide, the molybdenum, and the metal are selected. In step 1010, a total weight of a fuel mixture is prepared by mixing the selected respective amounts of the naturally fissioning actinide, the molybdenum, and the metal. With regard again to step 1008, in at least one embodiment, the respective amounts are selected in a body-centered cubic solid phase region of a phase diagram of the fuel mixture. In at least one other embodiment, the respective amounts are selected in a solid phase region, such as body-centered cubic solid phase region, of a phase diagram of a composition including molybdenum and the other metal, in which the composition may or may not include an actinide. Nuclear fuel rods, slugs, and claddings have been mentioned in the above descriptions. For exemplary purposes, two embodiments of nuclear fuel slugs and their claddings in exploded perspective view are illustrated in FIG. 12. A cylindrical fuel slug 1102 composed of a fission fuel mixture according to one or more embodiments described herein and a corresponding cladding cylinder 1104 are shown. An annular fuel slug 106 composed of a fission fuel mixture according to one or more embodiments described herein and a corresponding cladding cylinder 1108 are also shown. A nuclear fuel rod or pin is prepared from either embodiment illustrated in FIG. 12 by insertion of either slug (1102, 1106) into its corresponding cladding (1104, 1108) and sealing of the cladding to encapsulate the slug. A plenum chamber for collection of gases within the cladding, and a sodium bond for thermal conduction, may be included. Although the present disclosure has been illustrated and described herein with reference to preferred embodiments and specific examples thereof, it will be readily apparent to those of ordinary skill in the art that other embodiments and examples may perform similar functions and/or achieve like results. All such equivalent embodiments and examples are within the spirit and scope of the present disclosure, are contemplated thereby, and are intended to be covered by the following claims.
044779216	description	DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENT Generally, the illustrated embodiment of an apparatus 10 for use in X-ray lithographic systems for the manufacture of LS1 and VLS1 circuits, or for X-ray lithography research, comprises a high intensity, soft X-ray lithography source tube 12 and a processing chamber 14 mounted in operative associated therewith. Soft X-ray radiation 16, generated in the well-shielded tube 12, is permitted to escape therefrom only through a window 18, preferably formed of beryllium. The beryllium window 18 separates, in an airtight manner, the X-ray lithography source tube 12 from the processing chamber 14. The X-ray radiation 16 escaping the tube 12 through the window 18 is designed to operate on a work, such as a wafer 20, deposited on a work support 22, with the radiation 16 passing through a suitable mask 24, as known. See U.S. Pat. Nos. 3,742,229 and 3,742,230, mentioned above and assigned to M.I.T. See also U.S. Pat. No. 4,215,192, Buckley, "X-ray Lithography Apparatus and Method of Use," granted July 29, 1980. Preferably, the processing chamber 14 is evacuated and/or filled with an inert gas, such as for example helium, lest the soft X-ray radiation 16 becomes attenuated in air. Soft X-rays are of poorer penetrating power when compared to traditional X-rays. A source of vacuum 26 is shown provided adjacent the processing chamber 14 for evacuating both the chamber 14 and the tube 12, the latter via a suitable hose 28. Inert gas, such as helium, is shown admitted to the chamber 14 via a tube 30 and exiting therefrom via a port 32. The X-ray lithography source tube 12 essentially comprises an electron beam source 34 and a watercooled target 36. Preferably, the electron beam source 34 is a barium impregnated tungsten ring provided with a shield grid 38 and an extraction grid 40. A beam focus tube 42 is positioned concentrically within the electron beam source 34 so as to limit the range of the electron beam spot size on the target 36. Further adjustment in the electron beam spot size on the target 36 is made by varying the voltage applied to the beam focus tube 42. A high-voltage feedthrough 44 provides the power for operating the electron beam source 34, the grids 38 and 40, and the beam focus tube 42. The water cooled target 36 preferably is shaped as an inverted cone, and is multi-layered, as will be more fully described below and with particular reference to FIG. 2. The target 36 is concentrically mounted, and axially displaced from the ring-shaped electron beam source 34, within a suitable jacket assembly 46. Preferably, the jacket assembly 46 is removably secured within the X-ray lithography source tube 12 so as to facilitate the inspection, and exchange if need be, of the target 36. A water inlet-outlet tube assembly 48 is concentrically, and removably, mounted within the jacket assembly 46. Tube assembly 48 is provided to direct a high velocity water flow at the target 36 during the operation of the X-ray lithography apparatus 10. As mentioned and known, an electron beam from the ring-shaped electron beam source 34 is designed to strike the target 36. See the article by J. S. Gaines and R. A. Hansen, quoted above. The impact of the electrons on the target 36 causes some X-ray emission therefrom just because of the sudden deceleration of the electrons when they collide with free electrons in the target material. This bremsstrahlung (meaning "braking radiation" in German) is not as intense, however, as is the X-ray emission emanating directly from the atoms of the target 36 material. This latter X-ray emission occurs when an electron in one of the innermost orbits of the target atom is literally "knocked out" of the target atom by a high energy electron emanating from the electron beam source 34. An electron from the next lowest orbit of the target atom immediately takes the place of the knocked-out missing electron. This change of orbit by the electron in the target atom causes energy to be released from the target 36 as electromagnetic radiation of a particular wavelength, also referred to as line emission. For the inner electrons of atoms with a high atomic number, this radiation is at X-ray wavelength. The X-ray wavelength region is generally divided between a short X-ray radiation band (about 0.5 to about two Angstroms wavelengths) and a soft X-ray radiation band (about two Angstroms to about 100 Angstroms wavelength). In general, shorter wavelength X-rays are less easily absorbed and are, therefore, more penetrating than soft X-rays. Each metal target 36 element possesses a characteristic X-ray spectrum. This characteristic X-ray spectrum may be in the K-region, as for instance a target 36 formed of aluminum, producing K-line radiation; or it may be in the M-region, as for instance a target 36 formed of tungsten, producing M-line radiation; or it may be in the L-region, as for instance a target 36 formed of palladium, producing L-line radiation. A designer of X-ray lithography apparatus is faced, among others, with various hurdles in his quest to come up with an apparatus that has high stability, long lifetime and is relatively maintenance free. With specific respect to designing a target anode, the designer knows that the target must be capable of sustained high intensity operation. This suggests the selection and use of a relatively high-melting-point material. However, the designer also knows that the choice of the target anode material is dictated, for the most part, by its characteristic X-ray emission wavelength, which differs from one target material to the next. This is so since the selected characteristic X-ray emission wavelength for the target anode must also be compatible with the mask 24 and the resist materials used in the X-ray lithographic system. See the above-quoted article by Maldonado et al. The designer also strives to obtain as short an exposure time, and thus to achieve a longer life for the target anode, as he can for the particular resist material used in his X-ray lithographic system. See the above-quoted U.S. Pat. No. 4,215,192 to Buckley. Prompted by these and other design considerations, we have discovered that the multi-layered composite target cone 36 provides the most-desired attributes in terms of high stability, long lifetime, low maintenance and expected short exposure time to the X-ray lithographic apparatus 10 of the invention. A longitudinal section of the multi-layered composite target cone 36 is shown on an enlarged scale in FIG. 2. The composite target cone 36 comprises at least an X-ray generating layer 50 and a water-interface layer 52. The X-ray generating layer 50 preferably is about five micrometers thick and is formed of a member of the class consisting of platinum, silver, palladium, rhodium, molybdenum, tungsten, silicon, aluminum and copper. The methodology of selecting the specific member of this class for the layer 50 will be more fully discussed below. The water-interface layer 52 of the composite target cone 36 is the layer in contact with the cooling water. Here, good thermal contact with the cooling water and high thermal conductivity to the cooling water are of paramount consideration. We consider that best results overall are achieved by selecting for the bulk of the water-interface layer 52, a layer of high thermal conductivity material 54, which preferably is about fourteen mil thick and preferably is formed of copper or silver. Furthermore, this layer of high thermal conductivity material 54 preferably is covered at the water contacting surface by a thin layer 56 of high corrosion resistance, such as a palladium layer of about several micrometers thickness. This thin layer 56 of high corrosion resistance serves to protect the layer 54 of high thermal conductivity from the nucleate boiling effect of the high velocity turbulent cooling water flowing along the backside of the composite target cone 36. We have also found it advantageous to have a further thin layer 58 of a high-melting-point material, such as for instance a 0.2 micrometer thick layer of tantalum, covering the layer 54 of high thermal conductivity material and facing the X-ray generating layer 50. This layer 58 of high-melting-point material serves to protect the layer 54 of high thermal conductivity material from the adverse effects of the very high working temperatures to which the composite target cone 36 gets exposed on the electron beam side. The composite target cone 36, constructed as above described, does not suffer from interdiffusion between the layers and exhibits no cracking problems despite prolonged use. Table 1 provides some calculated performance data, based on experimental work, for the different target elements used as an X-ray generating layer 50 in the composite target cone 36 of the X-ray lithography apparatus 10 of the invention. The electron beam source 34 was a barium-impregnated tungsten ring designed to develop a 10 KW electron beam on the target 36, the X-ray window 18 a one mil thick disk of 3/4" diameter, the X-ray beam 16 take-off angle was 12.5.degree., the X-ray lithography source tube 12 was evacuated to a pressure of 1.times.10.sup.-7 torr, the composite target cone 36 cooled by water at an input water pressure of about 160 psi @ 4 gal/min, and the applied power was: acceleration=20 KV @ 500 mA, cathode filament=20 V @ 10 amp, extractor grid=1 KV @ 30 mA, and the focus tube=3 KV @ 5 mA. Column four of Table 1 shows the radiant intensity, I, from each target element when forming the X-ray generating layer 50 of the composite target cone 36. The radiant intensity, I, is stated in milliwatts per steradian. Column five of Table 1 shows, for each target element forming the X-ray generating layer 50, the expected exposure time, T.sub.E, for PMMA resist. It is to be understood that these exposure times vary considerably for resists faster than PMMA. For example, when using PBS resist and aluminum, palladium or rhodium as the target element forming the X-ray generating layer 50 of the composite target cone 36, exposure times under two minutes have been achieved. The exposure time, T.sub.E, equals Q.sub.A /(.PHI..sub.S .alpha.), WHERE Q.sub.A is 575 J/CM.sup.3, and .PHI..sub.S is the radiant flux density (irradiance) of the resist surface, measured in microwatts per square centimeter, and .alpha. is the absorption coefficient of the resist in cm.sup.-1. TABLE 1 __________________________________________________________________________ X-RAY LITHOGRAPHY TARGET ELEMENT COMPARISON CHARACTERISTIC RADIANT EMISSION THERMAL INTENSITY, I EXPOSURE TARGET WAVELENGTH CONDUCTIVITY MW/SR TIME, T.sub.E ELEMENT .ANG. W/CM/.degree.C. @ 10 KW MIN __________________________________________________________________________ PT(L) 1.31 0.697 229 3233 AG(L) 4.15 4.18 161 222 PD(L) 4.37 0.705 152 211 RH(L) 4.60 0.879 148 196 MO(L) 5.41 1.45 111 200 W(M) 6.98 1.99 19.6 416 SI(K) 7.13 0.216 182 41 AL(K) 8.34 2.15 116 46 CU(L) 13.36 4.59 12 287 __________________________________________________________________________ The use of the composite target cone 36 permits, as mentioned, the ready exchange of one target element for another when working with different resist and mask materials. Thus it has been shown and discribed an X-ray lithography apparatus 10, featuring a composite target cone 36 in an X-ray lithography source tube 12 of the apparatus 10, which apparatus satisfies the objects and advantages set forth above. Since certain changes may be made in the present disclosure without departing from the scope of the present invention, it is intended that all matter described in the foregoing specification or shown in the accompanying drawings, be interpreted in an illustrative and not in a limiting sense.
	abstract	The edges of the reticle are detected with respect to the microstructured patterns exposed by the stepper, and the shapes of the microstructured patterns at the surface and at the bottom of the photoresist are detected. The microstructured patterns are evaluated by calculating, and displaying on the screen, the dislocation vector that represents the relationship in position between the detected patterns on the surface and at the bottom of the photoresist. Furthermore, dislocation vectors between the microstructured patterns at multiple positions in a single-chip or single-shot area or on one wafer are likewise calculated, then the sizes and distribution status of the dislocation vectors at each such position are categorized as characteristic quantities, and the corresponding tendencies are analyzed. Thus, stepper or wafer abnormality is detected.
061987867	description	DETAILED DESCRIPTION FIG. 1 is a schematic diagram of the basic parts of a power generating system 8. The system includes a BWR 10 which contains a reactor core 12. Water 14 is boiled using the thermal power of reactor core 12, passing through a water-steam phase 16 to become steam 18. Steam 18 flows through piping in a steam flow path 20 to a turbine flow control valve 22 which controls the amount of steam 18 entering steam turbine 24. Steam 18 is used to drive turbine 24 which in turn drives electric generator 26 creating electric power. Steam 18 flows to a condenser 28 where it is converted back to water 14. Water 14 is pumped by feedwater pump 30 through piping in a feedwater path 32 back to reactor 10. The above described system is generally referred to as a closed loop system. The equations below show the basic relationships between the generation of power in the reactor core Q, the steam flow rate .omega..sub.s, the feedwater flow rate .omega..sub.FW, the reactor system pressure P.sub.s upstream of turbine control valve 22, the pressure P.sub.cv, downstream of turbine control valve 22, and the pressure P.sub.c in condenser 28. Typically, the pressure in P.sub.c in condenser 28 is considered to be zero. Also, the main turbine control valve flow characteristic C.sub.v changes from a relatively small value to a large value as control valve 22 traverses from a nearly closed position to its wide open position. The flow coefficient of turbine 24 is expressed as C.sub.T which may be considered relatively constant for small changes in steam flow. Typically, the steam flow rate .omega..sub.s is equal to the feedwater flow rate .omega..sub.FW when there are no significant alternate sources of water into the reactor system nor any leakage from the reactor system. The following equations depict the basic steady state relationships were secondary variables, such as heat losses, pumping energy and leakage flows, are ignored. The basic equations for system pressure control by main turbine control valve modulation and core power modulation are developed below: The steam flow .omega..sub.s is a function of turbine control valve 22 position C.sub.v, and the pressure drop cross control valve 22 is the difference between the system pressure P.sub.s and the pressure down stream P.sub.cv of flow control valve 22, which can be expressed as: EQU .omega..sub.s =C.sub.v P.sub.s -P.sub.cv +L Equation 1 The steam flow through turbine control valve 22 and turbine 24 are equal when there are no shunt flow paths between turbine control valve 22 and the turbine inlet. The steam flow .omega..sub.s is a function of the turbine flow coefficient C.sub.T, and the pressure difference between the pressure down stream P.sub.cv of turbine control valve 22 and condenser 28 pressure P.sub.c may be considered equal to zero relative to the system pressure P.sub.s. The expression is: EQU .omega..sub.s =C.sub.t (P.sub.CV -0) Equation 2 Equations 1 and 2 can be combined to calculate the system pressure P.sub.s in terms of turbine control valve 22 flow coefficient C.sub.v, the turbine flow coefficient C.sub.T and the pressure down stream P.sub.cv of turbine control valve 22. ##EQU1## The thermal power Q from reactor core 12 is approximately proportional to steam flow .omega..sub.s. The proportionally constant K relates these two parameters. The equation for reactor core power is: EQU Q=K.omega..sub.S Equation 4 Combining Equations 2 and 4, the thermal power Q out of core 12 can be expressed as: EQU Q+KC.sub.T *P.sub.CV Equation 5 Solving for the pressure down stream of turbine flow control valve 22 the equation becomes: ##EQU2## Combining Equation 3 and Equation 6 the system pressure P.sub.s can be determined in terms of core 12 thermal power Q, the constant that relates core power to steam flow K, main turbine flow control valve coefficient C.sub.v, and the turbine flow coefficient C.sub.T : ##EQU3## For the variable which controls the system pressure P.sub.s for a conventional method of reactor pressure control by turbine flow control valve modulation, Equation 7 is rearranged to: ##EQU4## The terms ##EQU5## are relatively constant for constant reactor power. This equation shows that the reactor system pressure P.sub.s is proportional to the inverse of the square of the turbine control valve flow coefficient C.sub.v, which is linearly proportional to the position of turbine control valve 22 position as previously discussed. When turbine flow control valve 22 closes in response to a decrease in reactor system pressure, the steam flow decreases in response to this flow control valve position change, thus reducing the steam flow rate which causes the reactor system pressure to increase to the desired value and vice versa. For the variable which controls the system pressure P.sub.s for the method of reactor pressure control by modulation of the reactor power Q in accordance with the present invention, Equation 8 is rearranged to: ##EQU6## The terms ##EQU7## are relatively constant for constant steam flow. This equation shows that the reactor system pressure P.sub.s is proportional to the square of the core power Q for the term involving the control valve flow coefficient C.sub.v and linear with power for the term involving the turbine coefficient C.sub.T. The power Q is actually changed by changes in the control rod density in the reactor core or by changes in the flow through the reactor core. When the control rod density decreases or the flow through the reactor core increases in response to a decrease in reactor system pressure, the core power increases, which in turn, causes the reactor system pressure to increase back to the desired value and vice versa. FIG. 2 is a schematic flow diagram illustrating core thermal power modulation pressure control of power generating system 8 in accordance with one embodiment of the present invention. As described above, power generating system 8 includes BWR 10 that produces steam 18. Steam 18 flows from BWR 10 through steam path 20 to and through turbine control valve 22 to turbine 24 then to condenser 28 where steam 18 is converted to liquid water 14. Liquid water 14 then flows back to BWR 10 through feedwater flow path 32. Condenser water flow path 60, containing pump 58, connects condenser 28 with heat sink 62. Condenser water is pumped by pump 58 from condenser 28 to heat sink 62 and back to condenser 28 in closed loop flow path 60. Turbine 24 drives electric generator 21 generating electric power. Bypass valve 54 permits steam to flow directly from BWR 10 to condenser 28 bypassing turbine 24. A control rod drive 34 and control rod controller 36 change control rod density within core 12 of BWR 10 to vary or modulate the thermal output from core 12. Water recirculated through core 12 also is used to control thermal output. A recirculation pump 40 pumps water through piping in a recirculation flow path 42. Typically, recirculation pump 40 is a variable speed pump which provides for control and modulation of the recirculation water flow rate. A flow control valve 44 for controlling recirculation flow rate is also included in recirculation flow path 42. Recirculation controller 38 controls the speed of recirculation pump 40 and the operating open position of flow control valve 44. A pressure sensor 46 measures steam pressure in flow path 20. Operator control station 50 communicates with a pressure controller 48, a turbine valve controller 52 and a core thermal power controller 64. In turbine control valve modulation mode, system steam pressure is controlled by first measuring the steam pressure in steam path 20 with pressure sensor 46 which inputs the reading into pressure controller 48. A pressure setpoint is put into pressure controller 48 by the operator at operator control station 50. If the pressure is higher or lower than the setpoint pressure, a signal is sent to turbine valve controller 52 which in turn sends a signal to the main turbine control valve 22 to open or close. Opening turbine control valve 22 allows more steam into turbine 24 and thus lowers system pressure. Closing turbine control valve 22 creates higher pressure in the system. A boiling water reactor power generation plant may have more than one turbine control valve 22. Typically there are four turbine control valves 22 in the system which operate in either full arc mode where all valves move together, or partial arc mode where one or more valves modulate and the remaining valves stay in a full open position. Also, if a system safety pressure setpoint is exceeded, a signal is sent to bypass valve 54 to open to divert steam directly to condenser 28, bypassing turbine 24, and thereby lowering system pressure. Recirculation flow control 38 sends a signal to either variable speed recirculation pump 40 or to control valve 44 to control recirculation flow rate and thereby maintain a constant thermal output from core 12. Condenser 28 operates by condenser water removing thermal energy from steam 18 flowing from turbine 24 thereby converting steam 18 to water 14. The condenser water is pumped by a condenser pump 58 through piping in a closed loop flow path 60 from condenser 28 to a heat sink 62 and back to condenser 28. Heat sink 62 dissipates the thermal energy from the condenser water before it is recirculated to condenser 28. Typically, changeover from conventional control valve modulation pressure control mode, described above, to core thermal power modulation pressure control mode is effected by the plant operator at operator control station 50. However, changeover to core power modulation mode may be effected automatically when predetermined requirement parameters are satisfied. Steam pressure in steam flow path 20 is measured by pressure sensor 46 which sends an input to pressure controller 48 and core thermal power controller 64. Pressure controller 48 sends a signal to turbine valve controller 52 which in turn sends a signal to main turbine control valves 22 to open to a constant position. Control valves 22 are usually set to wide open, but may be set to any other constant setting. Control valves 22 are typically set to at least 75 percent of wide open. To moderate core thermal power, core thermal power controller may either control core power by moderating control rod density within the reactor or may moderate recirculation water flow rate through reactor core 12. To moderate control rod density, a signal is sent by core thermal power controller 64 to control rod drive controller 36. Control rod drive controller 36 then directs control rod drive 34 to either raise or lower the control rods thereby changing or modulating the control rod density in reactor core 12. The core thermal power is inversely proportional to control rod density. For example, as the control rod density increases thermal power decreases, and conversely as control rod density decreases, core thermal power increases. To moderate recirculation flow rate, core thermal power controller 64 sends a signal to recirculation flow controller 38. Controller 38 then causes variable speed pump 40 to change speed thus modulating recirculation flow rate. Alternatively, controller 38 sends a signal to recirculation control valve 44 to modulate the open position of valve 44, thus modulating the recirculation flow rate of water through reactor core 12. Modulating recirculating water flow rate modulates reactor core thermal power output. FIG. 3 is a schematic functional control block diagram illustrating core thermal power modulation pressure control of power generating system 8 in accordance with an exemplary embodiment of the present invention. FIG. 3 illustrates function blocks for a sensed system pressure 200, a steam line pressure sensor 202, a pressure setpoint adjustment 204, and a summer or compensator 206. These function blocks are typically included in a conventional pressure regulation function which provides a steam flow demand signal to the turbine control system 210. As known in the art, turbine control system 210 may typically include a valve position characterizer, a valve position controller, an electric signal to hydraulic flow converter, a hydraulic cylinder, flow control valves, a valve position sensor, and a hydraulic power unit. Turbine control system 210 also includes a turbine load limit setpoint 212 function block and an increase bypass valve close bias 214 function block. If the pressure increases over the turbine load limit setpoint 212, and the increase is over the bypass close bias, the bypass valves will open routing steam directly to the condenser. FIG. 3 also illustrates power regulator 220, power control fault logic 230, power control on 232, neutron flux 234 and power control bias 240 function blocks. Function blocks for recirculation pump variable speed control system 250, recirculation flow control valve position control system 254, and control rod position control system 260 are also illustrated. In operation, steam pressure in pipe 200 is measured by pressure sensor 202 which sends a signal to summer or compensator 206 which compares the pressure to pressure setpoint 204. A signal is then sent to the turbine control system 210 and to summer 208. When core power control 232 is turned on, power control fault logic 230 is activated. Power control fault logic 230 will monitor power control system 220 for control system failures, position of the bypass valves, level of neutron flux 234, and power control system 220 operating parameters for acceptable values. If a variable is out of tolerance or a control system hardware is in a failed condition, power control logic 230 will not allow transfer to core power modulation mode. Also, if the plant is operating in the power control mode, fault logic 230 will automatically transfer back to turbine control valve modulation mode to maintain acceptable system pressure. When power control mode on 232 is turned on, a power control bias 240 will add a set signal to summer 208 which also receives the value of turbine load limit 212. These signals are summed with the pressure error signal from summer 206. The control signal from summer 208 is input to the power regulator 220 which, for example, may be a proportional plus integral controller. The output from the power regulator 220 is provided to one of the power control systems which can be either the recirculation pump variable speed system 250, the recirculation flow control valve position system 254, or the control rod position control system 260. FIG. 4 illustrates an operational relationship between the percent of rated reactor core power versus the percent of rated core flow for a BWR. The operational domain of conventional turbine control valve modulation pressure control 300 has an upper boundary of line 310 which represents the operating power limit for control valve modulation mode. The operational domain of core thermal power modulation pressure control 320 has a lower boundary line 330. Line 330 is based on an acceptable system stability and plant transient behavior during transfer from core thermal power modulation mode to turbine control valve modulation mode. The upper boundary line 340 of domain 320 represents the maximum power generated with thermal power modulation mode pressure control. As illustrated in FIG. 4, the maximum power generated from a BWR using thermal power modulation pressure control is greater than if turbine control valve power modulation pressure control is used. From the preceding description of various embodiments 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 to 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.
046844929	claims	1. In a water-cooled nuclear reactor including an openable reactor pressure vessel having a wall, feedwater distributors passing through the pressure vessel wall for flooding the pressure vessel, and a core barrel for fuel elements disposed below the feedwater distributors in the reactor pressure vessel, a repair fixture comprising a sealing box enclosing the feedwater distributors in the opened, flooded pressure vessel, means connected to said sealing box for pressing said sealing box liquid-tightly against the reactor pressure vessel wall enclosing at least some of the feedwater distributors, and evacuating means connected to said sealing box for emptying the feedwater distributors enclosed by said sealing box. 2. Apparatus according to claim 1, wherein the feedwater distributors having conduit connections leading to primary shutdown controls of the pressure vessel. 3. Apparatus according to claim 1, including a support ring lowerable into the pressure vessel, said sealing box being mounted on said support ring. 4. Apparatus according to claim 3, including a core barrel disposed in the pressure vessel below the feedwater distributor, said support ring including props engaging the core barrel when said support ring is lowered. 5. Apparatus according to claim 3, wherein said support ring has an outer periphery including lateral props engaging the inner surface of the pressure vessel wall for absorbing radial compressive forces of said sealing box. 6. Apparatus according to claim 5, including setting cylinders displacing said lateral props against the inner surface of the pressure vessel wall. 7. Apparatus according to claim 3, including means connected to said sealing box for displacing said sealing box in radial direction of said support ring. 8. Apparatus according to claim 3, including means connected to said sealing box for displacing said sealing box in circumferential direction of said support ring. 9. Apparatus according to claim 3, wherein said support ring has an axis of symmetry, and including a guide connected to said support ring in which said sealing box is displaceable vertically and parallel to said axis of symmetry. 10. Apparatus according to claim 3, including hydraulic cylinders connected to said sealing box for controlling displacements of said sealing box relative to said support ring. 11. Apparatus according to claim 10, wherein said hydraulic cylinders are operated by pure water preventing contamination of the water of the flooded pressure vessel. 12. Apparatus according to claim 3, including guide cams connected to said sealing box and adapted to a given reactor type for correctly displacing said sealing box in radial direction relative to said support ring. 13. Apparatus according to claim 3, wherein said support ring is divided into sections. 14. Apparatus according to claim 3, wherein said feedwater distributors are in the form of four feedwater distributors connected to the pressure vessel, and including guides being screwed to said support ring and movable along the periphery of said support ring toward said feedwater distributors. PG,24 15. Apparatus according to claim 1, wherein the pressure vessel has an axis of symmetry and contains fuel elements including outer fuel elements disposed at a given distance from said axis of symmetry, and the feedwater distributors protrude into the pressure vessel, and including a support ring on which said sealing box is mounted, said sealing box having a depth in radial direction of the pressure vessel being only great enough to cover the protrusion of said feedwater distributors, and said support ring and said sealing box leaving an unobstructed inside diameter of the pressure vessel being more than twice said given distance when said sealing box is pressed against the pressure vessel wall. 16. Apparatus according to claim 1, wherein said sealing box has rims to be pressed against the pressure vessel wall, and gaskets disposed on said rims. 17. Apparatus according to claim 16, wherein each of said gaskets is substantially L-shaped and has a longer and a shorter leg, said longer leg being pressed against the pressure vessel wall and having a narrow sealing edge and a wide sealing lip, and a shallow fillet disposed between said sealing lip and sealing edge. 18. Apparatus according to claim 1, wherein said evacuating means include a pump. 19. Apparatus according to claim 18, including a float switch disposed in said sealing box and connected to said pump for controlling said pump. 20. Apparatus according to claim 1, wherein the pressure vessel has a given water level, and including an air venting hose connected to said sealing box and leading above the water level. 21. Apparatus according to claim 1, wherein said sealing box is at least partly transparent. 22. Apparatus according to claim 21, including transparent windows disposed in said sealing box.
	abstract	A method is provided for decontaminating biological pathogens residing in an enclosure of an electronic device. The method includes: identifying materials used to encase the enclosure of the electronic device; tailoring x-ray radiation to penetrate the materials encasing the enclosure; and directing x-ray radiation having a diffused radiation angle towards the electronic device.
044329334	abstract	In processes utilizing a source of laser energy for achieving a thermonuclear fusion reaction, it is necessary to have fusion fuel prepared in a configuration with minute dimensions and the present invention contemplates preparing this fuel by introducing into hollow microspheres (typically comprised of glass) of predetermined size, in the range of 250 to 2,000 micrometers, a gaseous thermonuclear fuel. One way of accomplishing this is to cause diffusion of gaseous thermonuclear fuel through the walls of the microsphere under conditions of suitable temperature and pressure so that it may be achieved in a reasonable time, after which the fuel can be frozen out on the walls of the microsphere to provide a fusion fuel in a hollow spherical shape. Suitable coatings of additional materials may be applied to the fueled microsphere by appropriate coating methods to complete complex thermonuclear fuel pellet configurations.
	description	The present invention relates to a nozzle repair method for repairing a nozzle provided in a nuclear reactor vessel and a nuclear reactor vessel provided with a nozzle. For example, a nuclear power plant that includes a pressurized water reactor (PWR) uses light water as a nuclear reactor coolant and a neutron moderator while keeping the light water as high-temperature and high-pressure water which is not boiled throughout a reactor core, sends the high-temperature and high-pressure water to a vapor generator so as to generate a vapor by a heat exchange operation, and sends the vapor to a turbine generator so as to generate electric power. In such a nuclear power plant, there is a need to periodically inspect various structures of the pressurized water reactor in order to ensure sufficient safety or reliability. Then, when a problem is found after various inspections, a necessary portion involved with the problem is repaired. For example, in the pressurized water reactor, a nuclear reactor vessel body is provided with a plurality of instrumentation nozzles penetrating a lower end plate. Further, each of the instrumentation nozzles is formed so that an in-core instrumentation guide pipe is fixed to the upper end thereof inside the reactor and a conduit tube is connected to the lower end thereof outside the reactor. Then, a neutron flux detector capable of measuring a neutron flux is insertable from the instrumentation nozzle to a reactor core (a fuel assembly) through the in-core instrumentation guide pipe by using the conduit tube. The instrumentation nozzle is formed in a manner such that an in-core instrumentation cylinder formed of a nickel base alloy is fitted into an attachment hole of a nuclear reactor vessel body formed of low-alloy steel and is welded by a nickel base alloy. For that reason, there is a possibility that a stress corrosion crack may occur in the in-core instrumentation cylinder due to the long-term use. Thus, when the stress corrosion crack occurs, there is a need to repair the instrumentation nozzle. A nozzle repair method of the related art is disclosed in, for example, Patent Literature 1 below. The nozzle repair method disclosed in Patent Literature 1 includes forming a buttered-grooving portion by excavating an inner surface of a vessel in a substantially cylindrical shape in an area including a J-beveling portion, inserting a plug including a plug body portion with an inner end surface forming an extension portion of an outer surface of the buttered-grooving portion and a protrusion protruding from the inner end surface and having the same axis as the buttered-grooving portion into a nozzle hole so that the inner end surface substantially matches the outer surface of the buttered-grooving portion, forming a buttered-welding portion by buttered welding the buttered-grooving portion, forming a J-beveling portion in the buttered-welding portion, and inserting and welding the nozzle. Patent Literature 1: Japanese Patent Application Laid-open No. 2011-075453 In the above-described nozzle repair method of the related art, the buttered-grooving portion is formed on the inner surface of the vessel, the plug is inserted into the nozzle hole, and the plug is buttered welded. Then, the J-beveling portion is formed, and the nozzle is inserted and welded. For that reason, a new nozzle may be attached with high precision, but the repair becomes complex. As a result, a repair cost increases. The invention is made to solve the above-described problems, and an object of the invention is to provide a nozzle repair method and a nuclear reactor vessel capable of improving workability and decreasing a repair cost by easily performing a repair operation. According to an aspect of the present invention, a nozzle repair method for an instrumentation nozzle in which an in-core instrumentation cylinder is inserted into an attachment hole formed in a semi-spherical portion of a nuclear reactor vessel and the inner surface side of the semi-spherical portion is groove-welded so as to fix the in-core instrumentation cylinder, comprises: removing a connection portion with respect to the in-core instrumentation cylinder in a groove-welding portion; removing the in-core instrumentation cylinder from the semi-spherical portion; forming a plug attachment portion by removing the groove-welding portion; applying a pressing load to the semi-spherical portion by attaching a plug to the plug attachment portion; and welding and fixing the plug attached to the plug attachment portion. Accordingly, the connection portion with respect to the in-core instrumentation cylinder in the groove-welding portion is removed, the in-core instrumentation cylinder is removed from the semi-spherical portion, the plug attachment portion is formed by removing the groove-welding portion, and the plug is welded and fixed while the plug is attached to the plug attachment portion and a pressing load is applied to the semi-spherical portion. That is, when the in-core instrumentation cylinder is removed from the instrumentation nozzle, the entire groove-welding portion is removed, and the plug is fixed, the instrumentation nozzle is set to an unavailable state. Since the nuclear reactor vessel is provided with the plurality of instrumentation nozzles, even when a small number of instrumentation nozzles may not be used, the other instrumentation nozzles may be used. Accordingly, since the repair operation may be easily performed when the instrumentation nozzle is repaired, the workability may be improved and the repair cost may be decreased. Further, since the plug is attached to the plug attachment portion and is welded thereto while the pressing load is applied to the semi-spherical portion, it is possible to prevent a problem in which a gap is formed between the plug and the plug attachment portion even when the plug heated by welding is cooled. Thus, it is possible to highly precisely fix the plug to the plug attachment portion in a close contact state. Advantageously, in the nozzle repair method, the semi-spherical portion is formed so that an inner surface is provided with a buttered-welding layer having stress corrosion cracking resistance, and the plug is fixed in a manner such that an outer peripheral portion is welded to the buttered-welding layer while the plug is fitted into the plug attachment portion. Accordingly, since the outer peripheral portion of the plug is welded to the buttered-welding layer while the plug is fitted into the plug attachment portion, the welding operation for fixing the plug substantially does not give an influence of heat to the nuclear reactor vessel, and the heat treatment is not needed. Thus, the repair operation may be easily performed and the workability may be improved. Advantageously, in the nozzle repair method, the plug is provided with an upward protrusion portion, and an outer surface of the protrusion portion is welded and fixed to the semi-spherical portion while the plug is fitted into the plug attachment portion. Accordingly, since the outer surface of the protrusion portion is fixed to the semi-spherical portion by welding while the plug is fitted into the plug attachment portion, the amount of the welding material decreases. Thus, the welding cost may be decreased and the welding operation may be easily performed. Advantageously, in the nozzle repair method, the protrusion portion includes a vertical wall portion which is formed along an upper outer peripheral portion of the plug, and a concave portion is provided inside the vertical wall portion. Accordingly, since the upper outer peripheral portion of the plug is provided with the vertical wall portion and the concave portion is provided at the inside thereof, the weight of the plug may be decreased, and hence the cost thereof may be decreased. Also, since the welding heat is radiated through the concave portion, the welding quality may be improved. Advantageously, in the nozzle repair method, the plug attachment portion is an opening larger than the attachment hole, and includes a support surface supporting the plug, and the plug receives a pressing load while a lower surface is pressed by the support surface. Accordingly, when the plug is attached to the plug attachment portion and is fixed to the semi-spherical portion, the weight of the plug is supported by the support surface of the plug attachment portion, and hence the welding portion for fixing the plug to the semi-spherical portion may be simplified. Further, since the plug is attached to the plug attachment portion and is welded while the pressing load is applied thereto, it is possible to suppress a problem in which a gap is formed between the lower surface of the plug and the support surface even when the plug heated by welding is cooled, and hence it is possible to fix the plug to the plug attachment portion in a close contact state. Advantageously, in the nozzle repair method, the plug receives a pressing load while a pulling load is applied from the outside of the semi-spherical portion thereto through the attachment hole. Accordingly, since a pulling load is applied to the plug from the outside of the semi-spherical portion, an operator does not need to enter the semi-spherical portion, and hence a pressing load may be easily applied to the plug. Thus, the workability may be improved. According to another aspect of the present invention, a nuclear reactor vessel comprises: a nuclear reactor vessel body of which a lower portion is formed in a semi-spherical shape; a nuclear reactor vessel head which is formed in a semi-spherical shape and is attached to an upper portion of the nuclear reactor vessel body; an inlet nozzle and an outlet nozzle which are provided at the side portion of the nuclear reactor vessel body; a reactor core which is disposed inside the nuclear reactor vessel body and is formed by a plurality of fuel assemblies; a plurality of control rods which is insertable into the fuel assemblies; a control rod driving mechanism which moves the control rods in the vertical direction; a plurality of instrumentation nozzles which is provided at the lower portion of the nuclear reactor vessel body and into which a neutron flux detector is insertable; a plug which plugs any nozzle attachment hole of the plurality of instrumentation nozzles; and a pressing load generating jig which is provided in the plug. Accordingly, since the lower portion of the nuclear reactor vessel body is provided with the plurality of instrumentation nozzles and the unnecessary instrumentation nozzles are plugged by the plug, the repaired nuclear reactor vessel may ensure high stress corrosion cracking resistance. In this way, the repair cost may be decreased and the stress corrosion cracking resistance may be improved. Further, since the plug is provided with the pressing load generating jig, a pressing load may be easily applied to the plug, and hence the workability may be improved. Advantageously, in the nuclear reactor vessel, the pressing load generating jig includes a screw shaft which is connected to a lower surface of the plug. Accordingly, since the screw shaft is connected to the lower surface of the plug in advance, it is possible to apply a pressing load of pressing the plug against the plug attachment portion. Advantageously, in the nuclear reactor vessel, a deformation member is provided between the lower surface of the plug and a support surface formed in the nozzle attachment hole. Accordingly, since the deformation member is provided between the lower surface of the plug and the support surface of the nozzle attachment hole, the deformation member is deformed when a pressing load of pressing the plug against the plug attachment portion is applied thereto, and hence the lower surface of the plug and the support surface of the nozzle attachment hole may easily contact each other. According to the nozzle repair method of the invention, the connection portion with respect to the in-core instrumentation cylinder in the groove-welding portion is removed, the in-core instrumentation cylinder is removed from the semi-spherical portion, the plug attachment portion is formed by removing the groove-welding portion, and the plug is welded and fixed while the plug is attached to the plug attachment portion and a pressing load is applied to the semi-spherical portion. Since the repair operation is easily performed compared to the case when the instrumentation nozzle is repaired, the workability may be improved and the repair cost may be decreased. At this time, it is possible to suppress a problem in which a gap is formed between the plug and the plug attachment portion, and hence to highly precisely fix the plug to the plug attachment portion in a close contact state. Further, according to the nuclear reactor vessel, since the lower portion of the nuclear reactor vessel body is provided with the plurality of instrumentation nozzles and the unnecessary instrumentation nozzles are plugged by the plug, the repaired nuclear reactor vessel may ensure high stress corrosion cracking resistance. In this way, the repair cost may be decreased and the stress corrosion cracking resistance may be improved. Further, since the plug is provided with the pressing load generating jig, a pressing load may be easily applied to the plug, and hence the workability may be improved. Hereinafter, preferred embodiments of a nozzle repair method and a nuclear reactor vessel according to the invention will be described in detail with reference to the accompanying drawings. Furthermore, the invention is not limited to the embodiment. Further, when a plurality of embodiments is provided, the embodiments may be combined with one another. FIG. 2 is a schematic configuration diagram of a nuclear power plant, and FIG. 3 is a longitudinal sectional view illustrating a pressurized water reactor. A nuclear reactor of the embodiment is a pressurized water reactor (PWR) that uses light water as a nuclear reactor coolant and a neutron moderator while keeping the light water as high-temperature and high-pressure water which is not boiled throughout a reactor core, sends the high-temperature and high-pressure water to a vapor generator so as to generate a vapor by a heat exchange operation, and sends the vapor to a turbine generator so as to generate electric power. In a nuclear power plant that includes the pressurized water reactor of the embodiment, as illustrated in FIG. 2, a containment 11 accommodates a pressurized water reactor 12 and a vapor generator 13 therein. Here, the pressurized water reactor 12 and the vapor generator 13 are connected to a high-temperature-side supply pipe 14 through a low-temperature-side supply pipe 15, the high-temperature-side supply pipe 14 is provided with a pressurizer 16, and the low-temperature-side supply pipe 15 is provided with a primary cooling water pump 17. In this case, light water is used as a moderator and primary cooling water (coolant), and a primary cooling system is controlled at a high-pressure state of about 150 to 160 atm by the pressurizer 16 in order to prevent the primary cooling water from being boiled in the reactor core portion. Accordingly, in the pressurized water reactor 12, the light water as the primary cooling water is heated by low-enriched uranium or MOX as a fuel (an atomic fuel), and the high-temperature primary cooling water is sent to the vapor generator 13 through the high-temperature-side supply pipe 14 while being maintained at a predetermined high pressure by the pressurizer 16. In the vapor generator 13, the primary cooling water which is cooled by a heat exchange operation between the high-temperature and high-pressure primary cooling water and the secondary cooling water is returned to the pressurized water reactor 12 through the low-temperature-side supply pipe 15. The vapor generator 13 is connected to a vapor turbine 32 through a pipe 31 that supplies the heated secondary cooling water, that is, vapor, and the pipe 31 is provided with a main vapor isolation valve 33. The vapor turbine 32 includes a high-pressure turbine 34 and a low-pressure turbine 35, and is connected to a generator (a generation device) 36. Further, a moisture separation heater 37 is provided between the high-pressure turbine 34 and the low-pressure turbine 35. Here, a cooling water branch pipe 38 which is branched from the pipe 31 is connected to the moisture separation heater 37, the high-pressure turbine 34 and the moisture separation heater 37 are connected to each other by a low-temperature reheating pipe 39, and the moisture separation heater 37 and the low-pressure turbine 35 are connected to each other by a high-temperature reheating pipe 40. Further, the low-pressure turbine 35 of the vapor turbine 32 includes a condenser 41. Here, the condenser 41 is connected to a turbine bypass pipe 43 which extends from the pipe 31 and includes a bypass valve 42, and is connected to a water intake pipe 44 and a drainage pipe 45 which supply and discharge the cooling water (for example, sea water). The water intake pipe 44 includes a circulation water pump 46, and the other end thereof is disposed under the sea along with the drainage pipe 45. Then, the condenser 41 is connected to a pipe 47, a condensate pump 48, a grand condenser 49, a condensed water desalting device 50, a condensate booster pump 51, and a low-pressure feed water heater 52. Further, the pipe 47 is connected to a deaerator 53, and is provided with a water feeding pump 54, a high-pressure feed water heater 55, and a water feeding control valve 56. Accordingly, in the vapor generator 13, the vapor which is generated by the heat exchange operation with respect to the high-pressure and high-temperature primary cooling water is sent to the vapor turbine 32 (from the high-pressure turbine 34 to the low-pressure turbine 35) through the pipe 31. Then, the vapor turbine 32 is driven by the vapor so that the generator 36 generates electric power. At this time, the vapor which is sent from the vapor generator 13 is used to drive the high-pressure turbine 34, passes through the moisture separation heater 37 so that the vapor is heated while a moisture contained in the vapor is removed, and is used to drive the low-pressure turbine 35. Then, the vapor having been used to drive the vapor turbine 32 is cooled into condensed water by the sea water in the condenser 41, and is returned to the vapor generator 13 through the grand condenser 49, the condensed water desalting device 50, the low-pressure feed water heater 52, the deaerator 53, the high-pressure feed water heater 55, and the like. In the pressurized water reactor 12 of the nuclear power plant with such a configuration, as illustrated in FIG. 3, a nuclear reactor vessel 61 includes a nuclear reactor vessel body 62 and a nuclear reactor vessel head (an upper end plate) 63 attached to the upper portion thereof so that an in-core structure is inserted thereinto, and the nuclear reactor vessel head 63 is fixed to the nuclear reactor vessel body 62 by a plurality of stud bolts 64 and a plurality of nuts 65 so as to be opened and closed. The nuclear reactor vessel body 62 has a cylindrical shape of which the upper portion is opened by the separation of the nuclear reactor vessel head 63 and the lower portion is formed in a semi-spherical shape while being closed by a lower end plate 66. Then, the upper portion of the nuclear reactor vessel body 62 is provided with an inlet nozzle (an entrance nozzle) 67 which supplies the light water (coolant) as the primary cooling water and an outlet nozzle (an exist nozzle) 68 which discharges the light water. Further, the nuclear reactor vessel body 62 is provided with a water injection nozzle (a water injection nozzle) (not illustrated) separately from the inlet nozzle 67 and the outlet nozzle 68. In the inside of the nuclear reactor vessel body 62, an upper core support 69 is fixed to a portion above the inlet nozzle 67 and the outlet nozzle 68 and a lower core support 70 is fixed so as to be located in the vicinity of the lower end plate 66. The upper core support 69 and the lower core support 70 are formed in a disk shape and are provided with a plurality of flow holes (not illustrated). Then, the upper core support 69 is connected to an upper core plate 72 provided with a plurality of flow holes (not illustrated) at a lower portion thereof through a plurality of reactor core support rods 71. A core barrel 73 which has a cylindrical shape is disposed inside the nuclear reactor vessel body 62 with a predetermined gap with respect to the inner wall surface. Further, the upper portion of the core barrel 73 is connected to the upper core plate 72, and the lower portion thereof is connected to a lower core support plate 74 having a disk shape and a plurality of flow holes (not illustrated) formed therein. Then, the lower core support plate 74 is supported by the lower core support 70. That is, the core barrel 73 is suspended on the lower core support 70 of the nuclear reactor vessel body 62. The reactor core 75 is formed by the upper core plate 72, the core barrel 73, and the lower core support plate 74, and the reactor core 75 has a plurality of fuel assemblies 76 disposed therein. Although not illustrated in the drawings, each of the fuel assemblies 76 is formed by binding a plurality of fuel rods in a grid shape by a support grid. Here, the upper nozzle is fixed to the upper end, and the lower nozzle is fixed to the lower end. Further, the reactor core 75 has a plurality of control rods 77 disposed therein. The plurality of control rods 77 is formed as a control rod cluster 78 while the upper ends are evenly arranged, and is insertable into the fuel assembly 76. In the upper core support 69, a plurality of control rod cluster guide pipes 79 is fixed while penetrating the upper core support 69, and each control rod cluster guide pipe 79 is formed so that the lower end thereof extends to the control rod cluster 78 inside the fuel assembly 76. The upper portion of the nuclear reactor vessel head 63 that constitutes the nuclear reactor vessel 61 is formed in a semi-spherical shape, and a magnetic jack type control rod driving mechanism 80 is accommodated in a housing 81 which is integrated with the nuclear reactor vessel head 63. The plurality of control rod cluster guide pipes 79 is formed so that the upper ends thereof extend to the control rod driving mechanism 80, and control rod cluster driving shafts 82 which extend from the control rod driving mechanism 80 extend to the fuel assemblies 76 while passing through the inside of the control rod cluster guide pipes 79, thereby gripping the control rod cluster 78. The control rod driving mechanism 80 extends in the vertical direction so as to be connected to the control rod cluster 78, and a control rod cluster driving shaft 82 of which the surface is provided with a plurality of circumferential grooves formed in the longitudinal direction is moved in the vertical direction by the magnetic jack, thereby controlling the output of the nuclear reactor. Further, the nuclear reactor vessel body 62 is provided with a plurality of instrumentation nozzles 83 which penetrates the lower end plate 66, and each of the instrumentation nozzles 83 is formed so that the upper end inside the reactor is connected to the in-core instrumentation guide pipe 84 and the lower end outside the reactor is connected to a conduit tube 85. In each of the in-core instrumentation guide pipes 84, the upper end is connected to the lower core support 70 and upper and lower connection plates 86 and 87 for suppressing a vibration are connected to the in-core instrumentation guide pipes. A thimble pipe 88 is provided with a neutron flux detector (not illustrated) capable of measuring a neutron flux, and is insertable to the fuel assembly 76 while penetrating the lower core support plate 74 from the conduit tube 85 along the instrumentation nozzle 83 and the in-core instrumentation guide pipe 84. Accordingly, the nuclear fission inside the reactor core 75 is controlled in a manner such that the control rod cluster driving shaft 82 is moved by the control rod driving mechanism 80 so as to extract the control rod 77 from the fuel assembly 76 by a predetermined amount. Then, the light water charged into the nuclear reactor vessel 61 is heated by the generated thermal energy, and the high-temperature light water is discharged from the outlet nozzle 68 so as to be sent to the vapor generator 13 as described above. That is, neutrons are discharged by the nuclear fission of the atomic fuel forming the fuel assembly 76, and the light water as the moderator and the primary cooling water decreases the movement energy of the discharged high-speed neutrons so as to form thermal neutrons. Accordingly, new nuclear fission may easily occur, and the generated heat is stolen and cooled. Meanwhile, when the control rod 77 is inserted into the fuel assembly 76, the number of neutrons generated inside the reactor core 75 may be adjusted. Further, when the entire control rod 77 is inserted into the fuel assembly 76, the nuclear reactor may be emergently stopped. Further, the nuclear reactor vessel 61 is formed so that an upper plenum 89 communicating with the outlet nozzle 68 is provided above the reactor core 75 and a lower plenum 90 is provided therebelow. Then, a down comer portion 91 which communicates with the inlet nozzle 67 and the lower plenum 90 is formed between the nuclear reactor vessel 61 and the core barrel 73. Accordingly, the light water flows from the inlet nozzle 67 into the nuclear reactor vessel body 62, flows downward to the down comer portion 91, reaches the lower plenum 90, rises while being guided upward by the spherical inner surface of the lower plenum 90, passes through the lower core support 70 and the lower core support plate 74, and flows into the reactor core 75. The light water which flows into the reactor core 75 increases in temperature while cooling the fuel assembly 76 by absorbing the thermal energy generated from the fuel assembly 76 constituting the reactor core 75, passes through the upper core plate 72, rises to the upper plenum 89, and is discharged through the outlet nozzle 68. In the nuclear reactor vessel 61 with such a configuration, the instrumentation nozzle 83 is formed in a manner such that the in-core instrumentation cylinder is fitted into an attachment hole formed in the lower end plate 66 of the nuclear reactor vessel body 62 and the upper end of the in-core instrumentation cylinder is fixed to the inner surface of the lower end plate 66 by groove-welding. In this case, the nuclear reactor vessel body 62 is formed by buttered-welding a stainless steel to the inner surface of low-alloy steel as a base material, and the in-core instrumentation cylinder of the nickel base alloy is welded to the nuclear reactor vessel body 62 by the material of the nickel base alloy while being fitted into the attachment hole of the nuclear reactor vessel body 62. For that reason, there is a possibility that a stress corrosion crack may occur in the in-core instrumentation cylinder due to the long-term use. Thus, when the stress corrosion crack occurs, there is a need to repair the instrumentation nozzle 83. However, since the nuclear reactor vessel body 62 is formed of low-alloy steel, a heat treatment for removing a stress is needed after the welding operation. However, the heat treatment is difficult in that the inside of the nuclear reactor vessel body is a high radiation field. Therefore, a nozzle repair method of the first embodiment includes removing a connection portion with respect to the in-core instrumentation cylinder in the groove-welding portion, removing the in-core instrumentation cylinder from the lower end plate 66, processing the plug attachment portion by removing the groove-welding portion, applying a pressing load to the lower end plate 66 by attaching a plug to the plug attachment portion, and fixing the plug attached to the plug attachment portion by welding. At this time, since the plug is fixed by removing the groove-welding portion and the in-core instrumentation cylinder, the instrumentation nozzle 83 may not be used. Since the nuclear reactor vessel 61 is provided with a plurality of instrumentation nozzles 83, even when a small number of instrumentation nozzles 83 may not be used, the other instrumentation nozzles 83 may be used. Accordingly, since the repair operation may be easily performed compared to the case when the instrumentation nozzle 83 is repaired, the workability may be improved and the repair cost may be decreased. Further, since the plug is attached to the plug attachment portion and is welded thereto while the pressing load is applied to the lower end plate 66, it is possible to prevent a problem in which a gap is formed between the plug and the plug attachment portion even when the plug heated by welding is cooled. Accordingly, it is possible to highly precisely fix the plug to the plug attachment portion in a close contact state. FIG. 1 is a cross-sectional view illustrating an instrumentation nozzle of a nuclear reactor vessel which is repaired by a nozzle repair method according to the first embodiment of the invention, FIG. 4 is a flowchart illustrating the nozzle repair method of the first embodiment, FIG. 5 is a schematic diagram illustrating an operation of drawing a thimble tube, FIG. 6 is a schematic diagram illustrating an operation of removing an in-core structure from a nuclear reactor vessel, FIG. 7 is a schematic diagram illustrating a water stopping operation in an in-core instrumentation cylinder, FIG. 8 is a schematic diagram illustrating an operation of cutting a conduit tube, FIG. 9 is a schematic diagram illustrating an operation of attaching a water stopping cap, FIG. 10 is a schematic diagram illustrating an operation of cutting the in-core instrumentation cylinder, FIG. 11-1 is a schematic diagram illustrating an operation of trepanning the in-core instrumentation cylinder, FIG. 11-2 is a cross-sectional view illustrating the trepanned in-core instrumentation cylinder, FIG. 12 is a cross-sectional view illustrating an operation of drawing the in-core instrumentation cylinder, FIG. 13 is a schematic diagram illustrating an operation of machining a groove-welding portion in the instrumentation nozzle, FIG. 14 is a schematic diagram illustrating an operation of machining a plug attachment portion in the instrumentation nozzle, FIG. 15 is a schematic diagram illustrating an operation of attaching a plug to the instrumentation nozzle, FIG. 16 is a schematic diagram illustrating an operation of attaching a housing to the instrumentation nozzle, FIG. 17 is a schematic diagram illustrating an operation of applying a pressing force to the plug, FIG. 18-1 is a schematic diagram illustrating the instrumentation nozzle in which a pressing force is applied to the plug, FIG. 18-2 is a diagram illustrating a pressing force for the plug, FIG. 19 is a schematic diagram illustrating an operation of welding the plug to the instrumentation nozzle, and FIG. 20 is a schematic diagram illustrating the instrumentation nozzle to which the plug is welded. Hereinafter, a nozzle repair method of the first embodiment will be described in detail with reference to the cross-sectional view of FIG. 1, the flowchart of FIG. 4, and the schematic diagrams from FIGS. 5 to 20. As illustrated in FIGS. 4 and 5, in step S11, in the pressurized water reactor 12, the lower portion of the nuclear reactor vessel 61 is shielded by a shielding member 101, the conduit tube 85 is shielded by a shielding member 102, and the thimble tube (the neutron flux detector) 88 inserted into the nuclear reactor vessel 61 is extracted to the outside. Then, as illustrated in FIGS. 4 and 6, in step S12, in the pressurized water reactor 12, the nuclear reactor vessel head 63 is separated from the nuclear reactor vessel body 62 constituting the nuclear reactor vessel 61, and an in-core structure (an upper in-core structure 12A and a lower in-core structure 12B) accommodated therein is raised and removed by a suspending tool 153. In this case, a nuclear reactor building 111 is provided with a cavity 112 capable of storing cooling water, and an appliance temporary placement pool 114 is formed near a nuclear reactor pool 113 where the pressurized water reactor 12 is supported in a suspended state. For that reason, the upper in-core structure 12A and the lower in-core structure 12B are temporarily disposed while being immersed into the cooling water of the appliance temporary placement pool 114. As illustrated in FIG. 7, in the nuclear reactor vessel body 62, an inner surface of a base material 201 formed of low-alloy steel is provided with a buttered-welding layer 202 formed of stainless steel. Then, the instrumentation nozzle 83 has a configuration in which an in-core instrumentation cylinder 204 formed of a nickel base alloy (for example, Inconel 600/trademark) is inserted and positioned into an attachment hole 203 formed in the lower end plate 66 of the nuclear reactor vessel body 62 in the vertical direction and a groove-welding portion 206 (a lower welding portion 206a and a main welding portion 206b) formed of a nickel base alloy (for example, Inconel 600) is provided in a grooving portion 205 formed in the inner surface of the lower end plate 66. As illustrated in FIGS. 4 and 7, in step S13, a water stopping plug handling device (not illustrated) is provided above the cavity 112 and a water stopping plug attachment device (not illustrated) gripping a water stopping plug 115 moves downward inside the cooling water of the cavity 112. Then, the water stopping plug 115 is fitted to the upper end of the in-core instrumentation cylinder 204 constituting the instrumentation nozzle 83 of the nuclear reactor vessel body 62 so as to plug the upper end. Further, as illustrated in FIGS. 4 and 8, in step S14, the conduit tube 85 connected to the lower end of the in-core instrumentation cylinder 204 is cut. At this time, the cooling water inside the in-core instrumentation cylinder 204 is removed, and a foreign matter mixing preventing head is attached to the in-core instrumentation cylinder 204 in the instrumentation nozzle 83 which does not need to be repaired. As illustrated in FIGS. 4 and 9, in step S15, a water stopping cap 116 is fixed to the lower portion of the instrumentation nozzle 83. In this case, the water stopping cap 116 includes a casing 116a of which an upper end is opened and a lower end is closed, a pipe 116b which is connected to the lower portion of the casing 116a, and an opening/closing valve 116c which is provided in the pipe 116b. Meanwhile, the outer surface of the lower end plate 66 is provided with a buttered-welding layer 207 formed of the stainless steel in advance, and the buttered-welding layer 207 is inspected by an ultrasonic inspection device. For that reason, the water stopping cap 116 has a configuration in which the upper end of the casing 116a is welded and fixed to the buttered-welding layer 207 of the lower end plate 66 so as to cover the lower portion of the in-core instrumentation cylinder 204 from the downside. In this case, it is checked whether a leakage occurs by inspecting the welding portion of the casing 116a through a visual test using a camera (not illustrated). Furthermore, a jack 117 is provided inside the water stopping cap 116 so as to support the in-core instrumentation cylinder 204. When the water is stopped at the upper and lower ends of the existing in-core instrumentation cylinder 204 of the instrumentation nozzle 83, as illustrated in FIGS. 4 and 10, in step S16, the upper portion of the in-core instrumentation cylinder 204 in the instrumentation nozzle 83 is cut (or machined) by a cutting device (not illustrated), and the upper portion of the cut in-core instrumentation cylinder 204 is collected. As illustrated in FIGS. 4 and 11-1, in step S17, the groove-welding portion 206 of the in-core instrumentation cylinder 204 fixed to the lower end plate 66 is trepanned (as a trepanning portion 208) by a machining device (not illustrated), and as illustrated in FIG. 11-2, an opening gap 209 is formed between the in-core instrumentation cylinder 204 and the groove-welding portion 206. That is, the trepanning portion 208 as the connection portion with respect to the in-core instrumentation cylinder 204 in the groove-welding portion 206 is removed. At this time, the trepanning process is performed from the upper end of the groove-welding portion 206, that is, the inner surface of the lower end plate 66 to the downside of the groove-welding portion 206, that is, a base material 201 of the lower end plate 66. Furthermore, when the groove-welding portion 206 of the in-core instrumentation cylinder 204 is trepanned by a cutting device, produced chips are collected by a suction device (not illustrated). As illustrated in FIGS. 4 and 12, in step S18, the in-core instrumentation cylinder 204 is extracted and collected upward from the attachment hole 203 of the lower end plate 66 by using an extraction device (not illustrated). At this time, the jack 117 inside the water stopping cap 116 may be operated (lengthened) so as to press the in-core instrumentation cylinder 204 upward. Then, as illustrated in FIGS. 4 and 13, in step S19, the groove-welding portion 206 provided in the inner surface of the lower end plate 66 is removed by a machining device (or a discharge processing device or the like) not illustrated in the drawings. In step S20, the groove-welding portion 206 is removed by machining, and the remaining portion or the defect (crack) of the groove-welding portion 206 in a machined surface 210 is inspected by an eddy current inspection device. Here, when the entire groove-welding portion 206 is removed and the defect (crack) is not found, as illustrated in FIGS. 4 and 14, in step S21, a plug attachment opening (a plug attachment portion) 211 is formed in the inner surface of the lower end plate 66 by a machining device (or a discharge processing device or the like) not illustrated in the drawings. At this time, the plug attachment opening 211 is a columnar opening which is formed in the inner surface of the lower end plate 66. Here, it is desirable to align the axis of the plug attachment opening 211 to the axis of the attachment hole 203. Accordingly, the plug attachment opening 211 is formed so that an inner peripheral surface 211a becomes a vertical surface and an annular bottom surface (a support surface) 211b becomes a horizontal surface. Furthermore, when the inner surface (the buttered-welding layer 202) of the lower end plate 66 or the inner surface of the plug attachment opening 211 is not sufficiently processed, a finishing process is performed by a polishing device (not illustrated), so that the inner peripheral surface 211a and the bottom surface 211b are formed as flat surfaces. Further, it is checked whether the depth or the width (the inner diameter) of the plug attachment opening 211 is a predetermined depth or a predetermined width by measuring the depth or the width through a visual test using a camera (not illustrated). Then, as illustrated in FIGS. 4 and 15, in step S22, a plug 212 is attached to the plug attachment opening 211 by fitting. The plug 212 includes a columnar plug body 212a which has an outer diameter to be fitted into the plug attachment opening 211, a vertical wall portion 212b which is formed along an upper outer peripheral portion as a protrusion portion protruding upward in the upper surface of the plug body 212a, a concave portion 212c which is provided inside the vertical wall portion 212b, a protrusion 212d which protrudes downward in the lower surface of the plug body 212a, and a flat lower surface 212e which is provided around the protrusion 212d in the lower surface of the plug body 212a. Further, the plug 212 is formed so that the lower portion is integrated with a screw shaft 213 as a pressing load generating jig. The screw shaft 213 extends from the lower surface of the protrusion 212d in the plug 212 toward the attachment hole 203, and the front end (the lower portion) is provided with a screw portion 213a. In this case, the screw shaft 213 is formed so as to have a length in which the screw portion 213a extends from the attachment hole 203 to the outside of the lower end plate 66 when the plug 212 is attached to the plug attachment opening 211. When the plug 212 is fitted into the plug attachment opening 211, the plug body 212a is filled in the plug attachment opening 211, the protrusion 212d is fitted into the attachment hole 203, and the screw shaft 213 is inserted into the attachment hole 203. For that reason, in the plug 212, the annular vertical wall portion 212b protrudes upward from the inner surface of the lower end plate 66, and the concave portion 212c is flush with the inner surface of the lower end plate 66. Furthermore, when the plug 212 is fitted into the plug attachment opening 211, a joint surface between the plug 212 and the plug attachment opening 211 is inspected by a visual test using a camera (not illustrated). Here, as illustrated in FIGS. 4 and 16, an aerial space is formed in the upper portion of the instrumentation nozzle 83, that is, the periphery of the plug attachment opening 211 into which the upper portion of the attachment hole 203 is inserted, and the plug 212 is fixed to the aerial space. That is, a housing 121 of which a lower portion is opened is disposed above the plug attachment opening 211 in the lower end plate 66, and is fixed through a seal member 122 so as to seal an inner space. Then, the cooling water in the space is discharged through a pipe (not illustrated), so that the inside is formed as an aerial space A. Subsequently, a welding device 123 is carried into the housing 121 through a carrying pipe (not illustrated), and the water stopping cap 116 is removed. In this case, the housing 121 is fixed to the upper position of the plug attachment opening 211 in the lower end plate 66 by the seal member 122, the opening/closing valve 116c is opened, and the cooling water inside the housing 121 is discharged from the pipe 116b of the water stopping cap 116, so that the inside may be formed as the aerial space A. Then, as illustrated in FIGS. 4 and 17, in step S23, a pressing load is applied from the outside of the lower end plate 66 to the plug 212 attached to the plug attachment opening 211. First, the front end of the screw shaft 213 which protrudes from the attachment hole 203 at the outside of the lower end plate 66 penetrates a support plate (a pressing load generating jig) 214, and is temporarily fixed to the buttered-welding layer 207 of the lower end plate 66. Next, a nut (a pressing load generating jig) 215 is held by a nut holding portion 127 of a nut rotation tool 126, and the nut 215 is pressed against the screw portion 213a of the screw shaft 213 while the nut holding portion 127 is rotated. Then, as illustrated in FIG. 18-1, the nut 215 is threaded into the screw portion 213a of the screw shaft 213, so that the plug 212 is fixed. Here, when the nut 215 is further threaded into the screw portion 213a of the screw shaft 213, the plug 212 is pulled toward the downside, that is, the outside of the lower end plate 66 through the screw shaft 213 from the support plate 214 as the original point. Then, the lower surface 212e of the plug 212 is pressed against the bottom surface 211b of the plug attachment opening 211, and a pressing load exerted toward the lower end plate 66 is applied to the plug 212. In this case, as illustrated in FIG. 18-2, since the plug 212 which is fixed to the plug attachment opening 211 is formed so that a pressing load exerted toward the lower end plate 66 is applied thereto, the lower surface 212e is pressed against the bottom surface 211b. Then, as indicated by the two-dotted chain line of FIG. 18-2, the lower surface 212e of the plug 212 is deformed while being compressed, and the bottom surface 211b of the plug attachment opening 211 is slightly deformed while being compressed. Then, as illustrated in FIGS. 4 and 19, in step S24, the outer peripheral surface of the vertical wall portion 212b of the plug 212 and the inner surface of the lower end plate 66, that is, the surface of the buttered-welding layer 202 are fillet-welded to each other by the welding device 123 carried into the housing 121 while a welding head 125 is moved by a movement device 124. Then, as illustrated in FIG. 20, when a fillet-welding portion 216 is formed at the corner of the surface of the buttered-welding layer 202 in the lower end plate 66 and the outer peripheral surface of the vertical wall portion 212b while the plug 212 is fitted into the plug attachment opening 211, the plug 212 is fixed to the lower end plate 66. When the plug 212 is welded, the plug 212 is thermally expanded by the welding heat, and is contracted in a cooled state. For that reason, the plug 212 is contracted upward from the fillet-welding portion 216 as the original point after the plug is cooled. Thus, there is a concern that a gap may be formed between the lower surface 212e and the bottom surface 211b of the plug attachment opening 211. Incidentally, in the embodiment, as described above, the plug 212 is welded while being compressed and deformed by a pressing load applied toward the lower end plate 66. Accordingly, even when the plug is thermally expanded by the welding heat and is contracted upward in a cooled state from the fillet-welding portion 216 as the original point, the lower surface 212e is not separated from the bottom surface 211b of the plug attachment opening 211, and a gap is not formed therebetween. Furthermore, as the material of the plug 212 attached to the plug attachment opening 211 and the welding material (the fillet-welding portion 216) used to fix the plug 212 to the inner surface of the lower end plate 66, it is desirable to use a nickel base alloy (for example, Inconel 690) as a welding material having higher stress corrosion cracking resistance than nickel base alloy (for example, Inconel 600) as the welding material of the groove-welding portion 206 or the existing in-core instrumentation cylinder 204. However, the plug 212 or the welding material (the fillet-welding portion 216) may be formed of the same material as the existing in-core instrumentation cylinder 204 or the groove-welding portion 206. For example, both may be formed of stainless steel. Subsequently, in step S25, the fillet-welding portion 216 is inspected by a visual test using a camera (not illustrated). Here, when a sealing performance is ensured due to the non-existence of the leakage in the fillet-welding portion 216, the housing 121 or the welding device 123 is removed. Further, the foreign matter mixing preventing head which is attached to the in-core instrumentation cylinder 204 in the peripheral instrumentation nozzle 83 that does not need to be repaired is separated. Furthermore, here, the aerial space A is formed around the plug attachment opening 211, and the plug 212 is welded in the aerial space A. However, for example, the water may be stopped at the upper end of the nuclear reactor vessel body 62 by the seal plate, and the entire water therein may be discharged so as to form the aerial space. Then, as illustrated in FIGS. 4 and 5, in step S26, the in-core structure (the upper in-core structure 12A and the lower in-core structure 12B) is returned into the nuclear reactor vessel body 62, the nuclear reactor vessel head 63 is attached, and the thimble tube (the neutron flux detector) 88 which is extracted to the outside is inserted into the nuclear reactor vessel 61 so as to be restored. In the pressurized water reactor 12, the shielding member 101 below the nuclear reactor vessel 61 and the shielding member 102 of the conduit tube 85 are removed. As illustrated in FIG. 1, in the repaired instrumentation nozzle 83, the plug attachment opening 211 is formed in the inner surface of the lower end plate 66 at the upper end of the attachment hole 203 in the attachment hole 203 of the nuclear reactor vessel body 62 having the buttered-welding layer 202 formed of stainless steel as the inner surface of the base material 201 formed of low-alloy steel, the plug 212 is fitted into the plug attachment opening 211, and the outer peripheral surface of the vertical wall portion 212b of the plug 212 and the buttered-welding layer 202 of the lower end plate 66 are fixed and closed by the fillet-welding portion 216. Then, the plug 212 is fitted and fixed to the plug attachment opening 211, and the lower surface 212e is held while contacting the bottom surface 211b of the plug attachment opening 211. Furthermore, although a pressing load exerted toward the lower end plate 66 is applied to the plug 212 while the screw shaft 213 is fixed to the lower end plate 66 by the nut 215 through the support plate 214, this configuration is used in that the pressing load does not give a bad influence on the lower end plate 66. However, the support plate 214 and the nut 215 may be separated from the screw shaft 213 while the nut 215 is released or the screw shaft 213 may be also removed by cutting. In this way, the nozzle repair method of the first embodiment includes removing the connection portion (the trepanning portion 208) with respect to the in-core instrumentation cylinder 204 in the groove-welding portion 206, removing the in-core instrumentation cylinder 204 from the lower end plate 66, forming the plug attachment opening 211 by removing the groove-welding portion 206, applying a pressing load to the lower end plate 66 by attaching the plug 212 to the plug attachment opening 211, and welding and fixing the plug 212 attached to the plug attachment opening 211. Accordingly, the trepanning portion 208 with respect to the in-core instrumentation cylinder 204 in the groove-welding portion 206 is removed, the in-core instrumentation cylinder 204 is removed from the lower end plate 66, the plug attachment opening 211 is formed by removing the groove-welding portion 206, and the plug 212 is welded and fixed to the plug attachment opening 211. That is, when the in-core instrumentation cylinder 204 is removed from the instrumentation nozzle 83, the entire groove-welding portion 206 is removed, and the plug 212 is fixed, the instrumentation nozzle 83 is set to an unavailable state in a closed state. Since the lower end plate 66 of the nuclear reactor vessel 61 is provided with the plurality of instrumentation nozzles 83, even when a small number of instrumentation nozzles 83 may not be used, the other instrumentation nozzles 83 may be used. Accordingly, since the repair operation is easily performed when the instrumentation nozzle 83 is repaired, the workability may be improved and the repair cost may be decreased. When the plug 212 is welded to the plug attachment opening 211, the welding operation is performed in a state where a pressing load exerted toward the lower end plate 66 is applied to the plug 212 fitted into the plug attachment opening 211 in advance. For that reason, it is possible to suppress a problem in which a gap is formed between the plug 212 and the plug attachment opening 211 even when the plug 212 heated by welding is cooled and hence to highly precisely fix the plug 212 to the plug attachment opening 211 in a close contact state. In the nozzle repair method of the first embodiment, the lower end plate 66 has a configuration in which the buttered-welding layer 202 having stress corrosion cracking resistance is provided on the inner surface of the base material 201 and the outer peripheral portion is fixed to the buttered-welding layer 202 by welding while the plug 212 is fitted into the plug attachment opening 211. Accordingly, the welding operation for fixing the plug 212 substantially does not give an influence of heat on the base material 201 of the nuclear reactor vessel body 62, and the heat treatment is not needed. Thus, the repair operation may be easily performed, and the workability may be improved. In the nozzle repair method of the first embodiment, the outer peripheral surface of the vertical wall portion 212b is fixed to the buttered-welding layer 202 by welding while the plug 212 is provided with the upward vertical wall portion 212b and the plug 212 is fitted into the plug attachment opening 211. Accordingly, the amount of the welding material decreases. Thus, the welding cost may be decreased and the welding operation may be easily performed. In the nozzle repair method of the first embodiment, a concave portion 211c is provided at the inside of the vertical wall portion 211b of the plug 212. Accordingly, since the weight of the plug 212 may be decreased, the cost may be decreased. Also, since the welding heat is radiated through the concave portion 211c, the welding quality may be improved. In the nozzle repair method of the first embodiment, the plug attachment opening 211 is formed after the groove-welding portion 206 is removed by machining and the non-existence of the defect in the machined surface 210 is checked. Accordingly, since it is possible to reliably check a portion having a stress corrosion crack by checking the existence of the defect in the machined surface 210 obtained by machining the groove-welding portion 206, it is possible to maintain the high-quality nuclear reactor vessel body 62. In the nozzle repair method of the first embodiment, the plug attachment opening 211 is an opening larger than the attachment hole 203, the bottom surface 211b supporting the plug 212 is provided, and the lower surface 212e of the plug 212 is pressed against the bottom surface 211b so as to apply a pressing load thereto. Accordingly, when the plug 212 is attached and fixed to the plug attachment opening 211, the weight of the plug 212 is supported by the bottom surface 211b of the plug attachment opening 211, and hence the welding portion for fixing the plug 212 to the lower end plate 66 may be simplified. Further, since the plug 212 is attached to the plug attachment opening 211 and is welded thereto while the pressing load is applied thereto, it is possible to suppress a problem in which a gap is formed between the lower surface 212e of the plug 212 and the bottom surface 211b even when the plug 211 heated by welding is cooled and hence to fix the plug 212 to the plug attachment opening 211 in a close contact state. In the nozzle repair method of the first embodiment, a pressing load is applied to the plug 212 in a manner such that a pulling load is applied thereto through the attachment hole 203 from the outside of the lower end plate 66. Accordingly, the operation is performed on the plug 212 from the outside of the lower end plate 66, the operator does not need to enter the inside of the lower end plate 66, and the pressing load may be easily applied to the plug 212. Thus, the workability may be improved. In the nozzle repair method of the first embodiment, the lower portion of the nuclear reactor vessel 61 is shielded by the shielding member 101, the conduit tube 85 is shielded by the shielding member 102, the thimble tube 88 inserted into the nuclear reactor vessel 61 is extracted to the outside, the in-core structure is removed from the nuclear reactor vessel body 62, the groove-welding portion 206 and the in-core instrumentation cylinder 204 are removed under the water, and the processing of the plug attachment opening 211 and the welding of the plug 212 are performed in the atmosphere. Accordingly, it is possible to improve the safety by reducing the exposure. Further, in the nuclear reactor vessel of the first embodiment, the plurality of instrumentation nozzles 83 is provided in the lower end plate 66 of the nuclear reactor vessel body 62, and any attachment hole 203 of the plurality of instrumentation nozzles 83 is plugged by the plug 212 formed of a nickel base alloy having high stress corrosion cracking resistance. Accordingly, the unnecessary instrumentation nozzle 83 provided in the lower end plate 66 of the nuclear reactor vessel body 62 is plugged by the plug 212, and the repaired nuclear reactor vessel 61 ensures high stress corrosion cracking resistance. Thus, the repair cost may be decreased and the stress corrosion cracking resistance may be improved. Then, the screw shaft 213, the support plate 214, and the nut 215 are provided as the pressing load generating jig for the plug. Accordingly, the pressing load may be easily applied to the plug 212, and hence the workability may be improved. In this case, the pressing load of pressing the plug 212 against the plug attachment opening 211 may be easily applied to the plug in a manner such that the screw shaft 213 is fixed to the lower surface of the plug 212 in advance. Furthermore, in the first embodiment, the lower portion of the nuclear reactor vessel 61 and the conduit tube 85 are shielded, the thimble tube 88 is extracted to the outside from the nuclear reactor vessel 61, the in-core structure is removed from the nuclear reactor vessel body 62, the groove-welding portion 206 and the in-core instrumentation cylinder 204 are removed under the water, and the processing of the plug attachment opening 211 and the welding of the plug 212 are performed in the atmosphere. However, the invention is not limited thereto. For example, a configuration may be employed in which the in-core structure is removed from the nuclear reactor vessel body 62, a thimble stand is disposed inside the nuclear reactor vessel 61, the thimble tube 88 of the instrumentation nozzle 83 which is not repaired is supported, the groove-welding portion 206 and the in-core instrumentation cylinder 204 are removed under the water, and the processing of the plug attachment opening 211 and the welding of the plug 212 are performed in the atmosphere. FIG. 21 is a schematic diagram illustrating an operation of attaching a plug used in a nozzle repair method according to a second embodiment of the invention. Furthermore, the same reference numerals will be given to the same components having the same functions as the above-described embodiment, and the detailed description thereof will not be presented. The nozzle repair method of the second embodiment is different from that of the first embodiment in that the plug and the pressing load generating jig are different. As illustrated in FIG. 21, the plug attachment opening 211 which is formed in the inner surface of the lower end plate 66 includes the inner peripheral surface 211a and the bottom surface 211b. Meanwhile, a plug 311 which is attached to the plug attachment opening 211 includes a plug body 311a, a vertical wall portion 311b, a concave portion 311c, a protrusion 311d, and a lower surface 311e, and also includes a female screw portion 311f provided in the protrusion 311d. When the plug 311 is fitted into the plug attachment opening 211, the plug body 311a is filled in the plug attachment opening 211, the protrusion 311d is fitted into the attachment hole 203, the annular vertical wall portion 311b protrudes upward from the inner surface of the lower end plate 66, and the concave portion 311c is flush with the inner surface of the lower end plate 66. Here, an operation of applying a pressing load to the plug 311 attached to the plug attachment opening 211 from the outside of the lower end plate 66 is performed. A screw shaft 312 is prepared as the pressing load generating jig to be threaded into the female screw portion 311f of the plug 311. The screw shaft 312 is set to have a length in which the screw shaft may be threaded into the female screw portion 311f of the plug 311. First, a support plate 313 is temporarily fixed to the buttered-welding layer 207 of the lower end plate 66 corresponding to the attachment hole 203 at the outside of the lower end plate 66. Next, the screw shaft 312 is held by a rotation tool (not illustrated), and is caused to penetrate the support plate 313. Then, the front end of the screw shaft 312 is pressed against the female screw portion 311f in a rotation state, the screw shaft 213 is threaded into the female screw portion 311f, and the screw shaft is fixed to the plug 311. Here, when the screw shaft 312 is further threaded into the female screw portion 311f, the plug 311 is pulled toward the downside, that is, the outside of the lower end plate 66 through the screw shaft 312 from the support plate 313 as the original point. Then, a pressing load exerted toward the lower end plate 66 is applied to the plug 311 while the lower surface 311e is pressed against the bottom surface 211b of the plug attachment opening 211. Here, the outer peripheral surface of the vertical wall portion 311b of the plug 311 and the inner surface of the lower end plate 66, that is, the surface of the buttered-welding layer 202 are fillet-welded. The plug 311 is thermally expanded by the welding heat and is contracted in a cooled state. However, since the plug 311 is welded while being compressed and deformed due to a pressing load exerted toward the lower end plate 66, the lower surface 311e is not separated from the bottom surface 211b of the plug attachment opening 211 even when the plug is thermally expanded and contracted in a cooled state, and hence a gap therebetween is suppressed. In this way, in the second embodiment, the plug 311 is provided with the female screw portion 311f, the screw shaft 312, and the support plate 313 as the pressing load generating jig. Accordingly, a pressing load may be easily applied to the plug 311, and hence the workability may be improved. In this case, since the female screw portion 311f is formed in the lower surface 311e of the plug 311 in advance, a pressing load of pressing the plug 311 against the plug attachment opening 211 may be applied to the plug. FIG. 22 is a schematic diagram illustrating an operation of attaching a plug used in a nozzle repair method according to a third embodiment of the invention. Furthermore, the same reference numerals will be given to the same components having the same functions as the above-described embodiment, and the detailed description thereof will not be presented. The nozzle repair method of the third embodiment is different from those of the first and second embodiments in that the plug and the pressing load generating jig are different. As illustrated in FIG. 22, the plug attachment opening 211 which is formed in the inner surface of the lower end plate 66 includes the inner peripheral surface 211a and the bottom surface 211b. Meanwhile, the plug 321 which is attached to the plug attachment opening 211 includes a plug body 321a, a vertical wall portion 321b, a concave portion 321c, a protrusion 321d, and a lower surface 321e. When the plug 321 is fitted into the plug attachment opening 211, the plug body 321a is filled in the plug attachment opening 211, the protrusion 321d is fitted into the attachment hole 203, the annular vertical wall portion 321b protrudes upward from the inner surface of the lower end plate 66, and the concave portion 321c is flush with the inner surface of the lower end plate 66. Here, an operation of applying a pressing load to the plug 321 attached to the plug attachment opening 211 from the outside of the lower end plate 66 is performed. A pressing cylinder 331 is prepared as the pressing load generating jig that presses the plug 321 against the plug attachment opening 211. The pressing cylinder 311 includes a pressing portion 334 which is supported by the support base 322 through a driving rod 333. First, a pressing cylinder 331 is operated so as to lengthen the driving rod 333 so that the pressing portion 334 contacts the concave portion 331c of the plug 321. Next, the driving rod 333 is further lengthened by the pressing cylinder 331 so that the plug 321 is pressed by the pressing portion 334. Then, the lower surface 321e of the plug 321 is pressed against the bottom surface 211b of the plug attachment opening 211, so that a pressing load exerted toward the lower end plate 66 is generated. In this state, the outer peripheral surface of the vertical wall portion 321b of the plug 321 and the inner surface of the lower end plate 66, that is, the surface of the buttered-welding layer 202 are fillet-welded. Subsequently, the pressing cylinder 331 is removed. The plug 321 is thermally expanded by the welding heat and is contracted in a cooled state. However, since the plug 321 is welded while being compressed and deformed by a pressing load exerted toward the lower end plate 66, the lower surface 321e is not separated from the bottom surface 211b of the plug attachment opening 211 even when the plug is thermally expanded and contracted in a cooled state, and hence a gap therebetween is suppressed. In this way, in the third embodiment, the pressing cylinder 331 is prepared as the pressing load generating jig, and the plug 321 is pressed by the pressing cylinder 331 from the inside of the lower end plate 66, so that a pressing load is applied to the plug. Accordingly, there is no need to perform various processing operations for the plug 321. Thus, the structure of the plug 321 may be simplified and the manufacturing cost may be decreased. FIG. 23 is a schematic diagram illustrating an operation of attaching a plug used in a nozzle repair method according to a fourth embodiment of the invention, and FIG. 24 is a bottom view illustrating the plug of the fourth embodiment. Furthermore, the same reference numerals will be given to the same components having the same functions as the above-described embodiment, and the detailed description thereof will not be presented. The nozzle repair method of the fourth embodiment is different from that of the first embodiment in that the plug is different. As illustrated in FIGS. 23 and 24, the plug attachment opening 211 which is formed in the inner surface of the lower end plate 66 includes the inner peripheral surface 211a and the bottom surface 211b. Meanwhile, the plug 341 which is attached to the plug attachment opening 211 includes a plug body 341a, a vertical wall portion 341b, a concave portion 341c, a protrusion 341d, and a lower surface 341e. Further, the plug 341 is provided with a plurality of groove portions 341f which is formed in the radial direction of the lower surface 341e at the same interval in the circumferential direction. The plurality of groove portions 341f serves as a deformation member that is provided between the lower surface 341e of the plug 341 and the bottom surface 211b of the plug attachment opening 211. When the plug 341 is fitted into the plug attachment opening 211, the plug body 341a is filled in the plug attachment opening 211, the protrusion 341d is fitted into the attachment hole 203, the annular vertical wall portion 341b protrudes upward from the inner surface of the lower end plate 66, and the concave portion 341c is flush with the inner surface of the lower end plate 66. At this time, the plug 341 is formed so that the lower surface 341e excluding the groove portions 341f contacts the bottom surface 211b of the plug attachment opening 211. Here, an operation of applying a pressing load to the plug 341 attached to the plug attachment opening 211 from the outside of the lower end plate 66 is performed. That is, when the screw shaft 342 of the plug 341 is pulled downward, the lower surface 341e of the plug 341 is pressed against the bottom surface 211b of the plug attachment opening 211, and hence a pressing load exerted toward the lower end plate 66 is applied thereto. In this state, the outer peripheral surface of the vertical wall portion 341b of the plug 341 and the inner surface of the lower end plate 66, that is, the surface of the buttered-welding layer 202 are fillet-welded. At this time, since the plug 341 is welded while being compressed and deformed by a pressing load exerted toward the lower end plate 66, the lower surface 341e is not separated from the bottom surface 211b of the plug attachment opening 211 even when the plug is thermally expanded by the welding heat and is contracted in a cooled state, and hence a gap therebetween is suppressed. Then, since the plug 341 has a configuration in which the lower surface 341e is provided with the plurality of groove portions 341f, the strength of the lower portion is slightly degraded. Accordingly, the plug may be easily compressed and deformed. In this way, in the fourth embodiment, the lower surface 341e of the plug 341 is provided with the plurality of groove portions 341f. Accordingly, when a pressing load of pressing the plug 341 against the plug attachment opening 211 is applied to the plug, the strength of the lower portion is slightly degraded by the plurality of groove portions 341f, and hence the plug may be easily compressed and deformed. Thus, the plug 341 may be easily compressed and deformed. Furthermore, in the fourth embodiment, the plurality of groove portions 341f is formed in the lower surface 341e of the plug 341 as the deformation member of the invention, but the invention is not limited to this configuration. For example, the concave portion may be proved instead of the groove portion 341f. Further, the concave portion may be provided separately from the plug 341. Further, in the above-described embodiments, the machining operation is performed so that the axis of the plug attachment opening 211 matches the axis of the attachment hole 203, but the machining operation may be performed so that the axis of the plug attachment opening 211 is aligned to the radial direction of the lower end plate 66. Further, the plug attachment opening 211 is formed as a columnar shape, but the invention is not limited to the shape. For example, a prismatic shape, a semi-spherical shape, or a conical shape may be employed. Further, in the above-described embodiments, the vertical wall portions 212b, 311b, 321b, and 341b having the concave portions 212c, 311c, 321c, and 341c at the inside thereof are provided as the protrusion portions protruding upward from the upper surfaces of the plug bodies 212a, 311a, 321a, and 341a, but the concave portions 212c, 311c, 321c, and 341c may not be provided. Further, in the above-described embodiment, a method of repairing the instrumentation nozzle 83 provided in the lower end plate 66 of the nuclear reactor vessel body 62 has been described, but the method may be also used to repair the instrumentation nozzle provided in the upper end plate of the nuclear reactor vessel head 63. Further, a case has been described in which the nozzle repair method of the invention is applied to the pressurized water reactor, but the nozzle repair method may be also applied to a boiling-water nuclear reactor. 61 NUCLEAR REACTOR VESSEL 62 NUCLEAR REACTOR VESSEL BODY 63 NUCLEAR REACTOR VESSEL HEAD 66 LOWER END PLATE (SEMI-SPHERICAL PORTION) 83 INSTRUMENTATION NOZZLE 84 IN-CORE INSTRUMENTATION GUIDE PIPE 85 CONDUIT TUBE 88 THIMBLE PIPE 126 NUT ROTATION TOOL 201 BASE MATERIAL 202 BUTTERED-WELDING LAYER 203 ATTACHMENT HOLE 204 IN-CORE INSTRUMENTATION CYLINDER 205 GROOVING PORTION 206 GROOVE-WELDING PORTION 208 TREPANNING PORTION (CONNECTION PORTION) 210 MACHINED SURFACE 211 PLUG ATTACHMENT OPENING (PLUG ATTACHMENT PORTION) 211b BOTTOM SURFACE (SUPPORT SURFACE) 212, 311, 321, 341 PLUG 212b, 311b, 321b, 341b VERTICAL WALL PORTION (PROTRUSION PORTION) 212c, 311c, 321c, 341c CONCAVE PORTION 212e, 311e, 321e, 341e LOWER SURFACE (SUPPORT SURFACE) 213, 312 SCREW SHAFT (PRESSING LOAD GENERATING JIG) 214, 313 SUPPORT PLATE (PRESSING LOAD GENERATING JIG) 215 NUT (PRESSING LOAD GENERATING JIG) 216 FILLET-WELDING PORTION 311f FEMALE SCREW PORTION 331 PRESSING CYLINDER
	description	This application is a divisional of U.S. patent application Ser. No. 13/046,391, filed on Mar. 11, 2011, which claims priority to U.S. Provisional Patent Application No. 61/355,474, filed Jun. 16, 2010, and to U.S. Provisional Patent Application No. 61/431,341, filed Jan. 10, 2011, the entire contents of all of which are hereby incorporated by reference. The present invention relates to fuel channels for nuclear reactors. More particularly, the invention relates to an annulus spacer for use in the fuel channel of a CANDU-type nuclear reactor. The CANDU (“CANada Deuterium Uranium”) reactor is a heavy water or light water cooled, heavy-water moderated, fission reactor capable of using fuels composed of natural uranium, other low-enrichment uranium, recycled uranium, mixed oxides, fissile and fertile actinides, and combinations thereof. In one embodiment, the invention provides an annulus spacer for a fuel channel assembly of a nuclear reactor, the fuel channel assembly including a calandria tube and a pressure tube positioned at least partially within the calandria tube, the annulus spacer comprising a garter spring configured to surround a portion of the pressure tube to maintain a gap between the calandria tube and the pressure tube, the garter spring including a first end and a second end; a connector coupled to the first end and the second end of the garter spring, the connector allowing movement of the annulus spacer when the pressure tube moves relative to the calandria tube during thermal cycles of the fuel channel assembly; and a girdle wire positioned substantially within the garter spring and configured to form a loop around the pressure tube. In another embodiment, the invention provides an annulus spacer for a fuel channel assembly of a nuclear reactor, the fuel channel assembly including a calandria tube and a pressure tube positioned at least partially within the calandria tube, the annulus spacer comprising a garter spring configured to surround a portion of the pressure tube to maintain a gap between the calandria tube and the pressure tube; and a girdle wire positioned substantially within the garter spring and configured to form a loop around the pressure tube, the girdle wire including a first segment and a second segment that overlaps the first segment to form an overlap, the overlap extending between approximately 45 degrees and approximately 135 degrees to reduce the possibility of girdle wire twisting. In yet another embodiment, the invention provides a method of manufacturing an annulus spacer for a fuel channel assembly of a nuclear reactor, the method comprising providing a garter spring wire having a trapezoidal cross-section, a first end, and a second end; bending the garter spring wire into a coil to induce compressive strain on an inner portion of the trapezoidal cross-section and tensile strain on an outer portion of the trapezoidal cross-section such that the garter spring wire obtains a rectangular cross-section; positioning a girdle wire within the coil formed by the garter spring wire; and coupling the first end of the garter spring wire to the second end of the garter spring wire to form a toroid. Other aspects of the invention will become apparent by consideration of the detailed description and accompanying drawings. Before any embodiments of the invention are explained in detail, it is to be understood that the invention is not limited in its application to the details of construction and the arrangement of components set forth in the following description or illustrated in the following drawings. The invention is capable of other embodiments and of being practiced or of being carried out in various ways. FIG. 1 is a perspective of a reactor core of a CANDU-type reactor 6. A generally cylindrical vessel, known as a calandria 10, contains a heavy-water moderator. The calandria 10 has an annular shell 14 and a tube sheet 18 at a first end 22 and second end 24. A number of fuel channel assemblies 28 pass through the calandria 10 from the first end 22 to the second end 24. As illustrated in FIG. 2, each fuel channel assembly 28 is surrounded by a calandria tube (CT) 32. The CT 32 forms a first boundary between the heavy water moderator of the calandria 10 and the fuel channels assemblies 28. A pressure tube (PT) 36 forms an inner wall of the fuel channel assembly 28. The PT 36 provides a conduit for reactor coolant and fuel assemblies 40. An annulus space 44 is defined by a gap between the PT 36 and the CT 32. The annulus space 44 is normally filled with a circulating gas, such as dry carbon dioxide, nitrogen, air or mixtures thereof. The annulus space 44 and gas are part of an annulus gas system. The annulus gas system has two primary functions. First, a gas boundary between the CT 32 and PT 36 provides thermal insulation between hot reactor coolant and fuel within the PTs 36 and the relatively cool CTs 32. Second, the annulus gas system provides indication of a leaking calandria tubes, pressure tubes 36, or their connections via the presence of moisture in the annulus gas. An annulus spacer 48 is disposed between the CT 32 and PT 36. Functionally, the annulus spacer 48 serves roles in ensuring the safe, long-term operation of CANDU-type nuclear reactors. The annulus spacer 48 maintains the gap between the PT 36 and the corresponding CT 32, while allowing the passage of the annulus gas through and around the annulus spacer 48. More specifically, the annulus spacer 48 substantially minimizes the risk of contact between the CT 32 and PT 36 under Design Level A and B service conditions and Level C transients for the design life of the fuel channel, with the exception of a design basis earthquake with a fueling machine attached. The PT 36 would be inspected after such an earthquake, and if significant permanent deformation or damage has taken place, shall be replaced. The annulus spacer 48 limits heat transfer from the PT 36 to the heavy-water moderator during normal operating conditions, thus increasing the thermal efficiency of the reactor, and ensuring that hot PTs 36 are not locally cooled. Thermal gradients in the wall of a PT 36 can permit hydrogen (deuterium) diffusion along the gradient above threshold hydrogen concentrations. High hydrogen concentrations may allow hydride accumulation and the potential for unstable cracking during the PT 36 design life. Other functions of the annulus spacer 48 include accommodating relative axial movement between the PT 36 and CT 32 while limiting wear/scratches/deformation/damage to the PTs 36 and CTs 32, so that integrity and performance are maintained throughout the design life of the fuel channel. The annulus spacers 48 are configured to withstand the annulus gas environmental conditions without substantial degradation for the design life of the fuel channel. The annulus spacer 48 is further configured to limit parasitic neutron absorption and thereby reduce the fuel burn-up penalty by careful selection of spacer dimensions, spring cross-section, geometry, connections, and material. The performance requirements of the annulus spacer 48 are primarily based upon the functional requirements. In some embodiments, the annulus spacer 48 may withstand the maximum predicted PT 36 to CT 32 vertical interaction load specified in the applicable Fuel Channel Design Specification, without impeding the functional requirements of the spacer design or causing unacceptable deformation to the PT 36 or the CT 32. In some embodiments, the cross-section of the annulus spacer 48 is optimized as a square shape in order to maximize the load bearing capability in bending while minimizing the amount of material used. The annulus spacer 48 may also withstand PT to CT movement caused by the predicted number of thermal cycles and PT axial elongation specified in an applicable Fuel Channel Design Specification without impeding the functional requirements of the annulus spacer design or causing unacceptable deformation/wear to the PT 36 or the CT 32. Additionally, the annulus spacer 48 may withstand a maximum predicted diametral increase of the PT 36 specified in an applicable Fuel Channel Design Specification without nip-up. Nip-up occurs when the limit of unconstrained diametral expansion of the PT 36 at the location of the annulus spacer 48 has been reached. The annulus spacers 48 may also remain in their design location so as to prevent PT 36 to CT 32 contact throughout the life of the fuel channel. In some embodiments, annulus spacer 48 axial positions may be verifiable during fuel channel inspections throughout the life of the reactor, so as to ensure that PT to CT contact will not occur before the end of the next inspection interval. From a safety perspective, in some embodiments, the annulus spacer 48 may not result in unacceptable consequences that may affect reactor safeguards analysis. If required by the safety analysis, the annulus spacer 48 may allow contact of a PT 36 with the CT 32 surrounding it over a large enough area to permit a sufficient dissipation of heat for preventing fuel channel failure under a postulated event initiated by an accident condition such as a loss of coolant accident (LOCA). In the illustrated embodiment, the annulus spacer material does not interact with the PT material at high temperature during transients so as to compromise the integrity of the PT 36. In some embodiments, the annulus spacers 48 may also not cause local stresses in the PT 36 that could initiate premature PT failure. During a severe fuel channel flow blockage event, the annulus spacer 48 may not significantly increase the amount of molten material that might be present in the affected channel. The annulus spacers 48 may allow relatively unimpeded annulus gas flow for leak before break detection purposes. In some embodiments, the annulus spacer 48 withstands the fuel channel environmental conditions throughout its design life. As the annulus spacer 48 is located in the fuel channel annulus space 44, its temperature can be influenced by either the hot PT 36 (approximately 300° C.) or the cooler CT 32 (approximately 80° C.), depending on which component it is contacting. If the annulus spacer 48 is in contact with both tubes it will experience a temperature gradient between the temperatures of the PT 36 and the CT 32. The temperature of the annulus spacer 48 is further influenced by the contribution of gamma heating, although this effect should be small and is dependent on the spacer material selected. Nevertheless, the impact of gamma heating on the environmental conditions may be assessed once a spacer material and design are selected. The environment within the fuel channel annulus space 44 is primarily circulating carbon dioxide maintained at a low dew point, containing a small addition of oxygen. The annulus space 44 also experiences a relatively high fast neutron and thermal neutron flux. As there is a slow increase in the dew point of the annulus gas over time, the annulus gas system is periodically purged to maintain the sensitivity of its leak detection function. In the case of an abnormal operating occurrence or a Design Basis Accident (such as a fuel channel leak or pressure tube rupture), fuel channels other than the source channel may be exposed to extended periods of low temperature, moist annulus conditions. The annulus spacer 48 can be manufactured from materials that are stable under irradiation and are capable of withstanding the environmental conditions detailed above, such that any change in mechanical properties or geometries will not affect its integrity or location. The annulus spacer 48 directly interfaces with the PT 36, the CT 32, and the gas of the annulus gas system. Thus, the annulus spacers 48 should be compatible with these components. The annulus spacers 48 may permit circulation of the annulus gas (comprised of CO2 and small additions of O2) along the fuel channel annulus 44. The annulus spacers 48 can allow the fuel channel annulus to be efficiently dried if water leaks into it. The annulus spacers 48 do not reduce the design life or affect the integrity of the PT 36 or the CT 32 under all normal operating conditions. The annulus spacers 48 do not significantly interfere with the axial expansion of the PT 36, so as to affect the relative axial loading of the PT 36 or the CT 32. Annulus spacers 48 do not cause the formation of a stress riser on either the PT 36 or CT 32 greater than the maximum allowable value determined by the lower bound value for K1H. This is needed to demonstrate there is an adequate margin against delayed hydride cracking (DHC) initiation in either tube by the design loading conditions. FIG. 3 is a more-detailed perspective of an annulus spacer 48 installed within the annulus space 44 between the PT 36 and CT 32. The annulus spacer 48 includes a garter spring 52 and a girdle wire 56. The garter spring 52 is formed from a length of coiled wire 60. Two ends 64 and 68 of the coiled wire 60 are connected so that the garter spring 52 forms a toroid 72. The garter spring 52 is dimensioned to fit tightly around the PT 36. The garter spring 52 is resilient so that it may be expanded to a dimension greater than an outside diameter 76 of the PT 32 during installation, yet fit tightly and securely once positioned. In the illustrated embodiment, the garter spring 52 is formed from a nickel-chromium based alloy such as INCONEL X-750. In other embodiments, the garter spring 52 may be formed of other alloys, including zirconium-based alloy such as ZIRCALOY or a zirconium-niobium-copper alloy. In still other embodiments, the garter spring 52 may be formed of an alloy including, but not limited to, a combination of zirconium, niobium, and copper. The girdle wire 56 is held within an annular cavity 80 formed by the coiled wire 60 of the garter spring 52. The girdle wire 56 has two functions. First, the girdle wire 56 provides a fail-safe in the event that the garter spring 52 breaks. The girdle wire 56 will capture the separated garter spring 52. Second, in some embodiments the girdle wire 56 improves the ability to detect a position of the annulus spacer 48 using eddy current testing (ECT) techniques. Detecting the position of the annulus spacer 48 is necessary in order to verify the location of the annulus spacer 48 in order to ensure that the annulus spacer 48 meets a variety of functional, performance, safety, environmental and inter-facing system requirements. The girdle wire 56 helps the annulus spacer 48 be detectable by providing a loop of continuous conductivity. In the illustrated embodiment, the girdle wire 56 is formed of a zirconium-based alloy such as ZIRCALOY. In other embodiments, the girdle wire 48 can be formed from a variety of other alloys. In the illustrated embodiment, an outer segment 84 of the girdle wire 56 overlaps upon an inner segment 88 of the girdle wire 56. An overlap 92 is provided in order to ensure that the girdle wire 56 forms a continuous loop or overlapping loop within the garter spring 52, in order to ensure that the garter spring 52 is captured in the event of failure. However, overlapping portions of the girdle wire 56 may oxidize or move relative to each other over time and during operation of the reactor, degrading the conductivity, and thereby detectability, of the annulus spacer 48. As such, annulus spacers including girdle wires with overlaps may include garter springs that are welded into a continuous loop (as shown in FIGS. 8A-8D) to facilitate detection. In other embodiments, ends of the girdle wire 56 may be welded together with substantially no overlap. The conductivity, and thereby detectability, of girdle wires with welded ends typically does not degrade over time or during operation of the reactor. Various types of connectors to connect ends of garter springs (such as those shown in FIGS. 9A-15C) may therefore be employed in annulus spacers having welded girdle wires. In some embodiments, ends of the girdle wire 56 may be both overlapped and welded together. Additionally or alternatively, in some embodiments, both the girdle wire 56 and the garter spring 52 may include ends that are welded together. A girdle wire 56 according to the present invention alleviates a problem known in the nuclear industry as garter spring hang-up. During installation over the PT 36, an overlap 92 of 180 degrees or more could allow the girdle wire 56 to twist upon itself. A twisted girdle wire 56 may, in turn, prevent the garter spring 52 from compressing about the PT 36 when installed. If the garter spring 52 is blocked from compressing due to a twisted girdle wire 56, the functional and performance requirements of the annulus spacer 48 may not be met. In particular, a hung-up garter spring 52 may result in the annulus spacer 48 shifting between inspections and overhauls to the point where PT 36 to CT 32 contact could occur. A hang-up may also result in an annulus spacer installation tool jamming or failing. The annulus spacer 48 illustrated in FIG. 3 includes features designed to prevent girdle wire twisting. First, the overlap 92 of the girdle wire 56 has been dimensioned in order to minimize the risk of girdle wire twisting. In other words, it is desired that the overlap 92 be long enough to ensure that at least 360 degrees of continuous girdle wire 56 is provided within the coils 60 of the garter spring 52. However, the overlap 92 of the girdle wire 56 of a given configuration should be short enough to ensure that the girdle wire 56 is unlikely to twist upon itself In other words, it is desired that the overlap 92 of the girdle wire 56 remains substantially co-planar with non-overlapping portion of the girdle wire 56 during installation and operation. The amount of overlap 92 that meets these requirements will depend on the geometry, dimensions and mechanical characteristics of the wire from which the girdle wire 56 is formed. In the illustrated embodiment, it has been found that girdle wire overlaps of between approximately 1 degree and approximately 179 degrees, preferably between approximately 45 and 135 degrees, and even more preferably 75 to 105 degrees minimize the risk of the girdle wire twisting upon itself It should be recognized, however, that these ranges of overlap are based upon a girdle wire of the configuration illustrated. FIG. 4 illustrates a second feature of the girdle wire 56 intended to reduce the risk of girdle twisting. Specifically, the girdle wire 56 is formed with a nominal diameter 96 substantially smaller than the diameter 76 of the pressure tube 36. In one example, the girdle wire 56 is pre-formed with an approximately 3.5 inch diameter. A pressure tube 36 of the configuration illustrated has a diameter 76 of approximately 4.4 inches. When installed within the garter spring 52, a girdle wire 56 of this dimension will exhibit a desirable tendency to compress around the outside diameter 76 of the pressure tube 36. This compression provides additional retention for the annulus spacer 48 position, as well as further minimizing the risk of the girdle wire 56 twisting upon itself. FIG. 5 illustrates another embodiment of a girdle wire 556. In this embodiment, the girdle wire 556 is preformed with a plurality of kinks 558. The kinks 558 are intended to minimize the ability of the girdle wire 556 to twist upon itself The kinks 558 obstruct relative motion between an inner segment 588 and an outer segment 584. FIG. 6 illustrates a cross-section of yet another embodiment of a girdle wire 656. The cross-section is of an overlap 692 of the girdle wire 656. In this embodiment, the girdle wire 656 has a u-shaped cross-section. Because of the u-shaped cross section, an outer segment 684 of the overlap 692 nests, or mates, with an inner segment 688. The nesting minimizes lateral relative movement between the inner segment 688 and the outer segment 684, thus minimizing the risk of the girdle wire 656 twisting upon itself. FIG. 7 illustrates a cross-section of yet another embodiment of a girdle wire 756. The cross-section is of an overlap 792 of the girdle wire 756. In this embodiment the girdle wire 756 has a deep v-shaped cross-section. Because of the v-shaped cross section, an outer segment 784 of the overlap 792 nests, or mates, with an inner segment 788. Like the embodiment of FIG. 6, the nesting minimizes lateral relative movement between the inner segment 788 and the outer segment 784, thus minimizing the risk of the girdle wire 756 twisting upon itself. It should be appreciated that the features identified in FIGS. 3, 4, 5, 6, and 7 may be combined in various embodiments of the invention. Thus, for example, an overlap 92 having the dimensions of FIG. 3 may be combined with the kinks 558 of FIG. 5, in a girdle wire have a non-circular cross-section such as that disclosed in FIG. 6 or 7. Furthermore, the invention may include any other combination of the concepts disclosed herein. FIGS. 8A and 8B illustrate another embodiment of an annulus spacer 800 in more detail. Similar to the annulus spacer 48 discussed above, the annulus spacer 800 includes a garter spring 804 and a girdle wire 808. In the illustrated embodiment, the garter spring 804 includes two end portions 812, 816 that are turned into each other and welded together so that the garter spring 804 forms a toroid. In some embodiments, the two end portions 812, 816 may be welded at a single location by, for example, a single spot weld 820 (FIG. 8C). In other embodiments, the two end portions 812, 816 may be welded at locations spaced approximately 180 degrees from each other around the coil circumference by, for example, two spot welds 824 (FIG. 8D). Such embodiments increase the strength and redundancy of the connection while maintaining the flexibility required to enable free rolling of the annulus spacer 800 between the pressure tube and the calandria tube. In still other embodiments, the end portions 812, 816 may be welded at a plurality of locations by, for example, a plurality of spot welds spaced around the circumference of the garter spring 804. In other embodiments, such as the embodiment shown in FIG. 8A, the end portions 812, 816 may be welded in a continuous strip partially or completely around the circumference of the garter spring 804. In any embodiment, the end portions 812, 816 may be welded together by, for example, laser welding or electron beam welding. In some embodiments, such as the illustrated embodiment, less than about four overlaps or interlocks of the end portions 812, 816 are interlocked and welded together to maintain flexibility of the annulus spacer 800. In some embodiments, the cross-section of the garter spring 804 at the two end portions 812, 816 is optimized to have an approximately square or rectangular shape. This ideal cross-section is created by manufacturing the garter spring 804 from a straight or slightly curved wire that has an optimized trapezoidal cross-section. The wire is wound or bent via the manufacturing process into a coil, inducing compressive strain on the inner portion, or intrados, of the cross-section and tensile strain on the outer portion, or extrados, of the cross-section. The resultant cross-sectional shape is approximately square or rectangular and is optimized in order to maximize the strength of the garter spring 804 in bending, while minimizing the amount of material required to carry the necessary loads. By minimizing the amount of material required to carry a load, the parasitic effect of the garter spring 804 due to neutron absorption, attenuation, and reflection is lessened, which in turn directly leads to higher uranium utilization and efficiency of the reactor core. The garter spring cross-section has also been optimized in order to maximize the surface area in contact between the inter-wound end portions 812, 816 of the garter spring 804. This feature increases the strength and quality of the weld or plurality of welds. This feature also increases the ease and repeatability of creating the weld or plurality of welds that meet stringent nuclear industry standards, resulting in a lower cost per unit. Although not further discussed, this manufacturing technique may be applied to any garter spring disclosed herein. FIGS. 9A and 9B illustrate a third embodiment of an annulus spacer 900. In the illustrated embodiment, the annulus spacer 900 includes a garter spring 904, a girdle wire 908, and a universal joint 912. The universal joint 912 includes a first hinge 916 and a second hinge 920, each having a groove 924, 928 to receive portions of the garter spring 904. A first end portion 932 of the garter spring 904 is positioned within the groove 924 of the first hinge 916 to connect the first end portion 932 to the universal joint 912. A second end portion 936 of the garter spring 904 is positioned within the groove 928 of the second hinge 920 to connect the second end portion 936 to the universal joint 912. In some embodiments, the end portions 932, 936 of the garter spring 904 may be, for example, press-fit or welded within the grooves 924, 928 to secure the end portions 932, 936 to the universal joint 912. In some embodiments (see, for example, FIG. 9C), a groove, notch, or hole 940 may be formed in each hinge 916, 920 so that each end of the garter spring 904 can be bent, punched, or generally deformed to sit in the groove, notch, or hole 940 to help inhibit the garter spring 904 from becoming unraveled from the universal joint 912. The hinge portions 916, 920 of the universal joint 912 are coupled together by a cross-shaft 940. The cross-shaft 940 allows the universal joint 914 to pivot or bend in any direction, thereby maintaining the flexibility of the annulus spacer 900. As shown in FIG. 9B, the cross-shaft 940 defines a channel 944 through which the girdle wire 908 passes such that the girdle wire 908 does not interfere with movement of the universal joint 914. FIGS. 10A and 10B illustrate a fourth embodiment of an annulus spacer 1000. In the illustrated embodiment, the annulus spacer 1000 includes a garter spring 1004, a girdle wire 1008, and a sleeve joint 1012. As shown in FIG. 10C, the sleeve joint 1012 includes a tubular member 1016 and a threaded shoulder 1020, 1024 positioned at each end of the tubular member 1016. In some embodiments, such as the embodiment shown in FIG. 10D, the threaded shoulders 1020, 1024 may be helical partially or totally around the circumference of the tubular member 1016. Referring back to FIGS. 10A-10C, a first end portion 1028 of the garter spring 1004 surrounds a portion of the tubular member 1016 and is captured by the first shoulder 1020 and a second end portion 1032 of the garter spring 1004. The first end portion 1028 is stopped from rotation by a face 1026 formed on the second shoulder 1024. A second end portion 1032 of the garter spring 1004 surrounds another portion of the tubular member 1016 and is captured by the second shoulder 1024 and the first end portion 1028 of the garter spring 1004. The second end portion 1032 is stopped from rotation by a face 1022 formed on the first shoulder 1020. The threaded shoulders 1020, 1024, or flanges, inhibit the end portions 1028, 1032 of the garter spring 1004 from sliding axially off of the tubular member 1016. This arrangement of the first and second spring ends 1028 1032 with the tubular member 1016 requires that the first and second spring ends 1028, 1032 are axially overlapped with each other such that they are interlocked. In some embodiments, the end portions 1028, 1032 may also be secured to the tubular member 1016 by, for example, crimping, welding, press fitting, or fasteners. The illustrated tubular member 1016 is short enough to not inhibit the spacer 1000 from rolling freely under operating conditions. The end portions 1028, 1032 have predetermined clock positions relative to each other when the spacer 1000 is unassembled (e.g., in straight form) such that a minimum torsion is required in the garter spring 1004 to keep the tubular member 1016 secure and prevent it from becoming loose relative to the end portions 1028, 1032. In some embodiments (see, for example, FIG. 9C), grooves, notches, or holes may be formed in the tubular member 1016 so that each end of the garter spring 1004 can be bent, punched, or generally deformed to sit in a corresponding groove, notch, or hole to help inhibit the garter spring 1004 from becoming unraveled from the sleeve joint 1012. FIGS. 11A and 11B illustrate a fifth embodiment of an annulus spacer 1100. In the illustrated embodiment, the annulus spacer 1100 includes a garter spring 1104, a girdle wire 1108, and a ball joint 1112. The ball joint 1112 includes a female portion 1116 and a male portion 1120, each having a groove 1124, 1128 to receive portions of the garter spring 1104. A first end portion 1132 of the garter spring 1104 is positioned within the groove 1124 of the female portion 1116 to connect the first end portion 1132 to the ball joint 1112. A second end portion 1136 of the garter spring 1104 is positioned within the groove 1128 of the male portion 1120 to connect the second end portion 1136 to the ball joint 1112. In some embodiments, the end portions 1132, 1136 of the garter spring 1104 may be, for example, press-fit or welded within the grooves 1124, 1128 to secure the end portions 1132, 1136 to the ball joint 1112. In some embodiments (see, for example, FIG. 9C), a groove, notch, or hole may be formed in each joint portion 1116, 1120 so that each end of the garter spring 1104 can be bent, punched, or generally deformed to sit in the groove, notch, or hole to help inhibit the garter spring 1104 from becoming unraveled from the ball joint 1112. The male portion 1120 of the ball joint 1112 includes a spherical boss 1140 that is inserted into the female portion 1116 to couple the portions 1116, 1120 together. The female portion 1116 defines slots 1144 that allows the female portion 1116 to slightly deflect to facilitate assembly of the ball joint 1112. When assembled, the spherical boss 1140 allows the ball joint 1112 to pivot or bend in any direction, thereby maintaining the flexibility of the annulus spacer 1100. FIGS. 12A and 12B illustrate a sixth embodiment of an annulus spacer 1200. In the illustrated embodiment, the annulus spacer 1200 includes a garter spring 1204, a girdle wire 1208, and a bellows joint 1212. The bellows joint 1212 includes a generally flexible bellows 1216 having a first flange or shoulder 1220 and a second flange or shoulder 1224. A first end portion 1228 of the garter spring 1204 surrounds a portion of the bellows 1216 and is captured by the first flange 1220. A second end portion 1232 of the garter spring 1204 surrounds another portion of the bellows 1216 and is captured by the second flange 1224. The flanges 1220, 1224 inhibit the end portions 1228, 1232 of the garter spring 1204 from sliding axially off of the bellows 1216. In some embodiments, the end portions 1228, 1232 may also be secured to the bellows 1216 by, for example, adhesives, welding, or fasteners. The bellows 1216 bends in any direction to maintain the flexibility of the annulus spacer 1200. In some embodiments (see, for example, FIG. 9C), grooves, notches, or holes may be formed in the bellows 1216 so that each end of the garter spring 1204 can be bent, punched, or generally deformed to sit in the corresponding groove, notch, or hole to help inhibit the garter spring 1204 from becoming unraveled from the bellows joint 1212. FIGS. 13A and 13B illustrate a seventh embodiment of an annulus spacer 1300. In the illustrated embodiment, the annulus spacer 1300 includes a garter spring 1304, a girdle wire 1308, and a flexible conduit joint 1312. The flexible conduit joint 1312 includes a first connector 1316 having a flange or shoulder 1320, a second connector 1324 having a flange or shoulder 1328, and a flexible conduit 1332 extending between the first and second connectors 1316, 1324. In some embodiments, the flexible conduit 1332 is formed from, for example, a flexible metal conduit or a braided wire hose. Portions of the flexible conduit 1332 are inserted into grooves 1336, 1340 in the connectors 1316, 1324 to couple the conduit 1332 to the connectors 1316, 1324. In some embodiments, the flexible conduit 1332 may be secured within the grooves 1336, 1340 by press-fitting, welding, adhesives, or the like. A first end portion 1344 of the garter spring 1304 surrounds a portion of the first connector 1316 and is captured by the flange 1320. A second end portion 1348 of the garter spring 1304 surrounds a portion of the second connector 1324 and is captured by the flange 1328. The flanges 1320, 1328 inhibit the end portions 1344, 1348 of the garter spring 1304 from sliding axially off of the connectors 1316, 1324. In some embodiments, the end portions 1344, 1348 may also be secured to the connectors 1316, 1324 by, for example, adhesives, welding, or fasteners. In some embodiments (see, for example, FIG. 9C), a groove, notch, or hole may be formed in each connector 1316, 1324 so that each end of the garter spring 1304 can be bent, punched, or generally deformed to sit in the groove, notch, or hole to help inhibit the garter spring 1304 from becoming unraveled from the flexible conduit joint 1312. When assembled, the flexible conduit 1332 bends in any direction to maintain the flexibility of the annulus spacer 1300. FIGS. 14A-14C illustrate an eighth embodiment of a connector 1400 for use with an annulus spacer. In the illustrated embodiment, the connector 1400 is a multiple component connector including a first connector 1404 and a second connector 1408. The first connector includes a flange 1412, or shoulder, and a male connector extension 1416 that mates with the second connector 1408. The second connector 1408 includes a flange 1420, or shoulder, and a locking female penetration 1424 that mates with the first connector 1404. A first end portion of the garter spring is positioned within a groove 1432 in the first connector 1404 and is captured by the flange 1412. A second end portion of the garter spring is positioned within a groove 1436 in the second connector 1408 and is captured by the flange 1420. The flanges 1412, 1420 inhibit the end portions of the garter spring from sliding axially off of the connectors 1404, 1408. In some embodiments, the end portions may also be secured within the grooves 1432, 1436 by press-fitting, welding, adhesives, or the like. In some embodiments (see, for example, FIG. 9C), a groove, notch, or hole may be formed in each connector 1404, 1408 so that each end of the garter spring can be bent, punched, or generally deformed to sit in the groove, notch, or hole to help inhibit the garter spring from becoming unraveled from the multiple component connector 1400. After each end of the garter spring is coupled to the corresponding connector 1404, 1408, the male connector extension 1416 is inserted into the locking female penetration 1424. The connectors 1404, 1408 are then rotated relative to each other approximately 90 degrees such that projections 1440 on the male connector extension 1416 slide into recesses 1444 in the locking female penetration 1424 to lock the connectors 1404, 1408 together. FIGS. 15A-15C illustrate a ninth embodiment of an annulus spacer 1500. In the illustrated embodiment, the annulus spacer 1500 includes a garter spring 1504, a girdle wire 1508, and a connecting joint 1512. The connecting joint 1512 includes a male portion 1516 and a female portion 1520, each having a groove 1524, 1528 to receive portions of the garter spring 1504. A first end portion 1532 of the garter spring 1504 is positioned within the groove 1524 of the male portion 1516 to connect the first end portion 1532 to the connecting joint 1512. A second end portion 1536 of the garter spring 1504 is positioned within the groove 1524 of the female portion 1520 to connect the second end portion 1536 to the connecting join 1512. In some embodiments, the end portions 1532, 1536 of the garter spring 1504 may be, for example, press-fit or welded within the grooves 1524, 1528 to secure the end portions 1532, 1536 to the connecting joint 1512. In some embodiments (see, for example, FIG. 9C), a groove, notch, or hole may be formed in each joint portion 1516, 1520 so that each end of the garter spring 1504 can be bent, punched, or generally deformed to sit in the groove, notch, or hole to help inhibit the garter spring 1504 from becoming unraveled from the connecting joint 1512. The male portion 1516 of the connecting joint 1512 includes a cylindrical boss 1540 that is inserted into the female portion 1520 to couple the portions 1516, 1520 together. The female portion 1520 defines slots 1544 that lock the male portion 1516 and the female portion 1520 together during assembly. When assembled, the cylindrical boss 1540 allows the male and female portions 1516, 1520 to rotate relative to each other, thereby maintaining the flexibility of the annulus spacer 1500. Thus, the invention provides, among other things, an annulus spacer for the fuel channel of a nuclear reactor. Although the invention has been described in detail with reference to certain preferred embodiments, variations and modifications exist within the scope and spirit of one or more independent aspects of the invention. In addition, annulus spacers including any variations and/or combinations of garter springs and girdle wires disclosed herein are also within the scope of the invention. Various features and advantages of the invention are set forth in the following claims.
062597582	summary	BACKGROUND OF THE INVENTION The present invention is directed to hydrogen peroxide decomposer for use in water-cooled nuclear reactors, including boiling water reactors and pressurized water reactors, for the mitigation of corrosion phenomena in such systems. Steel pressure vessels and piping exposed to high temperature water are prone to corrosion due to oxidation of the various metals therein by oxidizing agents, particularly oxygen, present in the high temperature water. Corrosion of such vessels and piping can lead to a variety of problems, including stress corrosion cracking, crevice corrosion and erosion corrosion, leading to leakage and/or bursting of such vessels and piping. In nuclear reactors, significant amounts of heat energy is generated by reactor processes occurring in the reactor core. A liquid coolant, typically water, is used to remove heat from the reactor core and facilitate its conversion to a useable form. A reactor vessel is provided to contain the reactor coolant around the reactor core to effect such heat removal. Further, piping is provided to facilitate transport of the coolant to steam generators or turbines, where heat energy is ultimately converted to electricity. The materials used in the construction of nuclear reactor vessels and piping are elected for their ability to withstand rigorous loading, environmental and radiation conditions. Such materials include carbon steel, low alloy steel, stainless steel and nickel-based, cobalt-based and zirconium-based alloys. Despite careful material selection, corrosion and, particularly, intergranular stress corrosion cracking (or, simply, stress corrosion cracking (SCC)), is a problem in steel pressure vessels and piping used in nuclear reactors. SCC, as used herein, refers to cracking propagated by static or dynamic tensile stressing in combination with corrosion at the crack tip. Unfortunately, the nuclear reactor environment is conducive to both tensile stressing and corrosion. Nuclear reactor pressure vessels and piping are subject to a variety of stresses. Some are attributable to the high operating pressure required to maintain high temperature water in a liquid state. Stresses also arise due to differences in thermal expansion of the materials of construction. Other sources include residual stresses from welding, cold working, and other metal treatments. Nuclear reactors are also susceptible to SCC because of the water chemistry environment of its process systems, which is favourably disposed to corrosion. In this respect, the presence of oxidizing agents, such as oxygen, hydrogen peroxide, and various short-lived radicals, which arise from the radiolytic decomposition of high temperature water in boiling water reactors, contribute to SCC. Hydrogen peroxide is particularly unstable as it has the ability to act as both an oxidizing agent and a reducing agent. Hydrogen peroxide can act as an oxidizing agent, leading to the formation of water according to the following reaction: EQU H.sub.2 O.sub.2 +2H.sup.+ +2e.sup.-.fwdarw.2H.sub.2 O As a reducing agent, hydrogen peroxide is oxidized to oxygen according to the following reaction: EQU H.sub.2 O.sub.2.fwdarw.O.sub.2 +2H.sup.+ +2e.sup.- Because of its ability to act as both an oxidizing agent and a reducing agent, hydrogen peroxide is highly unstable and will spontaneously decompose into water and oxygen according to the following reaction: EQU 2H.sub.2 O.sub.2.fwdarw.2H.sub.2 O+O.sub.2 This will happen if aqueous hydrogen peroxide contacts a metallic surface whose electrode potential lies within this region of instability, which is typically the case in the BWR environment. Stress corrosion cracking is of great concern in boiling water reactors (BWR's) which utilize light water as a means of cooling nuclear reactor cores and extracting heat energy produced by such reactor cores. Stress corrosion cracking causes leakage or bursting of such vessels or piping resulting in the loss of coolant in the reactor core. This compromises the reactor process control, which could have dire consequences. To mitigate stress corrosion cracking phenomenon in BWR's, it is desirable to reduce the electrochemical corrosion of metal components that are exposed to aqueous fluids. ECP Electrochemical Corrosion Potential is a measure of the thermodynamic tendency for corrosion to occur, and is a fundamental parameter in determining rates of stress corrosion cracking. ECP has been clearly shown to be a primary variable in controlling the susceptibility of metal components to stress corrosion cracking in BWRs. FIG. 1 shows the observed and predicted crack growth rate as a function of ECP for furnace sensitized Type 304 stainless steel at 27.5 MPa in 288.degree. C. water over the range of solution conductivities from 0.1 to 0.3 .mu.S/cm. For type 304 stainless steel (containing 18-20% Cr, 8-10.5% Ni, and 2% Mn), it is known that if the ECP of such steel exposed to high temperature water at about 288.degree. C. can be reduced to values below -230 mV (Standard Hydrogen Electrode--SHE) (hereinafter the "critical corrosion potential"), the stress corrosion cracking problem of such steel can be greatly reduced. The same generally applies for other types of steels. A well-known method to reduce the ECP to less than -230 mV.sub.SHE and thereby mitigate SCC of steel pressure vessels and piping in nuclear reactors, is to inject hydrogen gas to the recirculating reactor feedwater. The injected hydrogen gas reduces oxidizing species in the water, such as dissolved oxygen. This has the very desirable benefit of reducing the corrosion potential of the steel vessel or piping carrying such high temperature water. As illustrated in FIG. 2, ECP of 304SS in 288.degree. C. water increases more rapidly with continued addition of hydrogen peroxide when compared to the ECP values measured at the same levels of oxygen concentration. Further, even with the use of hydrogen gas injection, SCC in BWRs continues to occur at unacceptable rates when hydrogen peroxide is present. This is illustrated in FIG. 3, where stress corrosion cracking is shown to occur in BWRs, even with the addition of hydrogen gas, when 20-30 ppb of hydrogen peroxide is present. This information suggests that the presence of hydrogen peroxide in reactor systems is a significant contributor to stress corrosion cracking of metal components. Moreover, the present practice of injecting hydrogen gas into the process liquid does not appear to completely assist in the decomposition of hydrogen peroxide and therefore does not bring about the concomitant reduction in ECP that is expected. SUMMARY OF INVENTION In one broad aspect, the present invention provides a corrosion resistant alloy having a surface exposed to aqueous liquid consisting of oxidizing species, including hydrogen peroxide, that increase the ECP of the alloy. The surface of the alloy is coated with a coating comprised of Mn, Mo, Zn, Cu, Cd, oxides thereof, or chemical compounds thereof. These metals and their compounds assist in causing the decomposition of hydrogen peroxide, thereby reducing the ECP of the alloy. These metals and their compounds can be present as a pre-existing coating on the alloy, or may be deposited in-situ into the aqueous liquid for subsequent deposition on the surface of the alloy after injection. According to another broad aspect of the present invention there is provided a corrosion resistant alloy cooling tube in a water-cooled nuclear reactor having a surface exposed to an aqueous cooling medium containing hydrogen peroxide, the surface being coated with a coating comprising matter selected from the group consisting of manganese, molybdenum, zinc, copper, cadmium, oxides thereof, chemical compounds thereof and mixtures thereof, for causing decomposition of the hydrogen peroxide. According to another aspect of the present invention there is provided a water-cooled nuclear reactor comprising metal piping, such metal piping having a surface exposed to an aqueous liquid containing hydrogen peroxide, the surface being coated with a coating comprising matter selected from the group consisting of manganese, molybdenum, zinc, copper, cadmium, oxides thereof, chemical compounds thereof and mixtures thereof, for causing decomposition of the hydrogen peroxide. According to another aspect of the present invention there is provided a method for lowering the electrochemical corrosion potential of a metal alloy, for use in a cooling tube in a water-cooled nuclear reactor, having a surface exposed to an aqueous liquid containing hydrogen peroxide, comprising the step of coating the surface with matter selected from the group consisting manganese, molybdenum, zinc, copper, cadmium, oxides thereof, chemical compounds thereof and mixtures thereof, for causing decomposition of the hydrogen peroxide. In a further aspect of the present invention, there is provided a method of lowering the electrochemical corrosion potential of metal alloy cooling tubes in a water-cooled nuclear reactor, the tubes having surfaces exposed to an aqueous liquid containing hydrogen peroxide, comprising the step of injecting matter into said water, said matter selected from the group consisting of manganese, molybdenum, zinc, copper, cadmium, oxides thereof, chemical compounds thereof and mixtures thereof, for causing decomposition of the hydrogen peroxide.
	abstract	An apparatus and method are described which enable real time measurements to measure the margin to criticality in a process for manufacturing fissile materials. An exemplary apparatus includes a neutron source capable of being modulated, an optional moderator to reduce the thermal energy of neutrons from the neutron source, a collimator for controlling the direction of any neutrons emanating in use from the target, a plurality of detector arrays positioned in predetermined locations relative to a process vessel for detecting process variables and for sending signals representative of the process variables in real time to a processor for receiving the signals and converting the detected process variables into margin to criticality measurements.
050846256	claims	1. A portable apparatus for selectively receiving, transporting, and releasing a sample, comprising: a portable storage member for storing and transporting the sample, said storage member including a top side and a bottom side, an adjustable top door located on the top side of said storage member, said top door permitting the sample to enter said storage member through said top side when said top door is in an open position, and said top door isolating the sample within said storage member when said top door is in a closed position, and an adjustable bottom door located on the bottom side of said storage member, said bottom door isolating the sample in said storage member when said bottom door is in the closed position, and said bottom door permitting the sample to leave said storage member through said bottom side when said bottom door is in the open position. a base member including a channel permitting the sample to pass through said base member, an adjustable bottom door supported by said base member, said bottom door including a channel capable of permitting the sample to pass through said bottom door when said bottom door is in an open position such that said bottom door channel is in registration with said base member channel, and said bottom door capable of retaining the sample when said bottom door is in a closed position, a storage member supported by said base member and located above said bottom door, said storage member including a channel capable of permitting the sample to be received, to be stored when said bottom door is in the closed position, and to be released when the bottom door is in the open position, and an adjustable top door supported by said storage member, said top door including a channel capable of permitting the sample to pass into said storage member channel when said top door is in the open position such that said top door channel is in registration with said storage chamber channel, and said top door capable of isolating the sample when said top door is in the closed position. a portable storage member for storing and transporting the samples, said storage member including a top side and a bottom side, said storage member including a plurality of storage chambers arrayed circumferentially with respect to a central axis, an adjustable top door located on the top side of said storage member, said top door including a channel capable of being selectively placed in registration with said respective storage chambers thereby permitting the samples to selectively enter said respective storage chambers through said top side when said top door is in an open position, and said top door isolating the respective samples within said storage chambers by placing said top door channel out of registration with said respective storage chambers when said top door is in a closed position, and a plurality of adjustable bottom doors located respectively on the bottom sides of said respective storage chambers, said bottom doors isolating the samples in said respective storage chambers when said bottom doors are respectively in the closed position, and said bottom doors permitting the samples to leave said respective storage chambers from said bottom side when said respective bottom doors are in respective open positions. means for transporting a plurality of sample containers from a loading area to the thermal analysis apparatus, said transport means including a plurality of sample chambers for housing the respective sample containers, said transport means including means for isolating the sample containers from the environment and for permitting loading and unloading the sample containers respectively to and from a sample receiving portion of the thermal analysis apparatus, and means for supporting said transport means adjacent to the sample receiving portion of the thermal analysis apparatus, said support means permitting registration of said respective sample chambers with the sample receiving portion of the thermal analysis apparatus. the thermal analysis apparatus is a differential thermal analysis (DTA) apparatus, and said transport means includes a channel adapted to receive a sparge tube of the DTA apparatus. means for transporting a plurality of sample containers from a loading area to the differential thermal analysis apparatus, said transport means including a plurality of sample chambers for housing the respective sample containers, said transport means including a channel adapted to receive a sparge tube of the differential thermal apparatus and said transport means including means for isolating the sample containers from the environment and for permitting loading and unloading the sample containers respectively to and from a sample receiving portion of the differential thermal analysis apparatus, and means for supporting said transport means adjacent to the sample receiving portion of the differential thermal analysis apparatus, said support means including means for engaging with said transport means said support means permitting registration of said respective sample chambers with the sample receiving portion of the differential thermal analysis apparatus. loading a plurality of sample containers into a plurality of storage chambers of a portable transporter in a containment area, the transporter serving to isolate the sample containers from the environment, transporting the transporter containing the sample containers to an analytical apparatus, placing the transporter in close proximity to a sample receiving portion of the analytical apparatus, permitting a sample container to move directly from a storage chamber in the transporter to the receiving portion of the analytical apparatus, moving the transporter away from the analytical apparatus, after an analysis is performed, placing the transporter in close proximity to the sample receiving portion of the analytical apparatus, moving the sample container from the analytical apparatus directly into a storage chamber in the transporter, and in the transporter, isolating the sample container from the environment. 2. An apparatus for selectively receiving, retaining, and releasing a sample, comprising: 3. A portable apparatus for selectively receiving, transporting, and releasing a plurality of samples, comprising: 4. The apparatus described in claim 3 wherein said bottom doors are supported by a base member located below and supporting said storage member. 5. The apparatus described in claim 3 wherein said bottom doors include springs for biasing said doors in the closed position. 6. The apparatus described in claim 3 wherein said bottom doors include actuator means for engaging a hand-held implement for opening said doors. 7. The apparatus described in claim 3 wherein said bottom doors include stop means for preventing said doors moving more than a predetermined distance when closing and for compressing a door-biasing spring when opening. 8. The apparatus described in claim 3, further including a channel for receiving a sparge tube of a differential thermal analysis apparatus. 9. An apparatus for handling sample containers for a thermal analysis apparatus and for isolating the sample containers from an adjacent environment, comprising: 10. The apparatus described in claim 9 wherein: 11. The apparatus described in claim 9 wherein said support means includes means for engaging said transport means. 12. The apparatus described in claim 9 wherein said support means includes means for locking together with said transport means. 13. The apparatus described in claim 9 wherein said support means includes means for engaging with the thermal analysis apparatus. 14. An apparatus for handling sample containers for a differential thermal analysis apparatus and for isolating the sample containers from an adjacent environment, comprising: 15. The apparatus described in claim 14 wherein said support means includes means for engaging said transport means. 16. The apparatus described in claim 14 wherein said support means includes means for locking together with said transport means. 17. The apparatus described in claim 14 wherein said support means includes means for engaging with the differential thermal analysis apparatus. 18. A method of handling hazardous analytical samples, comprising the steps of:
	claims	1. A garment for protection from ultraviolet radiation, comprising:A) a torso garment comprising a front side and a rear side that extend from a first edge to an end, said torso garment further comprises first and second lateral sides and first and second shoulder sections, extending from said first and second lateral sides and said first and second shoulder sections are first and second sleeves respectively, said torso garment further comprises securing means to secure a neck gaiter, said securing means includes fasteners, spring snaps assemblies, hook and loop fasteners, and zipper assemblies; andB) first and second hand covers that extend from said first and second sleeves respectively, said first and second hand covers each comprise an elastic band, said elastic band, a distal end, and third and fourth lateral sides define an interior face, said interior face comprises a thumb loop and at least first and second finger loops. 2. The garment for protection from ultraviolet radiation set forth in claim 1, further characterized in that said elastic band, said distal end, and said third and fourth lateral sides also define an exterior face. 3. The garment for protection from ultraviolet radiation set forth in claim 2, further characterized in that each said interior and exterior faces may fold internally within said first and second sleeves respectively. 4. The garment for protection from ultraviolet radiation set forth in claim 2, further characterized in that each said interior and exterior faces may fold externally onto said first and second sleeves respectively. 5. The garment for protection from ultraviolet radiation set forth in claim 1, further characterized in that each said elastic band is sewn to said first and second hand covers respectively. 6. The garment for protection from ultraviolet radiation set forth in claim 1, further characterized in that said thumb loop is positioned at a predetermined distance from said elastic band without reaching said distal end. 7. The garment for protection from ultraviolet radiation set forth in claim 1, further characterized in that said torso garment further comprises a neckband that extends from said first edge, said neckband comprises securing means to secure a neck gaiter. 8. The garment for protection from ultraviolet radiation set forth in claim 1, further characterized in that said securing means are positioned at said neckband. 9. The garment for protection from ultraviolet radiation set forth in claim 7, further characterized in that said securing means are positioned at an interior side of said neckband. 10. The garment for protection from ultraviolet radiation set forth in claim 7, further characterized in that said securing means are positioned at an exterior side of said neckband. 11. The garment for protection from ultraviolet radiation set forth in claim 1, further characterized in that said spring snaps assemblies comprise caps, sockets, studs, and posts. 12. The garment for protection from ultraviolet radiation set forth in claim 1, further characterized in that said neck gaiter comprises an exterior side and an interior side that extend between a top end and a bottom end. 13. The garment for protection from ultraviolet radiation set forth in claim 1, further characterized in that said neck gaiter also comprises said securing means to secure onto said neckband. 14. The garment for protection from ultraviolet radiation set forth in claim 1, further characterized in that said torso garment and said first and second hand covers are made of stretchable fabrics/materials. 15. The garment for protection from ultraviolet radiation set forth in claim 14, further characterized in that said stretchable fabrics/materials include spandex, cotton, cotton blends, nylon, polyesters, and combinations thereof. 16. The garment for protection from ultraviolet radiation set forth in claim 1, further characterized in that said neck gaiter is made of stretchable fabrics/materials. 17. The garment for protection from ultraviolet radiation set forth in claim 16, further characterized in that said stretchable fabrics/materials include spandex, cotton, cotton blends, nylon, polyesters, and combinations thereof.
051587408	abstract	A trepan is provided in a flat end face of an end plug about an external opening of an axial bore through the end plug. The trepan includes an annular groove encircling the external opening and an annular end face portion extending between the annular groove and the external opening. The ratio of the depth of the groove to the diameter of the annular end face portion of the trepan is within a range of from about 1:3 to 1:6. The ratio of the depth of the groove to the width thereof is within a range of from about 1:1 to 1:2. When an axially-directed welding arc is applied to the trepan in the flat end face of the end plug a melting of the material of the trepan into the end face occurs which forms a weld seal such that the trepan is replaced by a shallow concavity overlying the weld seal and extending across the region substantially encircled originally by the annular groove of the trepan.
039379710	description	A focused shield blank 10 is preferably cast or molded from molten lead to present a generally rectangular body having a planar top side 12, a rounded bottom side 14 opposite the top side 12 and longitudinal and transverse edges 16 and 18, respectively, generally perpendicular to the planar top side 10. The shield blanks 10 may be uniformly cast having overall standard dimensions that are compatible with a radiation therapy machine (not shown) in connection with which the blanks 10 are to be ultimately used. If desired, the edges 16 and 18 may be subsequently trimmed for purposes of reducing weight if the full area of the blank 10 is not required. Normally the blanks 10 are at least 2 inches thick. The method of making a focused shield 20 from the shield blank 10 in accordance with the present invention includes the use of a shield blank holding fixture 22 especially constructed for placement on a conventional cutting table 24 of a band saw 26. The fixture 22 may be molded or fabricated in any suitable manner out of fiber glass, plastic, or the like and is comprised of a flat bottomed receptacle 28 having an upwardly facing concavity 30 presenting a rounded, shield blank supporting surface 32 provided with a radius of curvature complementary to the curvature of the bottom side 14 of the shield blank 10. An opening 34 is provided at the bottom center of the concavity 30 to provide clearance for a saw band 36. Reference to FIG. 4 of the drawing will clearly illustrate the complementary nature of the respective radii of curvature of the shield blank 10 and the blank-supporting surface 32 of the fixture 22. Prior to the placement of the shield blank 10 in the concavity 30, a desired radiation field outline 38 corresponding to a predetermined area of a patient that is to be exposed to a field of radiation is drawn on a piece of paper or other sheet material 40. Initially the area to be exposed to a radiation field is outlined on an X-ray radiograph (not shown) of the patient by a radiotherapist and the outline as sketched is then reduced to the appropriate size on the paper 40 using a pantograph (not shown) resulting in the outline 38. It is, of course, to be understood that the degree of reduction of the outline is dependent upon the relative location of the completed focused shield 20 in a radiation therapy machine with respect to the source of the cobalt rays and relative location of the patient. The procedure employed in determing the extent of outline reduction and the relative distances involved between the patient and shield 20, as well as the distance from the shield 20 to the source of the radiation rays, is well known to those experienced in the field of rendering radiation treatments and will not be detailed herein. Although it is not essential, it is desirable to cut the paper 40 to a size corresponding to that of the transverse dimensions of a radiation beam, which has a rectangular cross section at the point the latter strikes the top side 12 of the shield 20. In any event, the desired radiation field outline 38 is to be selectively placed on the paper 40 at a predetermined point relative to a center marking 42 representing the center of a radiation beam when the focused shield 20 is operably inserted in a radiation therapy machine. Once the radiation field outline 38 has been properly located on the paper 40 the latter is then placed on the top side 12 of the shield blank 10 such that the center marking 42 is superimposed over a predetermined center 44 of the shield blank 10 as shown in FIG. 2. Normally the paper 40 is pasted or otherwise secured to the top side 12 to insure that there is no inadvertent displacement of the outline 38 once the paper 40 has been positioned on the shield blank 10. After the paper 38 has been secured to the top side 12, the shield blank 10 is positioned on the fixture 22 with the convexly rounded bottom side 14 of the shield blank 10 resting on the rounded, concave surface 32 of the bowl-shaped receptacle 28. Thus, it will be seen that as the shield blank 10 is normally manipulated, the actuated saw band 36 cuts an aperture 46 therein having a transverse configuration corresponding to that of the predetermined area of a patient to be exposed to a field of radiation as represented by the outline 38. Further, not only will the aperture 46 have the correct configuration, but the aperture 46 will also have beveled sidewalls 48 which are angularly disposed to be in parallelism with the angularity of the radiation rays passing through the aperture 46 adjacent the sidewalls 48 when the shield 20 is properly associated with a radiation therapy machine. By way of further description it will be seen in referring to FIG. 4 that were the blank 10 positioned at the exact center of the concavity 30 any aperture sidewall located at this point would be perpendicular or normal to the top side 12 and be in alignment with the center of a radiation beam during use of the shield 20. Additionally, it will be further observed that when the blank 10 is shifted in any direction away from center, the angularity of the corresponding cut is increased in accordance with the increased angularity of the radiation rays as the outer perimeter of a radiation beam is approached with the sidewalls 48 diverging as the rounded bottom side 14 is approached. Once the focused shield 20 has had the aperture 46 cut therein the focused shield is then securely attached to a mounting plate 50 by means of screws 52 for insertion into a radiation machine. The relative disposition of the focused shield 20, after it is inserted in a radiation machine, is depicted in FIG. 6 wherein the numeral 54 generally identifies a radiation beam as it would appear while emanating from a radiation source 56 of a radiation therapy machine. It will be seen in referring to FIG. 6 that the shield 20 stops the radiation rays of any portion of the radiation beam 24 that is not in direct linear alignment with the aperture 46 while those rays, identified by the numeral 58 in alignment with the aperture 20 pass therethrough in an unimpeded manner. The unique construction of the focused shield 20, typified by the beveled aperture sidewalls 48, permits those rays 58 proximal thereto to pass through the aperture 46 in substantial parallelism with their respective adjacent sidewall 48. As is further apparent in viewing FIG. 6, the radiation rays will be precisely focused on the desired area of the patient for full exposure within the boundary of the outline 38. There will be no unintended "dead" spots beneath the shield 20 because the relative angularity of all portions of the sidewalls 48 is the same as that of the radiation rays passing thereby. The aperture 46 is, in effect, disposed within the shield 20 in axial alignment with the radiation rays 58 passing therethrough and no sections of the upper edge 60 of the aperture 46 block off any of the radiation rays 58. Likewise, none of the radiation rays 58 entering the aperture 46 strike any portion of the sidewalls 48 because of the parallelism of the same with the radiation rays 58. Manifestly, it is to be understood that all radiation rays 58 entering the aperture 46 at the top side 10 will exit in an unimpeded manner at the bottom side 14. The angularity of the sidewalls 48 relative to the top side 12 is determined by the radius of curvature of the surface 32 and the rounded bottom wall 14. However, the required angularity of the sidewalls 48 and therefore the radius of the blank bottom side 14 and surface 32, needs to be calculated, keeping in mind the angle of divergence of the rays within the radiation beam 54. While it is not necessary, the edges 16 and 18 of the blank 10 may also be trimmed to present beveled edges 16a and 18a as shown in FIGS. 5 and 6 to reduce the size of the finished focused shield 20 for purposes of weight reduction. However, in no event should the blank 10 be reduced to a size less than that of the paper 40 or the transverse width of the radiation beam 54 at the point it strikes the top side 12. It is to be further understood that the blank 10 is to be kept in full contact with the surface 32 during the cutting of the aperture 46 to insure the proper angularity of the sidewalls 48. Once the blank 20 has been secured to the mounting plate 50 the focused shield 20 may be repeatedly used to treat a patient simply by inserting the assembled plate and shield into a radiation thereapy machine with the assurance that the exact same spot is being treated each time, that all rays in alignment with the aperture 46 enter the same, and that the rays exit the aperture in the exact same pattern they enter the aperture.
	summary
048662844	description	DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENT The irradiation device of FIG. 1 has a base 1 which is connected to a housing via pivots by means of a telescopic arm 2 and an intermediate arm 3 connected thereto. This housing consists of two elongated parts 4 and 5 which are juxtaposed in a folded state. These parts 4 and 5 accommodate radiation sources (such as high-pressure mercury vapour discharge lamps in which also cobalt and iron are present in the discharge vessel) with reflectors arranged behind them. In a practical embodiment there are provided two radiation sources per part. A radiation exit side is formed on the side of the two parts facing the base. If the two parts 4 and 5 are placed side by side, the entire system can be collapsed to a compact unit. The said parts 4 and 5 are pivotable with respect to each other about an axis perpendicular to the plane through the radiation exit side (the "horizontal" plane). This is realized by providing the end of the intermediate arm 3 (which is in the form of two elongated parallel metal bars 3a and 3b, see FIG. 2) with a coupling member 6 to which the parts 4 and 5 are secured by means of pivots 7 and 8. The longitudinal axes 7a, 8a of these pivots are perpendicular to the plane through the radiation exit side of the parts 4 and 5. The other end of the intermediate arm 3 is pivotably connected to a short pivotal bar 9 secured to the top of the telescopic arm 2. This metal short pivotal bar is bridged by a gas spring 10 whose ends are also pivotably connected to the telescopic arm 2 and the intermediate arm 3, respectively. FIG. 2 shows the irradiation device in the operating condition. The position of the housing is indicated by broken lines. The reference numeral 11 shows diagrammatically that the longitudinal axis of each part is perpendicular to the plane of the drawing. In the position 12 the longitudinal axis is in the plane of the drawing. The parts 4 and 5 are collapsed in this position and are located side by side. The gas spring 10 is shown diagrammatically. It is connected at one end in a position near the end of the telescopic arm and at the other end to the bar 3a. The bars 3a and 3b are pivotably secured to the short pivotal bar 9 which itself is pivotable with respect to arm 2 by means of a pivot (9a). The coupling member 6 is also connected to the ends of the bars 3a and 3b by means of pivots (6a, 6b). The gas spring 10 absorbs three pivotal movements, namely that of the short pivotal bar 9 with respect to arm 2, that of bars 3a and 3 b with respect to bar 9 and that of the arm 2 with respect to the bars 3a and 3b. FIG. 3 shows the irradiation device in a semi-collapsed state. The longitudinal direction of the juxtaposed parts 4 and 5 is shown in a broken line (12). Finally FIG. 4 shows the device in the fully collapsed state. By lowering arm 2 into the base 1 (via a pivot) a compact unit is obtained with member 6 being locked with the arm 2. The base 1 accommodates the electric ballasts of the lamps. The base has also wheels (such as 13, see FIG. 1). Due to the gas spring the device immediately assumes the position as shown in FIG. 3 when the said lock is released. In the operating condition of the irradiation device the synthetic material parts of the housing accommodate high-pressure mercury vapour discharge lamps with a power of 400 W. In addition to 20 mg of mercury the discharge vessel of such a lamp also comprises 0.16 mg of cobalt and 0.3 mg of iron. Such lamps mainly emit UV-A radiation (315-400 nm) and some UV-B radiation (280-315 nm) in addition to infrared radiation. When folding out the parts of the housing (dimensions of each part 25.times.15.times.60 cm) an irradiation field is obtained which is amply sufficient for irradiating the entire human body.
048760633	abstract	In a nuclear fuel bundle having a lower tie-plate, an upper tie-plate and a surrounding channel therebetween, an improved water rod is disclosed for preferable use when fuel rods held between the tie-plates are placed in a 9 by 9 array. Typically, seven fuel rods are omitted centrally of the 9 by 9 array with the middle or fifth row having three rods removed and paired rods being removed in the 4th and 6th row with displacement of the removed pair towards opposite corners. Into the volume created by the removal rods, there are placed two "D" sectioned rods, the "D" rods each being round in cross-section except for a truncating chord, this truncating chord defining the straight back of each "D". In the preferred embodiment one of the "D" water rods is provided with spacer tabs for maintaining spacers separating the fuel rod at their correct elevations. This rod is inserted with an alignment that permits the tabs to pass through the spacers. When the rod is fully inserted, it is rotated to a locking alignment so that the tabs capture the spacers. In this locked alignment, the straight back of the "D" is aligned to confront the straight back of the confronting "D" on the remaining water rod. When the remaining "D" rod is inserted with the backs confronted mutual, locking of both "D" sectioned water rod occurs. Provision is made for narrowing of the upper and lower sections of the water rods. The upper and lower sections, with reduced cross-sections are supported by the upper and lower tie plates, respectively. There results a water rod which efficiently utilizes the available space, has low manufacturing costs, and provides for spacer capture enabling ease of assembly and disassembly of the fuel bundle.
051951207	claims	1. In a radiological apparatus comprising at least one X-ray source irradiating an object to be examined that is located at a distance c from said source, and an antiscatter grid with a spatial frequency F.sub.g placed between said object and a receiver located at a distance d from said object, a method enabling the elimination of the image of said antiscatter grid that consists in: keeping the antiscatter grid fixed, simultaneously irradiating said object by a second source of X-rays having a spatial energy distribution that is substantially identical to that of the first source, and choosing the distance b between the two X-ray sources so that: ##EQU14## with a magnification ##EQU15## modifying the distance b between the two X-ray sources when the magnification G varies so as to always have the equality given by: ##EQU16## 2. A method according to claim 1 further consisting in:
	abstract	The invention relates to an optical filter material made of doped quartz glass, which at a low dopant concentration exhibits spectral transmission as high as possible of at least 80% cm−1 for operating radiation of 254 nm, transmission as low as possible in the wave range below approximately 250 nm, and an edge wavelength λc within the wave range of 230 to 250 nm. It was found that this aim is achieved by doping comprising a gallium compound, which in the wave range below 250 nm has a maximum of an absorption band and thus determines the edge wave range λc.
062755684	abstract	The invention relates to an X-ray apparatus which includes an adjustable X-ray filter. In order to adjust an intensity profile of the X-ray beam, an X-ray absorbing liquid is transported to filter elements of the X-ray filter. Such transport is susceptible to gravitational forces which lead to an irregular hydrostatic pressure distribution in the X-ray filter. In order to reduce the effects of the gravitational forces on the transport of the X-ray absorbing liquid, the duct connecting the filter elements to the reservoir is subdivided into sub-ducts and the reservoir is subdivided into chambers, each chamber being connected to at least one sub-duct. The X-ray apparatus also includes means for keeping the sub-ducts aligned with a horizontal plane.
043494650	claims	1. In a process for the treatment of solid, combustible radioactive wastes comprising plastic and/or rubber, wherein the wastes are contacted with sulfuric acid of a concentration greater than 16 moles per liter and are reacted with this sulfuric acid at an elevated temperature, and concentrated nitric acid or nitrogen oxides are added to the sulfuric acid, whereby oxidation of the wastes occurs below the surface of the sulfuric acid and gaseous by-products and a solid residue are formed, the improvement comprising: (1) subjecting the solid wastes, prior to their reaction with sulfuric acid, to mechanical processing which comprises (2) forming a suspension of the finely ground waste of step (1) with the sulfuric acid at a temperature of less 313.degree. K.; and (3) reacting the finely ground waste with the sulfuric acid in said suspension. 2. Process according to claim 1 wherein the primary comminution of step (b) comprises rendering the waste pieces brittle in liquid nitrogen and then grinding the waste pieces in a cold-grinding mill. 3. Process according to claim 1 or 2 wherein the suspension of the finely ground waste is formed with about 90% sulfuric acid. 4. Process according to claim 1 or 2 wherein the reacting of the finely ground waste with the sulfuric acid is carried out at a pressure of about 100 to 500 m bar. 5. Process according to claim 4 wherein the reacting of the finely ground waste with the sulfuric acid is carried out at a temperature less than or equal to about 493.degree. K. 6. Process according to claim 1 wherein the reacting of the finely ground waste with the sulfuric acid is carried out at a temperature less than or equal to about 493.degree. K. 7. Process according to claim 1 wherein said residue is separated from the remainder of the reaction mixture, is treated in order to remove valuable radioactive material, and is compacted, and wherein the remainder of the reaction mixture and the gaseous by-products are recovered in order to obtain sulfuric and nitric acids for recycling into the process. 8. Process according to claim 1 wherein said radioactive wastes contain radionuclides emitting alpha radiation.
	claims	1. A non-transitory computer readable storage medium storing computer executable instructions that when executed on a processor manage a graphical interface, the medium storing:instructions for providing a graphical interface, where the graphical interface:accesses a hardware device that is associated with a plurality of properties used to communicate with the hardware device, andaccesses a software device being accessible through the graphical interface, the software device being accessible to a computer;instructions for scanning for available hardware devices, whereintwo or more of the available hardware devices each respond to different commands, anda response to a given one of the commands identifies one of the available hardware devices, and the given one of the commands is user-defined;instructions for creating an additional hardware object for each hardware device detected and not already associated with a hardware object;instructions for providing a first interactive hardware object, where the first interactive hardware object:is accessible to the computer,is depicted in the graphical interface, andinteracts with the hardware device;instructions for providing a first configuration represented by the first interactive hardware object, where the first configuration represents a collection of properties used to communicate with the hardware device and a first collection of values associated with the properties;instructions for providing a second interactive hardware object, where the second interactive hardware object:is accessible to the computer,is depicted in the graphical interface, andinteracts with the hardware device;instructions for providing a second configuration represented by the second interactive hardware object, where the second configuration represents the same collection of properties as the first configuration and a second collection of values associated with the properties, wherein at least one value of a property differs between the first configuration and the second configuration;instructions for providing a software object, wherein the software object is representative of the software device, where the software object is depicted in the graphical interface and is configured to be interactive with the software device;instructions for displaying the first hardware object and the second hardware object simultaneously;instructions for receiving, from a user, a selection of at most one hardware object; andinstructions for communicating with the hardware device corresponding to the selected hardware object using the configuration represented by the hardware object. 2. The computer readable storage medium of claim 1, further comprising providing an analysis object, wherein said analysis object is adapted to communicate with at least one of said hardware object and said software object for analysis of data from at least one of said hardware object and said software object. 3. The computer readable storage medium of claim 2, wherein the analysis object filters data. 4. The computer readable storage medium of claim 2, wherein the analysis object plots data. 5. The computer readable storage medium of claim 1, further comprising:instructions for receiving code for execution by the hardware object. 6. The computer readable storage medium of claim 1, wherein a plurality of hardware objects are provided for a single hardware device. 7. The computer readable storage medium of claim 1, wherein a plurality of hardware objects are provided for a plurality of hardware devices. 8. The computer readable storage medium of claim 1, wherein at least one of instructions for providing at least one hardware object and providing at least one software object further comprises instructions for accessing at least one of a hardware object and a software object located on a remote computer. 9. The computer readable storage medium of claim 8, wherein instructions for accessing is performed through a web page. 10. The computer readable storage medium of claim 8, wherein instructions for accessing is performed over a network. 11. The computer readable storage medium of claim 10, wherein instructions for accessing is performed by passing commands over the network in a MATLAB environment. 12. The computer readable storage medium of claim 1, further comprising:instructions for modifying at least one of the hardware object and the software object. 13. The computer readable storage medium of claim 12, wherein modifying specifies a protocol for use by the hardware object for communication with the hardware device. 14. The computer readable storage medium of claim 12, wherein modifying modifies a value stored in an array of an array-based environment. 15. The computer readable storage medium of claim 1, further comprising:instructions for modifying a value stored in an array of an array-based environment, thereby modifying at least one of the hardware object and the software object. 16. The computer readable storage medium of claim 1, further comprising:instructions for exporting data from the graphical interface to an array-based environment. 17. The computer readable storage medium of claim 1, further comprising:instructions for converting user actions with the graphical interface into code. 18. The computer readable storage medium of claim 17, wherein the code is created in a MATLAB environment. 19. The computer readable storage medium of claim 17, wherein the code comprises steps to create an analysis object, configure the analysis object and write and read data from the analysis object. 20. The computer readable storage medium of claim 17, wherein the code comprises an analysis routine. 21. The computer readable storage medium of claim 1, wherein the graphical interface is implemented with an extensible API. 22. The computer readable storage medium of claim 1, further comprising:instructions for generating an analysis object so that the analysis object can be used in MATLAB. 23. The computer readable storage medium of claim 1, further comprising:instructions for generating an analysis object that can be used in SIMULINK. 24. The computer readable storage medium of claim 1, wherein the graphical interface is adapted to operate on a plurality of operating systems. 25. The computer readable storage medium of claim 1, wherein the graphical interface comprises a tree view, wherein the tree view groups the hardware objects and the software objects by a functionality characteristic. 26. The computer readable storage medium of claim 1, wherein the hardware object enables communication between the graphical interface and the hardware device, and the software object enables communication between the graphical interface and the software device. 27. A method for managing an interface, the method comprising:providing a graphical interface that provides interaction with an array-based environment, a hardware device and a software device being accessible through the graphical interface, the software device being accessible to a computer, the hardware device associated with a plurality of properties used to communicate with the hardware device;scanning for available hardware devices, whereintwo or more of the available hardware devices each respond to different commands, anda response to a given one of the commands identifies one of the available hardware devices, and the given one of the commands is user-defined;creating an additional hardware object for each hardware device detected and not already associated with a hardware object;providing a first hardware object, where the first hardware object:is accessible to the computer,is depicted in the graphical interface, andinteracts with the hardware device;providing a first configuration of the hardware device represented by the first hardware object, the first configuration representing a collection of properties used to communicate with the hardware device and a first collection of values associated with the properties;providing a second hardware object, where the second hardware object:is accessible to the computer,is depicted in the graphical interface, andinteracts with the hardware device;providing a second configuration of the hardware device represented by the second hardware object, the second configuration representing the same collection of properties as the first configuration and a second collection of values associated with the properties, wherein at least one value of a property differs between the first configuration and the second configuration;providing at least one software object, representative of the software device, where the software object is depicted in the graphical interface, and is configured to be interactive with the software device;updating the graphical interface when the first hardware object, the second hardware object, or the software object are changed in the array-based environment; anddisplaying the hardware object and the software object to a user. 28. The method of claim 27, further comprising:receiving code for execution by the hardware object. 29. The method of claim 27, wherein at least one additional hardware object is provided for the hardware device. 30. The method of claim 27, wherein additional hardware objects are provided for a plurality of hardware devices. 31. The method of claim 27, further comprising:providing an analysis object adapted to communicate with at least one of the hardware object and the software object. 32. The method of claim 27, wherein at least one of providing at least one hardware object and providing at least one software object further comprises accessing at least one of a hardware object and a software object located on a remote computer. 33. The method of claim 27, further comprising:modifying at least one of the hardware object and the software object. 34. The method of claim 33, wherein modifying specifies a protocol for use by the hardware object for communication with the hardware device. 35. The method of claim 33, wherein modifying modifies a value stored in an array of an array-based environment. 36. The method of claim 27, further comprising generating an analysis object that can be used in MATLAB. 37. The method of claim 27, further comprising generating an analysis object that can be used in SIMULINK. 38. The method of claim 27, wherein the hardware object enables communication between the graphical interface and the hardware device, and the software object enables communication between the graphical interface and the software device. 39. A computing device comprising:an array-based environment;a storage medium for storing and a processor for processing:a graphical interface, at least one hardware device and one software device being accessible through the graphical interface, the hardware device associated with a plurality of properties used to communicate with the hardware device;instructions for scanning for available hardware devices, whereintwo or more of the available hardware devices each respond to different commands, anda response to a given one of the commands identifies one of the available hardware devices, and the given one of the commands is user-defined;instructions for creating an additional hardware object for each hardware device detected and not already associated with a hardware object;a first hardware object, where the first hardware object:is accessible to the computer,is depicted in the graphical interface, andinteracts with the hardware device;a first configuration of the hardware device represented by the first hardware object, the first configuration representing a collection of properties used to communicate with the hardware device and a first collection of values associated with the properties;a second hardware object, where the second hardware object:is accessible to the computer,is depicted in the graphical interface, andinteracts with the hardware device;a second configuration of the hardware device represented by the second hardware object, the second configuration representing the same collection of properties as the first configuration and a second collection of values associated with the properties, wherein at least one value of a property differs between the first configuration and the second configuration;a plurality of software objects, each representative of a software device accessible to the computer, where each of the software objects is depicted in the graphical interface and is configured to be interactive with the software device; anda display device to display the first hardware object, the second hardware object, and the plurality of software objects to a user in a single graphical interface simultaneously, wherein the first hardware object, the second hardware object, and the plurality of software objects are accessible through both the array-based environment and the graphical interface. 40. The computing device of claim 39, wherein the system receives code for execution by the hardware objects. 41. The computing device of claim 39, wherein a plurality of hardware objects are provided for a single hardware device. 42. The computing device of claim 39, wherein a plurality of hardware objects are provided for a plurality of hardware devices. 43. The computing device of claim 39, wherein an analysis object is provided for communicating with at least one of the hardware objects and the software objects. 44. The computing device of claim 39, wherein at least one of the hardware objects and the software objects are located on a remote computer. 45. The computing device of claim 39, at least one of the hardware objects and the software objects are modified by the processor. 46. The computing device of claim 39, wherein at least one of the hardware objects and the software objects are modified by the processor such that a protocol is specified for use by the at least one of the hardware objects for communication with the hardware device. 47. The computing device of claim 39, wherein at least one of the hardware objects and the software objects are modified by the processor such that a value is stored in an array of an array-based environment. 48. The computing device of claim 39, wherein the hardware object enables communication between the graphical interface and the hardware device, and the software object enables communication between the graphical interface and the software device.
	summary
	summary
047073276	summary	The invention relates to a container system for a high-temperature nuclear reactor, including an outer metallic pressure vessel, and an inner metallic core barrel, feeding and discharge of cooling fluid taking place at the lower end of the core barrel. A reactor of this type has been described in German Published, Non-Prosecuted Application DE-AS No. 30 16 402, corresponding to U. S. Pat. No. 4,476,089. Among other things, that patent proposes the placement of a core barrel containing the nuclear reactor proper in a steel pressure vessel such as has been used heretofore in the construction of water-cooled reactors. It was assumed in that disclosure that the spaces between the pressure vessel and its internal parts would be in communication with each other everywhere and would be filled with the helium used as the cooling fluid. Furthermore, no statements were made in the above-mentioned publication regarding the manner in which the core barrel should be fastened in the pressure vessel. Additionally, there were no provisions for preventing contamination of spaces between the pressure vessel and core barrel during servicing. It is accordingly an object of the invention to provide a container system for a high-temperature nuclear reactor, which overcomes the hereinafore-mentioned disadvantages of the heretofore-known devices of this general type, and in which the part of the space between the core barrel and the pressure vessel, in which devices are disposed which need frequent testing or servicing (for instance, the absorber rod drives normally placed in the ceiling of the core barrel, or the feed device such as for spherical fuel elements) can be made accessible without the danger of a break-in of the ambient atmosphere (i.e., of oxygen-containing air) into the interior of the core barrel, which would lead to considerable corrosion of the internal parts thereof, formed of carbon blocks and/or graphite, or of the fuel elements. With the foregoing and other objections in view there is provided, in accordance with the invention, a container system for a high-temperature nuclear reactor, comprising an outer metallic pressure vessel having an inwardly-protruding flange, an inner metallic core barrel resting tightly on the flange, and means disposed below the flange at a lower end of the core barrel for feeding and discharging cooling fluid, the core barrel being gas-tight above the flange. A pad which is already required for holding the core barrel in the pressure vessel is constructed in such a way that it simultaneously serves as a partition between the upper part of the above-mentioned space and the lower part thereof. Only the feed and discharge devices for the cooling fluid are located in the space. If the weight of the core barrel is known, one of ordinary skill in the art can construct the required size of the sealing surface without difficulty so that with a suitable construction of the sealing surface, the required sealing effect is achieved due to the weight of the core barrel itself. (Due to the high operating temperatures, elastomer seals are out of the question, and preferably, metal O-rings are used). Since the core barrel should be removable for repair purposes, a form-locking connection of the core barrel and the mounting flange is dispensed with. A form-locking connection is one in which parts are locked together by virtue of their own shape. The weight of the core barrel is also sufficient to reliably prevent the core barrel from being lifted off if a slightly higher pressure prevails in the lower part of the space than in the upper part. (Such a pressure difference is 2 bar for an operating pressure of the reactor of, for instance, 40 bar). The upper space is advantageously also filled with helium which, however, is not contaminated because it does not circulate through the nuclear reactor and other parts of the plant (such as heat exchanger); therefore, no radioactive deposits can occur in the upper space, which would limit the desired accessibility of the core barrel. Since no piping carrying the cooling medium is present in the upper space, the core barrel can be made tight at that location, and the feedthroughs required for the absorber rod drives etc. can be made gas-tight by conventional means (such as sliding valves). In order to prevent possible contamination of the gas filling the upper space by microscopic leaks at the sealing surface between the core barrel and the pressure vessel, the upper space is advantageously kept at a slight overpressure. In accordance with another feature of the invention, the core barrel is disposed at a distance from the pressure vessel defining a space therebetween, the flange dividing the space into upper and lower spaces, and including an equalization line connected between the upper and lower spaces, and means for shutting off the line. In this way, larger pressure differences between the upper and the lower space are controlled. By limiting the pressure difference, the passage of larger leakage amounts in one direction or the other is prevented. This device is sufficient to equalize the pressure differences expected during normal operation. In accordance with a concomitant feature of the invention, there is provided another equalization line leading from the lower end to the upper end of the core barrel interconnecting the upper and lower spaces, and a rupture disc protector closing off the other line. These features are provided to take care of a sudden occurrence of a major leak in the pressure vessel. Rupture disc protectors are well or known components which are completely tight in normal operation but completely release a pipe line which is provided for this purpose and is constructed for obtaining a sufficiently fast pressure equalization if a given response pressure is exceeded. However, a line connecting the lower space to the upper space over the shortest path would favor the occurrence of natural convection, wherein the air penetrating into the upper space through the leak could also reach into the interior of the core barrel, which is to be avoided, as explained above. The second equalization line provides a direct path to the lower end of the lower space for the cold air which has penetrated through the leak into the upper space after the rupture disc has opened, and thus, the lower space is therefore also flooded. The temperature difference present in the gas which circulates in the core barrel by natural circulation after the cooling blowers have failed or have been shut down and which removes the decay heat, causes a stratification which prevents the penetration of the corroding air therein. 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 container system for a high-temperature nuclear reactor, 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.
	claims	1. In a computer system comprising a display, user interface selection device and user interface data entry device, a process for playing back a recorded collaborative electronic presentation in which a plurality of participant's interacted with displayed presentation data via an interactive virtual team worksite over a distributed computer network, wherein the collaborative electronic presentation was recorded by capturing and storing the interactions between each participant and the displayed presentation data, wherein each interaction event is timestamped and linked to a data file comprising the presentation data, said playback process comprising the process actions of:displaying a worksite window on the display to a team member logged onto said team worksite, wherein the worksite window comprises a plurality of sectors;inputting data and implementing commands entered by the team member playing back a recorded collaborative electronic presentation via said selection and data entry devices to,display a list of recorded collaborative electronic presentation sessions in the worksite window in response to a team member command to do so, said display of the sessions list comprising,assigning a name to each session for display purposes, anddisplaying a history sector of the window comprising a display area in which the list of recorded collaborative electronic presentation sessions is displayed by their assigned names,upon selection of a displayed recorded presentation session by the team member, display information about the session in the worksite window comprising displaying a presentation timeline in a timeline area of the history sector, wherein the timeline is a visual representation of the recorded interactions between each participant and the displayed presentation data over the course of the presentation, andplay back the recorded collaborative electronic presentation session selected by the team member in the worksite window in response to the team member's command to do so, wherein playing back the selected presentation session comprises displaying the presentation data from the associated data file in an order the presentation data was originally presented and reproducing the recorded interactions between each participant and the displayed presentation data at the same point in the presentation that they were originally performed, based on the timestamps. 2. The process of claim 1, wherein the process action of displaying the presentation timeline, comprises an action of displaying a timeline representing all or just a portion of the presentation depending on a zoom command entered by the team member, which has been rescaled to fit the timeline area. 3. The process of claim 1, wherein the process action of displaying the presentation timeline comprises displaying a horizontal line representing the length of the presentation in terms of time and displaying vertical lines representing recorded interactions each of which crosses the horizontal line at a point representing the time in the presentation the interaction took place. 4. The process of claim 3, wherein the process action of displaying vertical lines representing recorded interactions, comprises an action of color-coding the vertical lines to indicate the type of interaction they represent. 5. The process of claim 4, wherein the interactions comprises highlighting portions of the displayed data, using a pointer to call attention to a portion of the data, and/or modifying the data. 6. The process of claim 1, wherein the process action of playing back the recorded collaborative electronic presentation session in the worksite window comprises an action of displaying a collaborative presentation sector comprising a display area in which the presentation data and participant interactions associated with a current portion of the presentation session being played back are displayed. 7. The process of claim 6, wherein the process action of displaying the presentation timeline further comprises an action of displaying a current portion indicator that points to the part of the presentation in the timeline that is currently featured in the display area of the collaborative presentation sector. 8. In a computer system comprising a display, user interface selection device and user interface data entry device, a process for playing back a recorded collaborative electronic presentation in which a plurality of participant's interacted with displayed presentation data via an interactive virtual team worksite over a distributed computer network, wherein the collaborative electronic presentation was recorded by capturing and storing the interactions between each participant and the displayed presentation data, wherein each interaction event is timestamped and linked to a data file comprising the presentation data, said playback process comprising the process actions of:displaying a worksite window on the display to a team member logged onto said team worksite, wherein the worksite window comprises a plurality of sectors;inputting data and implementing commands entered by the team member playing back a recorded collaborative electronic presentation via said selection and data entry devices to,display a list of recorded collaborative electronic presentation sessions in the worksite window in response to a team member command to do so,upon selection of a displayed recorded presentation session by the team member, display information about the session in the worksite window, andplay back the recorded collaborative electronic presentation session selected by the team member in the worksite window in response to the team member's command to do so, wherein playing back the selected presentation session comprises displaying a collaborative presentation sector comprising a display area in which the presentation data and participant interactions associated with a current portion of the presentation session being played back are displayed and displaying the presentation data from the associated data file in an order the presentation data was originally presented and reproducing the recorded interactions between each participant and the displayed presentation data at the same point in the presentation that they were originally performed, based on the timestamps, andwherein displaying information about the selected presentation session in the worksite window, comprises displaying a presentation event listing in the display area of the history sector of in lieu of the list of recorded collaborative electronic presentation sessions, wherein the presentation event listing comprises an identification of the part of the presentation that is currently featured in the display area of the collaborative presentation sector and a list of all the recorded interactions between each participant and the displayed presentation data corresponding to the part of the presentation that is currently featured in the display area of the collaborative presentation sector. 9. In a computer system comprising a display, user interface selection device and user interface data entry device, a process for playing back a recorded collaborative electronic presentation in which a plurality of participant's interacted with displayed presentation data via an interactive virtual team worksite over a distributed computer network, wherein the collaborative electronic presentation was recorded by capturing and storing the interactions between each participant and the displayed presentation data, wherein each interaction event is timestamped and linked to a data file comprising the presentation data, said playback process comprising the process actions of:displaying a worksite window on the display to a team member logged onto said team worksite, wherein the worksite window comprises a plurality of sectors;inputting data and implementing commands entered by the team member playing back a recorded collaborative electronic presentation via said selection and data entry devices to,display a list of recorded collaborative electronic presentation sessions in the worksite window in response to a team member command to do so,upon selection of a displayed recorded presentation session by the team member, display information about the session in the worksite window,play back the recorded collaborative electronic presentation session selected by the team member in the worksite window in response to the team member's command to do so, wherein playing back the selected presentation session comprises displaying the presentation data from the associated data file in an order the presentation data was originally presented and reproducing the recorded interactions between each participant and the displayed presentation data at the same point in the presentation that they were originally performed, based on the timestamps, andjump to a point in the presentation selected by the team member and continuing the playback from that point. 10. In a computer system comprising a display, user interface selection device and user interface data entry device, a process for playing back a recorded collaborative electronic presentation in which a plurality of participant's interacted with displayed presentation data via an interactive virtual team worksite over a distributed computer network, wherein the collaborative electronic presentation was recorded by capturing and storing the interactions between each participant and the displayed presentation data, wherein each interaction event is timestamped and linked to a data file comprising the presentation data, said playback process comprising the process actions of:displaying a worksite window on the display to a team member logged onto said team worksite, wherein the worksite window comprises a plurality of sectors;inputting data and implementing commands entered by the team member playing back a recorded collaborative electronic presentation via said selection and data entry devices to,display a list of recorded collaborative electronic presentation sessions in the worksite window in response to a team member command to do so,upon selection of a displayed recorded presentation session by the team member, display information about the session in the worksite window,play back the recorded collaborative electronic presentation session selected by the team member in the worksite window in response to the team member's command to do so, wherein playing back the selected presentation session comprises,displaying the presentation data from the associated data file in an order the presentation data was originally presented and reproducing the recorded interactions between each participant and the displayed presentation data at the same point in the presentation that they were originally performed, based on the timestamps,inquiring of the team member whether the member's interactions with the displayed presentation data is to be recorded or not, andcapturing and storing the interactions between the team member and the displayed presentation data whenever the team member responds to the inquiry that the interactions are to be recorded, wherein each interaction event is timestamped and linked to a file comprising the presentation data. 11. The process of claim 10, wherein the process action of displaying a list of recorded collaborative electronic presentation sessions, further comprises including recorded collaborative electronic presentation sessions that comprise the interactions of a team member captured and stored during the play back of a previously recorded presentation session, and wherein the process action of playing back a recorded collaborative electronic presentation session further comprises playing back a recorded collaborative electronic presentation session selected from the list that comprises the interactions of a team member captured and stored during the play back of a previously recorded presentation session. 12. The process of claim 11, wherein the process action capturing and storing the interactions between the team member and the displayed presentation data comprises storing the interaction data in a separate file, and wherein the process action of playing back a recorded collaborative electronic presentation session that comprises the interactions of a team member captured and stored during the play back of a previously recorded presentation session, comprises playing back only those interactions of the team member stored in said separate in conjunction with the playback of the presentation data. 13. The process of claim 11, wherein the process action of capturing and storing the interactions between the team member and the presentation data comprises storing the interaction data in a file along with the interaction data associated with any other person whose interactions were captured and linked to the presentation data previously, and wherein the process action of playing back a recorded collaborative electronic presentation session that comprises the interactions of a team member captured and stored during the play back of a previously recorded presentation session, comprises playing back those interactions of the team member captured and stored during the play back of a previously recorded presentation session and those of any other person whose interactions were captured and linked to the presentation data previously. 14. The process of claim 10, wherein the collaborative electronic presentation was recorded by capturing and storing the interactions between each participant and the displayed presentation data in a separate file, and wherein the process action capturing and storing the interactions between the team member and the displayed presentation data during a play back comprises storing the interaction data in a separate file. 15. The process of claim 14, wherein the process action of displaying the list of recorded collaborative electronic presentation sessions further comprises displaying, for each session listing, a list of the participants whose interactions with the presentation data were captured and stored in a separate file, and wherein the process action of playing back a recorded collaborative electronic presentation session comprises, displaying the presentation data from the associated data file in an order it was originally presented, and reproducing only the recorded interactions between each participant whose name is selected by a team member playing back the presentation session from the list of the participants whose interactions with the presentation data were captured and stored in a separate file.
	claims	1. A nuclear core component hold-down assembly that accommodates top mounted instrumentation systems comprising:a base plate sized to seat within a top nozzle of a nuclear reactor fuel assembly above and spaced from an adapter plate of the top nozzle, the base plate having a number of openings that align with an equal number of holes in the adapter plate through each of which a corresponding control rod guide thimble in the fuel assembly is accessed;a vertical, hollow sleeve elongated along a dimension of elongation, having an axis along said dimension of elongation, the sleeve extending through and below a central opening in the base plate to mate with an upper opening in an instrument thimble in the fuel assembly, the sleeve extending vertically above the base plate and sized to extend through an upper core plate in the reactor when installed in a core of the reactor;a hold-down bar slidably mounted on the sleeve and having an axial travel length that is restrained a given distance below the top of the sleeve so that the sleeve extends above the hold-down bar when the hold-down bar is fully extended in a direction away from the base plate; anda spring concentrically positioned around the sleeve and extending substantially between the hold-down bar and the base plate. 2. The hold-down assembly of claim 1 wherein the spring comprises two concentric springs. 3. The hold-down assembly of claim 1 wherein the vertical, hollow sleeve is sized to extend above the upper core plate in the reactor when installed in the core of the reactor. 4. The hold-down assembly of claim 1 wherein the sleeve has an axially extending slot that extends over the travel length of the hold-down bar and the hold-down bar has a radially, inwardly extending pin that respectively travels within the slot. 5. The hold-down assembly of claim 1 wherein an interior hollow cavity of the vertical sleeve has two different diameters along its axial length between the base plate and the hold-down bar. 6. The hold-down assembly of claim 5 wherein an upper section of the interior hollow cavity of the vertical sleeve has a larger inner diameter to receive an instrumentation shroud, than a lower portion of the interior hollow cavity of the vertical sleeve which guides a top mounted instrumentation through the top nozzle adapter plate into the instrumentation thimble in the fuel assembly. 7. The hold-down assembly of claim 6 wherein the upper section of the interior hollow cavity of the vertical sleeve has a sufficient cavity length to accommodate a differential thermal and irradiation growth between the fuel assembly and a reactor vessel in which the fuel assembly will be supported. 8. An elongated nuclear reactor fuel assembly having an axial dimension along its elongated length, the fuel assembly comprising:a top nozzle having an adapter plate;a plurality of control rod guide thimbles extending into corresponding openings in the adapter plate;an instrumentation thimble extending into a central opening in the adapter plate; anda hold-down assembly that accommodates top mounted instrumentation systems comprising;the elongated nuclear reactor fuel assembly above and spaced from the adapter plate of the top nozzle, the base plate having a number of openings that align with an equal number of holes in the adapter plate through each of which the corresponding control rod guide thimble in the fuel assembly is accessed;a vertical, hollow sleeve elongated along a dimension of elongation, having an axis along said dimension of elongation, the sleeve extending through and below a central opening in the base plate to mate with an upper opening in the instrumentation thimble in the fuel assembly, the sleeve extending vertically above the base plate and sized to extend through an upper core plate in a nuclear reactor when installed in a core of the reactor;a hold-down bar slidably mounted on the sleeve and having an axial travel length that is restrained a given distance below the top of the sleeve so that the sleeve extends above the hold-down bar when the hold-down bar is fully extended in a direction away from the base plate; anda spring concentrically positioned around the sleeve and extending substantially between the hold-down bar and the base plate. 9. The elongated nuclear reactor fuel assembly of claim 8 wherein the vertical, hollow sleeve is sized to extend above the upper core plate in the reactor when installed in the core of the reactor. 10. The elongated nuclear reactor fuel assembly of claim 8 wherein the sleeve has an axially extending slot that extends over the travel length of the hold-down bar and the hold-down bar has a radially, inwardly extending pin that respectively travels within the slot. 11. The elongated nuclear reactor fuel assembly of claim 8 wherein an interior hollow of the vertical sleeve has two different diameters along its axial length between the base plate and the hold-down bar. 12. The elongated nuclear reactor fuel assembly of claim 11 wherein an upper section of the interior hollow of the vertical sleeve has a larger inner diameter to receive an instrumentation shroud, than a lower portion of the interior hollow of the vertical sleeve which guides a top mounted instrumentation through the top nozzle adapter plate into the instrumentation thimble in the fuel assembly. 13. The elongated nuclear reactor fuel assembly of claim 12 wherein the upper section of the interior hollow cavity of the vertical sleeve has a sufficient cavity length to accommodate a differential thermal and irradiation growth between the fuel assembly and a reactor vessel in which the fuel assembly will be supported. 14. A nuclear reactor power generating system having a core comprising a number of fuel assemblies, at least some of the fuel assemblies comprising:a top nozzle having an adapter plate;a plurality of control rod guide thimbles extending into corresponding openings in the adapter plate;an instrumentation thimble extending into a central opening in the adapter plate; anda hold-down assembly that accommodates top mounted instrumentation systems comprising;a base plate sized to seat within the top nozzle of a nuclear reactor fuel assembly above and spaced from the adapter plate of the top nozzle, the base plate having a number of openings that align with an equal number of holes in the adapter plate through each of which the corresponding control rod guide thimble in the fuel assembly is accessed;a vertical, hollow sleeve elongated along a dimension of elongation, having an axis along said dimension of elongation, the sleeve extending through and below a central opening in the base plate to mate with an upper opening in the instrumentation thimble in the fuel assembly, the sleeve extending vertically above the base plate and sized to extend through an upper core plate in a nuclear reactor when installed in a core of the reactor;a hold-down bar slidably mounted on the sleeve and having an axial travel length that is restrained a given distance below the top of the sleeve so that the sleeve extends above the hold-down bar when the hold-down bar is fully extended in a direction away from the base plate; anda spring concentrically positioned around the sleeve and extending substantially between the hold-down bar and the base plate. 15. The nuclear reactor power generating system of claim 14 wherein the vertical, hollow sleeve is sized to extend above the upper core plate in the reactor when installed in the core of the reactor. 16. The nuclear reactor power generating system of claim 14 wherein the sleeve has an axially extending slot that extends over the travel length of the hold-down bar and the hold-down bar has a radially, inwardly extending pin that respectively travels within the slot. 17. The nuclear reactor power generating system of claim 14 wherein an interior hollow of the vertical sleeve has two different diameters along its axial length between the base plate and the hold-down bar. 18. The nuclear reactor power generating system of claim 17 wherein an upper section of the interior hollow of the vertical sleeve has a larger inner diameter to receive an instrumentation shroud, than a lower portion of the interior hollow of the vertical sleeve which guides a top mounted instrumentation through the top nozzle adapter plate into the instrumentation thimble in the fuel assembly. 19. The nuclear reactor power generating system of claim 18 wherein the upper section of the interior hollow cavity of the vertical sleeve has a sufficient cavity length to accommodate a differential thermal and irradiation growth between the fuel assembly and a reactor vessel in which the fuel assembly will be supported.
045307830	description	EXAMPLE 1 A simulated, non-radioactive solution (A) is prepared having the following composition: NaNO.sub.3 : 357p PA1 HNO.sub.3 : 126p PA1 Fe.sub.2 (SO.sub.4).sub.3 : 60p PA1 H.sub.2 SO.sub.4 concentrated: 44p PA1 H.sub.2 O to make: 1000p PA1 LiNO.sub.3 : 0.75p PA1 Na.sub.2 CO.sub.3 : 33.7p PA1 NaNO.sub.3 : 2.33p PA1 (NH.sub.4).sub.2 CO.sub.3 : 1.50p PA1 H.sub.3 BO.sub.3 : 136p PA1 NaBO.sub.2 hydrate: 100p PA1 H.sub.2 O to make: 1000p PA1 Na.sub.3 PO.sub.4 : 30p PA1 FeCl.sub.3 : 15p PA1 CaCl.sub.2 : 10p PA1 H.sub.2 O to make: 1000p Separately there are synthetized, by using normal esterification methods: an unsaturated polyester of the type (II) by reacting 3440p of isopropylidene-bis-(p-phenyleneoxy-propanol-2) and 1160p of fumaric acid until an acid number of 15-20 has been reached and by dissolving the polymeric ester thus obtained in styrene in the ratio 60:40 approximately, adding hydroquinone as stabilizer in the concentration of 0.015 g %. A composition (B) is thus obtained. Another type (I) unsaturated polyester is obtained by reacting 1300p of neopentyl glycol, 340p of diethylene glycol, 735p of maleic anhydride, 1245p of isophthalic acid, until an average molecular weight above 1200 is attained. This unsaturated polymeric ester is mixed with styrene so as to obtain a solution containing about 40% of monomer. Said solution is stabilized with 0.014% of hydroquinone. A composition (C) is thus obtained. In order to prepare the emulsion which is an object of the present invention, 500p of composition (B) and 500p of the composition (C), are mixed in a vessel having a capacity of 2500 ml. 6p of commercial benzoyl peroxide, in the form of paste, in dibutyl phthalate containing 50% of peroxide are added to said mixture. 670p of solution (A) are added under stirring and at a controlled speed so as to complete the operation in 15 min. The formation of the emulsion occurs rapidly. Once the solution has been completely added, the mass is stirred for a further few minutes, whereafter 2p of N,N-dimethylamino-p-toluidine are slowly added. The stirring is continued for 2 more minutes to promote dissolution of the reagents. The stirrer is then removed and the organic material is allowed to polymerize, thus incorporating the sample solution. A solid block free from exudation and cracks is obtained. EXAMPLE 2 A simulated non-radioactive solution (D) is prepared, having the following composition The solution is to be maintained at a temperature in the order of 40.degree.-50.degree. C. to avoid precipitation of the boric acid. 1000p of a mixture of the compositions (B) and (C) obtained according to the indications of example 1, in the ratio 30/70, are separately prepared in a vessel having a capacity of 2500 ml, and 5p of 50% benzoyl peroxide paste in dibutyl phthalate are added, and stirring is started. As soon as the benzoyl peroxide has been dissolved, the addition of 670p of solution (D) is initiated, while maintaining it warm as stated hereinbefore. The emulsion is formed rapidly even under mild stirring. Once the addition of the solution has ended, 1p of N,N-dimethylamino-p-toluidine is added and the stirring is continued for some further minutes. The stirrer is removed and the emulsion is allowed to stay. The polyester polymerizes rapidly: a solid block, having a dry surface and free from cracks is thus obtained. EXAMPLE 3 An emulsion prepared as described in example 1 is cast, before it polymerizes, into glass tubes having an inner diameter of 20 mm. It is allowed to polymerize. From the solid thus obtained cylinders 20 mm high are cut, which are subjected to a rapid leaching test in a Soxhlet apparatus at 99.degree. C. (according to SOXHLET LEACH TEST PROCEDURE" for testing of "Solidified Radioactive Waste"). In order to quantify the resistance to leaching, the release of sodium is considered, as a measure of the "speed of leaching", the value of which in the present case is found to be 1.0.multidot.10.sup.-2 g cm.sup.-2 days.sup.-1 (expressed in units g cm.sup.-2 days.sup.-1 in order to be homogeneous with the data that are found in the literature). Said speed is calculated by the following formula ##EQU1## wherein: R.sub.Si =Soxhlet leaching speed based on leached ions (g cm.sup.-2 days.sup.-1) a=quantity of ions in solutions (g) A.sub.o =Na.sup.+ ions in the sample (g) W.sub.1 =initial weight of the test specimen. S=surface (cm.sup.2) t=time (days) Further it is found that the amount of Na.sup.+ ions extracted from the test specimen is lower than 10% after 72 hours. EXAMPLE 4 An emulsion prepared as described in example 2 is cast, before it polymerizes, into glass tubes having an inner diameter of 20 mm. It is allowed to polymerize. 20 mm high cylinders are cut from the solid thus produced, which cylinders are subjected to a leaching test as described in example 3. The values of the leaching speed are in the order of 1.5.multidot.10.sup.-2 g cm.sup.-2 days.sup.-1. The amount by weight of Na.sup.+ ion leached from the test specimen is in the order of 14% after 72 hours. EXAMPLE 5 An emulsion is prepared as in example 1, but with a ratio of composition (B) to composition (C) of 30 parts to 70 parts. Test specimen prepared as in the example 3, subjected to leaching, yield a value of leaching speed of 1.6.multidot.10.sup.-2 g cm.sup.-2 days, and a percentage of Na.sup.+ ion leached in of the order of 16% after 72 hours. EXAMPLE 6 An emulsion is prepared as in the example 1, it is cast into glass tubes having an inner diameter of 25 mm and cylinders are cut as described in the example 3 but having dimensions of 25 mm diameter and 50 mm height. Resistance to compression tests are carried out on said cylinders using an Instron 10 KN electronic dynamometer, according to norm ASTM D 695. The values obtained for the compression strength are in the order of 120 kg/cm.sup.2, measured at 10% of deformation, at which point the test specimen has not broken. The value of the elasticity modulus at compression measured simultaneously is in the order of 4500 kg/cm.sup.2. EXAMPLES 7-17 The following examples illustrate the importance of the variation of the content of each of two unsaturated polyesters (I) and (II) in the mixture on the speed of leaching determined by the method described in example 1, the type and concentration of emulsified solution being the same in all of the following examples, and being that described in example 1. A series of emulsions is prepared, wherein the composition of the mixtures varies as described in the following table, and wherein the ratio 1:1 is always maintained between the organic phase (immobilizing matrix) and the aqueous phase (radioactive waste, as dispersed phase). The emulsions are prepared as described in example 1. The test specimens are prepared and subjected to leaching as described in example 3. The results obtained are listed in the following table. Example 7 and 8 are comparison examples. TABLE ______________________________________ Mixture ratio Polyester Polyester leaching speed Example type (I) type (II) g cm.sup.-2 .multidot. days.sup.-1 ______________________________________ 7 0 100 unmeasurable: corroded sample 8 10 90 as above 9 20 80 80 .multidot. 10.sup.-2 10 30 70 50 .multidot. 10.sup.-2 11 40 60 35 .multidot. 10.sup.-2 12 50 50 8.1 .multidot. 10.sup.-2 13 60 40 2 .multidot. 10.sup.-2 14 70 30 1.2 .multidot. 10.sup.-2 15 80 20 2.8 .multidot. 10.sup.-2 16 95 5 15.0 .multidot. 10.sup.-2 17 100 0 25 .multidot. 10.sup.-2 ______________________________________ EXAMPLE 18 An emulsion is prepared as described in example 2 and test specimens are prepared therefrom as described in example 6. Said specimens, subjected to compression test according to norm ASTM D 695, evidence a compression strength of 95 kg/cm.sup.2 measured at 10% of deformation, at which point the test specimen has not broken. Simultaneously the elasticity modulus at compression is measured and has a value of about 4200 kg/cm.sup.2. EXAMPLE 19 A simulated non-radioactive solution (E) is prepared having the following composition: Separately, 1000p of mixture of resins having composition (B)+(C) prepared according to the indication of example 1, in ratio 70:30, are prepared in a vessel having a capacity of 2500 ml and 5p of 50% benzoyl peroxide paste in dibutyl phthalate are added, and the stirring is initiated. As soon as the benzoyl peroxide has been dissolved, the addition of 818 p of solution (E) is started, always continuing the stirring. Once the addition of solution is ended, 1p of N,N-dimethyalmine-p-toluidine is added and the stirring is continued for some further minutes. The emulsion thus prepared, is cast into a glass tube having an inner diameter of 20 mm. It is allowed to polymerize. 20 mm high cylinders are cut from the solid thus obtained obtained and they are subjected to leaching test as described in the example 3. The leaching speed values are in the order of 1.multidot.10.sup.-2 g cm.sup.-2 days.sup.-1. The amount by weight of Na.sup.+ ion extracted from the test specimen is in the order of 10% after 72 hours. EXAMPLE 20 An emulsion prepared as described in example 1 wherein, however, radionuclides as tracers with the following activities: Co.sup.58 =280 .mu.Ci; Cs.sup.137 =212 .mu.Ci Sr.sup.85 =593 .mu.Ci, are added, is cast before it polymerizes into a polyethylene container having the dimensions: diameter of 50 mm and height of 55 mm. It is allowed to polymerize and a specimen having diameter of 50 mm and height 50 mm is cut. The test specimen thus obtained is subjected to a long term leach test, as specified hereinafter. In order to quantify the leaching resistance, the release of radioisotopes is calculated with a quantity called "leaching rate" the value whereof being in the present case 1.5.multidot.10.sup.-5 cm days.sup.-1 for Co.sup.58, 5.1.multidot.10.sup.-5 cm days.sup.-1 for Cs.sup.137 and 4.9.multidot.10.sup.-5 cm days.sup.-1 for Sr.sup.85. The method used is the following: a test specimen of the piece containing radioactive wastes solidified according to the method described in the present application, said test specimen being constituted by a block of material having a cylindrical shape, the total geometrical surface whereof being between 10 and 200 cm.sup.2, is placed in a container of polytetrafluorethylene or polypropylene in such a way that the test specimen is suspended by means of wires covered with one of the aforesaid materials so as not to touch the surface of the container. The container is filled with deionized water in order to completely cover the test specimen, this latter being surrounded at each point by a layer having at least 1 cm thickness of deionized water. The dimensions of the vessel and the amount of deionized water should be chosen in such a way that the value of the ratio of the deionized water to the area of the total geometric surface of the test specimen be comprised in the range from 0.08 to 0.12 m. The container is sealed and kept for 300 days at 23.degree. C..+-.1.degree. C. in such a way that the water does not undergo any mechanical stirring. The deionized water, after a certain contact time is substituted with fresh deionized water with the following frequency: once a day for the first seven days, twice a week for the second week, once a week for the third, fourth, fifth, and sixth week and thereafter once a month for the remaining time period up to 300th day. The individual amounts of water used are collected and the pH, the sodium ion, the Co.sup.58, the Cs.sup.137 and Sr.sup.85 are determined by the usual analitical chemistry and radioactivity methods. The results of the leach test should be expressed for each component by the leaching rate R.sub.n.sup.i, EQU R.sub.n.sup.i =a.sub.n.sup.i /(A.sub.o.sup.i .multidot.F.multidot.t.sub.n .multidot..rho.), wherein: R.sub.n.sup.1 =leaching rate in m/s of the i-th component, during the n-th leaching period; a.sub.n.sup.i =radioactivity in s.sup.-1 or mass in kg leached during the n-th leaching period, of the i-th component leached; A.sub.o.sup.i =specific radioactivity in s.sup.-1 .multidot.kg.sup.-1 or concentration by weight initially present in the test specimen; F=exposed surface of the test specimen in m.sup.2 ; t.sub.n =time of the n-th leaching period in s; .rho.=mass by unit volume of the test piece in kg/m.sup.3. A.sub.o.sup.i and a.sub.n.sup.i should be corrected by taking into account the decay time of the radionuclide considered. For the most suitably solidified material the leaching rate R.sub.n becomes constant after a certain amount of continuous renewals of leaching solution, as it is seen from the diagram of the variation of R.sub.n as a function of time. Said value, virtually constant, should be indicated together with its accuracy. EXAMPLE 21 An emulsion prepared as described in example 19, wherein however radionuclides as tracers with following activities: Co.sup.58 =270 .mu.Ci Cs.sup.137 =201 .mu.Ci Sr.sup.85 =490 .mu.Ci PA0 Co.sup.58 =3.1.multidot.10.sup.-6 cm days.sup.-1 ; Cs.sup.137 =4.9.multidot.10.sup.-7 cm days.sup.-1 ; Sr.sup.85 =7.multidot.10.sup.-7 cm days.sup.-1. PA0 Co.sup.58 =166 .mu.Ci Cs.sup.137 =210 .mu.Ci Sr.sup.85 =490 .mu.Ci PA0 Co.sup.58 =2.7.multidot.10.sup.-5 cm days.sup.-1 ; Cs.sup.137 =2.6.multidot.10.sup.-4 cm days.sup.-1 ; Sr.sup.85 =4.9.multidot.10.sup.-6 cm days.sup.-1. are added, is cast into a polyethylene container. The specimens prepared as described in example 20 are subjected to leach tests in the way described in example 2, whereby the following values are obtained: EXAMPLE 22 An emulsion, prepared as described in the example 2, wherein, however, radionuclides as tracers with the following activities: are included in the solution (D), is cast into polyethylene containers. The specimens prepared as described in the example 20 are subjected to leach tests in the way described in the example 20, whereby the following values are obtained:
	summary
063317125	abstract	A focused ion beam apparatus having an ion source, a focusing optical system and a scanning electrode scans the focused ion beam across a desired region of a sample surface to form a cross-section in which stacked conductors separated by an insulating film are exposed. To prevent charge-up of the sample due to an electrically floating nature of a conductor, a thin hole is formed using the focused ion beam to extend from one conductor to another. Etched particles are adhered to a side surface of the hole due to thin hole formation, with the result that a conductive film is formed electrically connecting the conductors. A floating conductive film is put into contact with a non-floating film to thereby avoid charge-up during observation of the sample with a charged particle beam. A secondary charged particle detector detects secondary charged particles generated in response to ion beam irradiation and outputs a corresponding signal, and a display unit displays an image of the sample based on the output signal of the secondary charged particle detector.
044951426	summary	The present invention relates to a system for monitoring the state of a nuclear reactor core and, more particularly, to a monitor system for monitoring the state or condition of a nuclear reactor core, employing a radiation monitor suitable for monitoring the state of a nuclear reactor core at the time of loss of coolant accident (referred to as LOCA, hereinafter). The boiling water reactor has a reactor containment vessel consisting of a dry well and a pressure suppression chamber. The reactor pressure vessel is placed in the dry well, while the pressure suppression chamber is filled with cooling water. In the event of a LOCA in the boiling water reactor, the cooling water of high pressure and temperature in the pressure vessel is dischaged as steam into the dry well through the fractured portion of the primary loop recirculation system. This steam is introduced into the pressure suppression chamber and is condensed by the cooling water in the latter. As disclosed in Japanese Patent Laid-open Publication No. 65295/76, a radiation meter, thermometer, pressure gauge and a condensate drain level meter are installed in the dry well of the boiling water reactor, in order to monitor the leakage of the cooling water due to LOCA. This monitoring system permits the operator to confirm occurrence of LOCA, but cannot provide accurate information concerning the state or condition of fuel rods in the reactor core after occurrence of LOCA. Therefore, with this known monitor system, it is not possible to take suitable countermeasures after the occurrence of the LOCA. SUMMARY OF THE INVENTION It is, therefore, a major object of the invention to accurately determine the state of fuel rods after occurrence of LOCA. To this end, according to the invention, there is provided a monitoring system comprising: means for measuring a level of the radioactivity of iodine in the primary containment vessel of a nuclear reactor; means for measuring a level of the radioactivity of noble gas; and means for judging or determining the state or condition of fuel rods in the reactor core in the event of occurrence of LOCA, upon receipt of the output from the means for measuring the level of the radioactivity of iodine and the output from the means for measuring the level of the radioactivity of noble gas.
	description	This application claims the benefit of DE 10 2011 076 876.9, filed on Jun. 1, 2011. The present embodiments relate to an apparatus for irradiating patients with x-ray radiation. In medical technology, x-rays may be used for diagnosis and therapy. The design and mechanics of x-ray devices may be adjusted for the desired functions. Functions of this type are, for example, the conventional x-ray diagnostics with the recording of individual x-ray images, fluoroscopy, mammography and computed tomography. With fluoroscopy, contrary to conventional x-ray diagnostics, no static individual image is produced. Instead, dynamic processes in the body are made visible by brief snapshots or whole series of individual images such as, for example, swallowing or the movement of the esophagus. In fluoroscopic systems, a distinction may be made between under-couch devices and/or under-couch systems that position the x-ray tube below the patient couch, and over-couch systems, in which the x-ray tube is attached above the couch. The present embodiments may obviate one or more of the drawbacks or limitations in the related art. For example, patient access to x-ray apparatuses may be facilitated. With under-couch devices having x-ray emitters fixedly mounted below the patient support, the latitude available for lowering the patient support is limited. In order to achieve optimal illumination and separation from the image receiver, the x-ray emitter requires a specific distance. A floor-to-patient support distance that may not be downwardly changed may result therefrom. It may not be comfortable for the patient to climb onto the couch when the minimal couch height of approximately 80 to 90 cm is restricted by the fixed x-ray emitters. This applies to patients who are not in a good state of health. In accordance with the present embodiments, an apparatus that includes a lowerable patient support and an x-ray apparatus that may be positioned below the patient support is provided in order to irradiate patients with x-rays. Provision is made for the accommodation and/or lowering of at least one part of the x-ray apparatus in the patient support for the purpose of further lowering the patient support. This apparatus is an under-couch device, for example. The ability to lower part of the x-ray apparatus in the patient support may reduce the minimal patient support height. The patient is thus helped with climbing onto the patient support, so that less assistance has to be provided by hospital personnel. Any climbing aids provided in conventional devices may be dispensed with. The x-ray apparatus is an apparatus for generating x-rays. The x-ray apparatus may be formed with an x-ray emitter and an aperture housing. According to one embodiment, part of the aperture housing is accommodated in the patient support when the patient support is being lowered. In one embodiment, a mechanism that establishes a releasable contact between the patient support and the x-ray apparatus such that the x-ray apparatus is also subject to a vertical change in position of the patient support is provided. The releasable contact is disengaged for the lowering of the at least one part of the patient apparatus in the patient support such that the patient support and the x-ray apparatus may be displaced relative to one another. Contact may be established in a first phase of the lowering, and the patient support may be lowered together with the x-ray apparatus. A maximum value of the possible shared mutual lowering is achieved during the first phase, whereupon the contact is released, and a further lowering of the patient support is connected with a lowering and/or displacement of part of the x-ray apparatus in the patient support. The release of the contact between the patient support and the x-ray apparatus may be triggered by a force counteracting the lowering. This force is conveyed, for example, by the x-ray emitter landing on the floor. A sensor system may be provided, for example, connecting a light barrier that detects a threshold value for the height of the patient support and triggers a release of the contact. The lowering with existing contacts may be braked in a final phase. Both the release of contact triggered by a sensor and also the braking may be used to reduce any force effects (e.g., by the floor) when reaching the final position of a first phase of the lowering process in order to prevent damage to the x-ray apparatuses. The second phase of lowering includes displacing part of the x-ray apparatus in the patient support. In one embodiment, the mechanism may be configured to manufacture the contact in a defined relative position of the patient support and the x-ray apparatus. This is realized, for example, by a mechanism that includes a detent and a detent lever. In this embodiment, the detent lever is pressed out of the detent during the course of the release of the contact, and the contact is established by a renewed engagement. The function of reestablishing the contact is meaningful with respect to raising the patient support in order to achieve a recording position for the x-ray recording. In one embodiment, a safety system may be provided. The safety system detects whether the contact is established. The x-ray direction is blocked for x-ray recordings by the safety system, provided there is no contact. With a released contact and/or partially inserted x-ray apparatus, the optimal distances between the x-ray emitter and the patient for an x-ray recording are not present. An intentional triggering of an x-ray recording may be explicitly prevented. The raising or lowering of the x-ray emitter may be realized by a motor that is arranged laterally to the aperture housing, for example. “Laterally” may be an area of the entire plane at right angles to the beam direction (e.g., an arrangement behind the aperture housing). FIG. 1 shows an apparatus 3 for generating x-rays. The apparatus 3 includes an x-ray emitter 1 and an aperture housing 2. The x-ray emitter 1, from which focus F x-rays emanate, includes a rough pre-aperture 5 and a beam window 4. A pre-aperture 6 provided with the slots and elements 7 and 8 for determining a position of the focus F and an actual main aperture 10 for forming a useful x-ray beam bundle are integrated into the aperture housing 2. In order to protect against radiation, a scattered radiation seal 9 also exists between the x-ray emitter 1 and the aperture housing 2. With an under-couch device, the x-ray apparatus 3 radiates upwards (e.g., a position is rotated by 180 degrees with respect to FIG. 1). The x-ray apparatus 3 exhibits an expansion in the beam direction. The aperture structure is responsible for a minimum distance of the focus F from a patient being provided during operation. In conventional under-couch devices, this distance limits the possible lowering depth of the patient support and/or patient couch. As a result, access for the patient is made more difficult. The patient may climb onto a couch with a couch height of approximately 80 to 90 cm. This is difficult for many patients with respect to the size and/or state of health of the patient. To remedy this difficulty, present embodiments are explained in more detail below with the aid of FIG. 2. FIG. 2 shows one embodiment of an under-couch device 20 in two different positions. A position, in which recordings take place, is shown on the left side. The x-ray emitter 1 is a certain distance from the floor and is arranged for fluoroscopy with respect to the patient support 11. In order to enable the patient to climb onto the patient support 11, the patient support 11 is lowered as shown in the image to the right. The x-ray emitter 2 is lowered to the floor. The x-ray apparatus includes an area 19 that is accommodated when lowering the patient support 11 in a lower region 111 of the patient support 11. The patient support 11 may therefore be lowered further than with conventional systems by about the thickness D of the lower area 111. The patient support 11 is thus more easily accessible. FIG. 3 shows a mechanism used for the partial lowering of the x-ray apparatus. The x-ray tube 1 and the aperture housing 2 form an x-ray apparatus that may be lowered for the patient to climb on. An engaging lever 13, a detent 12 (e.g., with an opening of approximately 30 mm) and a linear guide 14 are shown in FIG. 3. The engaging lever 13, the detent 12, and the linear guide 14 are moved together with the x-ray apparatus 3. The engaging lever 13 includes a ball bearing attached below. The engaging lever 13 has a diameter of 30 mm, for example. The latching lever 13 is also provided with a spring support 23 and is moveably mounted (e.g., pivotable about axis 24) so that the spring support 23 presses the lever 13 into the detent 12 in an engaged position. The detent 12 is a stationary component with respect to the patient support 11. The mechanism is described with the aid of FIG. 4. The x-ray apparatus with the engaging mechanism 21 is visible in the center of FIG. 4. If the emitter is in the engaged position, the x-ray apparatus is in a normal working position. As indicated above right with respect to the entire system 20, the patient support is moved downwards for patient access (e.g., arrow 22). The emitter 1 rests on the floor when the couch moves downwards. Continuing the downward movement provides that the force of the spring on the detent lever 13 is no longer adequate, and the lever is pressed out of the detent 12. At a lower end of the lever is a ball bearing (not shown) that enables the leverage movement. The unit including emitter 1, engaging lever 12, linear guide 14 and spring support may therefore be moved upwards guided by the linear guide 14. If the table is moved upwards again after positioning the patient, the emitter unit reengages on account of the dead weight. No displacement of the emitter outside of a zero degree position arises. FIG. 5 provides information relating to the space available and/or traveling distance. FIG. 5 shows the emitter 1 in the engaged position (e.g., extended to a maximum). Reference character 17 indicates the maximum upward travelling distance. The maximum upward travelling distance ends just below the detector loader 15, which may traverse the couch from A to B. FIG. 6 shows a side view of the overall system, in which protection from x-ray radiation 18 is provided. FIG. 7 and FIG. 8 indicate possible positions for the motor-driven drive of the tubes. This drive is designated with reference character 19. Different variants of a motor may be provided (e.g., a space-saving drum motor or a normal electric motor). The type of drive may be a spindle drive or a rack drive. The arrangement of the motor may be lateral to the aperture (e.g., on the rear (FIG. 7) or adjacent to the aperture (FIG. 8) when viewed from the front). FIG. 9 shows the entire system from the front having a possible motor position of the motor 9. FIG. 10 is a front view of the x-ray apparatus, and FIG. 11 is a side view of the entire under-couch system. The apparatus is not restricted to the embodiments illustrated. For example, other mechanisms than a purely mechanical one may be provided for engagement. Developments may include, for example, a sensor system that detects positions of the couch and triggers an unlocking mechanism using a control signal. The forces developing in the mechanical embodiment illustrated may, for example, be reduced when coming into contact with the floor. In one embodiment, a safety system that prevents the x-ray tube from triggering if no unlocking and/or no engagement and thus no recording position of the x-ray emitter exists, may be applied. While the present invention has been described above by reference to various embodiments, it should be understood that many changes and modifications can be made to the described embodiments. It is therefore intended that the foregoing description be regarded as illustrative rather than limiting, and that it be understood that all equivalents and/or combinations of embodiments are intended to be included in this description.
	description	This application claims the benefit of U.S. Provisional Patent Application No. 60/620,304 filed by Manoj Prasad, Neal J. Snyderman, and Mark S. Rowland Oct. 19, 2004 and titled “Absolute Nuclear Material Assay.” U.S. Provisional Patent Application No. 60/620,304 is incorporated herein by this reference. The United States Government has rights in this invention pursuant to Contract No. W-7405-ENG-48 between the United States Department of Energy and the University of California for the operation of Lawrence Livermore National Laboratory. 1. Field of Endeavor The present invention relates to nuclear material assay and more particularly to an absolute nuclear material assay. 2. State of Technology United States Patent Application No. 2005/0105665 by Lee Grodzins and Peter Rothschild for a system of detection of neutrons and sources of radioactive material, published May 19, 2005, provides the following state of technology information: “There is a need to find sources of radiation and other nuclear material that are clandestinely transported across national boundaries. The sources of clandestine nuclear material may be in the form of “dirty bombs” (e.g., a conventional explosive combined with radioactive nuclides designed to spread radioactive contamination upon detonation), fissile material, and other neutron and radiation emitting sources that may present a hazard to the public. During recent years, the United States government has placed mobile vehicles at strategic areas with gamma ray detectors dedicated to the task of finding fissile material. Atomic explosives may be made from 235U, a rare, naturally occurring, isotope of uranium that lives almost 109 years, or 239Pu, a reactor-made isotope that lives more than 104 years. 235U decays with the emission of gamma ray photons (also referred to as ‘gammas’), principally at 185.6 keV and 205. 3 keV. 239Pu emits a number of gamma rays when it decays, the principal ones being at 375 keV and 413.7 keV. These gamma rays are unique signatures for the respective isotopes. But fissile material invariably contains other radioactive isotopes besides those essential for nuclear explosives. For example, weapons grade uranium may contain as little as 20% 235U; the rest of the uranium consists of other isotopes. The other uranium and plutonium isotopes reveal their presence by gamma rays emitted by their daughters. For example, a daughter of 238U emits a high energy gamma ray at 1,001 keV; a daughter of 232U, an isotope present in fissile material made in the former USSR, emits a very penetrating gamma ray at 2,614 keV; and a daughter of 241Pu emits gamma rays of 662.4 keV and 722.5 keV.” U.S. Pat. No. 4,201,912 issued May 6, 1980 to Michael L. Evans et al and assigned to The United States of America as represented by the United States Department of Energy, provides the following state of technology information: “A device for detecting fissionable material such as uranium in low concentrations by interrogating with photoneutrons at energy levels below 500 keV, and typically about 26 keV. Induced fast neutrons having energies above 500 keV by the interrogated fissionable material are detected by a liquid scintillator or recoil proportional counter which is sensitive to the induced fast neutrons. Since the induced fast neutrons are proportional to the concentration of fissionable material, detection of induced fast neutrons indicates concentration of the fissionable material.” U.S. Pat. No. 4,617,466 issued Oct. 14, 1986 to Howard O. Menlove and James E. Stewart and assigned to The United States of America as represented by the United States Department of Energy, provides the following state of technology information: “Apparatus and method for the direct, nondestructive evaluation of the .sup.235 U nuclide content of samples containing UF.sub.6, UF.sub.4, or UO.sub.2 utilizing the passive neutron self-interrogation of the sample resulting from the intrinsic production of neutrons therein. The ratio of the emitted neutron coincidence count rate to the total emitted neutron count rate is determined and yields a measure of the bulk fissile mass. The accuracy of the method is 6.8% (1.sigma.) for cylinders containing UF.sub.6 with enrichments ranging from 6% to 98% with measurement times varying from 3-6 min. The samples contained from below 1 kg to greater than 16 kg. Since the subject invention relies on fast neutron self-interrogation, complete sampling of the UF.sub.6 takes place, reducing difficulties arising from inhomogeneity of the sample which adversely affects other assay procedures.” U.S. Pat. No. 3,456,113 issued Jul. 15, 1969 to G. Robert Keepin provides the following state of technology information: “An apparatus and method of detecting, identifying and quantitatively analyzing the individual isotopes in unknown mixtures of fissionable materials. A neutron source irradiates the unknown mixture and the kinetic behavior of the delayed neutron activity from the system is analyzed with a neutron detector and time analyzer. From the known delayed neutron response of the individual fission species it is possible to determine the composition of the unknown mixture. Analysis of the kinetic response may be accomplished by a simple on-line computer enabling direct readout of isotopic assay.” Features and advantages of the present invention will become apparent from the following description. Applicants are providing this description, which includes drawings and examples of specific embodiments, to give a broad representation of the invention. Various changes and modifications within the spirit and scope of the invention will become apparent to those skilled in the art from this description and by practice of the invention. The scope of the invention is not intended to be limited to the particular forms disclosed and the invention covers all modifications, equivalents, and alternatives falling within the spirit and scope of the invention as defined by the claims. The present invention provides a system of absolute nuclear material assay of an unknown source. The present invention provides a system that relates, in detail, a correlated or uncorrelated chain of neutrons with what appears in an instrument (i.e., relates physical parameter to a measured quantity). How the chain of neutrons is used was traditionally related in a process that connects count sums to physical parameters of interest, such as multiplication. The limitations of prior art start with and are rooted in approximations in the detailed description of the neutron chain. These approximations, in the details of exactly how a chain is described and evolves in time, conspire to make the process of relating chains to physical parameters highly unstable. Prior art therefore relies on a process of calibration. For example, calibration means that of the five parameters needed to describe a physical system, four are determined independent of an assay measurement. The old assay process then proceeds by assuming the four parameters apply and are considered with a measurement of the fifth parameter, to be extracted from the assay measurement. In the present invention, the assay solution comes from the solution of a coupled set of equations where all five parameters are used to solve for a physical parameter of interest, such as multiplication. The present invention benefits from a complete understanding of an arbitrary chain and variously allows the extraction all five parameters, or four parameters given only one, or three parameters given only two, etc. In the present invention neutrons are measured in a neutron detector and five parameters are determine (mass, multiplication, alpha ratio, efficiency, and time constant) that describe the object that is being assayed. The present invention makes an assay for the purpose of determining these five parameters, given that one does not know these five parameters. A neutron is created by a physical process, either fission or an inducing nuclear reaction. The created neutron or neutrons then interact with the environment. If the environment contains more nuclear material (i.e., uranium), the first neutrons may create more neutrons by causing more fission or other nuclear reactions. The first and second and subsequent neutrons are the chain. A chain may start with an alpha particle creating a single neutron that subsequently creates hundreds of fissions. Another chain may start with a spontaneous fission creating three neutrons that go on to create hundreds of fissions. These chains evolve over time and some of the neutrons are absorbed or lost. Finally, some members of the chain are captured in a detector. The final captured neutrons may be counted as a simple sum or observed as a time dependent rate. What may start out as a chain of 1000 neutrons may result in a count of two neutrons during some snippet of time, in a detector. The specific numerical process of relating the relevant physical parameters (mass, multiplication, alpha ratio, efficiency, and time constant) to an observed quantity (how many 2's) is based on approximations in the prior art. Describing these chains, with all the numerical detail requires a way to relate the five physical parameters to how the chains are created. The present invention provides a method of absolute nuclear material assay of an unknown source comprising counted neutrons from the unknown source and uses a model (theory) to optimally fit the measured count distribution. The present invention begins by analytically solving for and efficiently computing the entire fission chain probability distribution for any given set of physical parameters (mass, multiplication, alpha ratio, efficiency, and time constant). This fission chain distribution is then used to simulate a data stream from which time dependent count distributions are constructed. The model randomly initiates fission chains at a rate dependent on the measured source strength and samples from the analytical fission chain probability distributions to artificially create data with statistical fluctuations with finite time counting. This approach allows the most direct modeling of the data as it is actually taken. It also allows complete control in modeling issues related to finite sampling, truncation errors from inherently truncated data, and dead time effects in the detector. Previous art could only compute the first few moments of the full idealized fission chain distribution and relate these to moments of measured data. The previous art is fundamentally flawed in modeling finite sample truncated data with idealized infinite population moments. This flaw manifests itself in an erratic and unstable reconstruction of the unknown physical parameters. The approach of the present invention, based on analytical fission chain probability distribution, is able to robustly and stably reconstruct physical parameters. A far more reaching significance of the present invention is that it provides a complete theoretical framework for modeling the entire neutron count distribution, not just its first few moments. Any measured count distribution and its model made with the five, or even more parameters, may be quantitatively compared for the purpose of optimally reverse engineering the 5 or more parameters that describe the unknown. Previous art based on the first few moments can only get at some small subset of the information contained in the data, and even then is flawed by issues of finite sample size and truncation errors. (Other parameters include, but are not limited to background contributions, external sources adding counts, (n,2n) neutron sources . . . ) The present invention provides a method of absolute nuclear material assay of an unknown source comprising counting neutrons from the unknown source and providing an absolute nuclear material assay utilizing a sampling method to distribute theoretical count distributions over time. The method utilizing a random sampling of a count distribution to generate a continuous time-evolving sequence of event-counts by spreading the count distribution in time. The present invention also provides an apparatus for absolute nuclear material assay comprising a multigate neutron multiplicity counter, a processor that solves three moment equations, a processor that provides fit to actual time dependence of the moments to get proper asymptotic moments, a processor that uses the estimated parameters to compute full count distribution, a processor that compares truncated data moments with untruncated and truncated theoretical moments, and a processor that provides adjustments to reduce bias. The present invention has use in providing an assay of nuclear material. The present invention also has uses in providing the amount of moderator and in providing a neutron lifetime. The present invention can be used to providing an operator a simple system for obtaining the mass, multiplication, detector efficiency, and the alpha-decay-created neutron rate. The invention is susceptible to modifications and alternative forms. Specific embodiments are shown by way of example. It is to be understood that the invention is not limited to the particular forms disclosed. The invention covers all modifications, equivalents, and alternatives falling within the spirit and scope of the invention as defined by the claims. Referring to the drawings, to the following detailed description, and to incorporated materials, detailed information about the invention is provided including the description of specific embodiments. The detailed description serves to explain the principles of the invention. The invention is susceptible to modifications and alternative forms. The invention is not limited to the particular forms disclosed. The invention covers all modifications, equivalents, and alternatives falling within the spirit and scope of the invention as defined by the claims. Fission is defined as the emission of multiple neutrons after an unstable nucleus disintegrates. For example, Pu240 decays at a rate of about 400 fissions per second per gram of Pu240 atoms. When the fission occurs, multiple neutrons are emitted simultaneously, with the number ranging from zero to eight neutrons. The present invention provides a system that utilizes a set of parameters that describe an unknown mass of fissile material. This simultaneous neutron emission characteristic is unique to fission. The present invention provides a system that utilizes a multiplicity counter and a neutron detector that is set up to see time grouped neutrons. The present invention has use in providing an operator a simple system for obtaining the mass, multiplication, detector efficiency, and the alpha-decay-created neutron rate. The characteristic of fission is that neutrons emit in groups. Random sources of neutrons are emitted with no regard for grouping, however, since the appearance of these neutrons at the detector are randomly spread in time, some may accidentally appear in close temporal proximity. An example is a neutron detector that counts neutrons for short periods of time, say ½ milli-second. This example time corresponds to a typical neutron diffusion time in a typical detector, the choice of which depends on the detector design and is not the subject here. If the ½ msec. period is counted once, the count may be three counts, or some other integer number, including zero. It is desirable to select an appropriate observation time, two to three times the typical neutron diffusion time, and then repeat the sampling of counts period many times to produce a histogram of counts described as the number of occurrences of each multiplet group (i.e., number of times 0, 1, 2, 3 . . . were observed, in sum, over say 10,000 repeated detection periods). Fission is unique in that it creates real correlations, while non-fission neutron sources create accidental correlations. The present invention provides a system that utilizes new developments in how fission neutron chains are modeled to simplify and remove problems related to the assay of unknown packages of fissioning material. In general, the present invention provides a system that describes the evolution of fission chains with enough detail that universal procedures can be defined for absolute assay. The absolute assay does not need pre-defined facts or assumptions about the neutron detector efficiency (e), neutron lifetime (L), instrumentation dead-time losses (D), the terrestrial background (B), or the fraction of alpha-decay-induced neutrons (A) while endeavoring to obtain neutron multiplication (M) and mass of fissioning material (m). Counting neutrons by looking for time-correlated groupings is called multiplicity counting. The groupings arise from the fission process where a portion of a fission chain is detected. The analysis of this type of data assists in deriving mass, multiplication, detector efficiency, and alpha ratio (mMeA). Other factors in the analysis include neutron lifetime (L), measurement gate width (T), the maximum size of neutron multiplets observed (n), the background correlation and count rate (B), and the generalized Poisson exponent Λ (Λ). Traditionally, the count rate (singles) and the number of doubles are used to solve for up to two of the parameters, unfortunately with a significant dependence on quantitative knowledge of the other parameters. Measurement of the number of singles and doubles is limited additionally because of the necessity of incomplete sampling of the fission chains (since no one can count for an infinite time). Prior approaches assume a complete sampling of the fission chain. The present invention provides a system that utilizes a process where the partial and full fission chain details are calculated exactly and are used to correctly interpret the measurements. The present invention provides a system that utilizes allows solving for all of the unknown parameters listed above. The premise for (definition) multiplication is that all neutrons in the fission chains are accounted for in the definition of nubar and multiplication (M). Nubar of the fission chain (N) and M must relate exactly (probability of fission=p) M=1/(1−pN). The first moment of the induced fission chain, started from one neutron, is (1−p)M and is what is intended to be measured. In practice the first moment is not actually measured because the populations of neutrons are always sampled incompletely. M is the multiplication defined for the full population. Measurement gives an incomplete sampling of the population and is always biased (incorrect) because of the finite sampling time. When the measured samples are biased, they no longer relate properly to the M derivation, therefore M is usually derived only approximately. The incomplete sampling problem applies to higher moments of the fission chain. These errors propagate to the other derived unknowns, regardless of how many moments are used in an analysis. Other errors arise from mistakes in understanding the matrix of unknown source containers (e.g., errors in L, A, e, and B). The present invention provides a system that utilizes measurements made with a multi-gate neutron multiplicity counter. A fit to the actual time dependence of the moments is used to get the proper asymptotic moments and dead-time losses inherent in the data. Since H-C2 inversion leads to estimates that are biased (wrong) because of the finite sampling problem and dead-time, there are two paths to solve for the rest of the parameters. One is to use the Prasad theory to compute libraries of count distributions that may be used as a lookup table and the other is to use the H-C style estimated parameters to compute the full count distribution that would have been measured if there was no finite sampling error. The present invention provides a system that compares the truncated data moments (measurement) with untruncated and truncated theoretical moments. The present invention provides a system that utilizes extending the moments approach to more unknowns. Also, using moments is the same as using only part of the measured data, in contrast to actually fitting the measured count distribution to a library of count distributions (theory). The present invention provides a system that utilizes furthering the field by fitting the measurement to theoretically calculated count distributions to find the optimal set of parameters that would explain the count distribution. Fitting the full count distribution is a better approach because it uses all the information in the count distribution. The present invention provides a fitting approach that can extract all unknowns, in contrast to the prior situation of deriving at most three unknowns from three moments. The present invention provides a system that extends the H-C approach by adding a new method for dead-time correction most noteworthy for high multiplication, allows for truncation corrections, and allows direct comparison of data to parameter-based (mMea) count distributions that are generated as a proof test. The present invention provides a system that utilizes several new steps, not all required depending on analysis objectives or measurement uncertainties. One is to create a fitting algorithm that preferentially weights the longer T gates in a fitting analysis so the short mode effects minimally alter the resulting assmyptote. This is called a “T-cut” approach that prefers to extract the fundamental mode. Another method is to observe dead-time effects as a function of T, by simulation with Applicants new count distribution calculation method. This results in multi-mode time dependences that may be specified to the data fitting process, so that D may be extracted. With specific time-dependence specification and understanding, the fitting routine is stable as the only free parameter is D. Another method is to specify the time dependence in terms of the fission chain topology. This results in two modes for the second moment time dependence, and three modes for the third moment time dependence. By specifying these constrained sets of time dependences, the fitting routine will be stable as the only free parameter is the assmyptote and Lshort and Llong. The present invention provides a system that utilizes computing the exact fission chain time evolution and count distribution as a function of M, m, e A, L, T, A, D, and B so that Applicants can simulate measurements. Regarding dead-time (D), a precursor to using count distributions for assay requires a method to add the dead-time. The present invention provides a system that utilizes distributed theoretical count distributions over time (i.e., time-tagging the count events as they would have been seen during a measurement). This is different from using a monte-carlo transport technique because such a technique can not sample rare events thoroughly enough. The Prasad count distribution generation technique completely fills in all rare events exactly so it can be sampled with uniform weight to form an accurate time-tagged stream of synthetic data. The present invention provides a system that utilizes random sampling of a count distribution to generate a continuous time-evolving sequence of event-counts spreads the count distribution in time, as it would be seen during the measurement. This is done by randomly initiating fission chains at a rate dependent on the source strength and sampling from Applicants analytical theory of fission chain probability distributions to artificially create a stream of realistic data. The final step is to alter the time-tagged data with “coincidence-sum limits” to create dead time in time-tagged data or summed-count distributions. “Coincidence-sum limits” are the removal of selected time-tagged counts based on their being located within a D seconds to another count. Here D would be called the dead-time. The present invention provides a system that utilizes dealing with dead time when using H-C style moments based analysis. Similar to the process of generating a count distribution, the impact of dead-time is a non-linear process at the core of the count distribution generating function. Having identified the impact of dead time on count distributions, the present invention provides a system that parameterized these effects in the form of corrections to the moments.First moment: Dcr=Tcrexp(−DTcr−DLTr2f).Second moment: Dr2f=Tr2fexp(−D[3Tcr−LTr2f+{2LTr3f/Tr2f}].Third moment: Dr3f=Tr3fexp[−D[5Tcr−LTr2f+{(2TcrTr2f2+3LTr4f)/Tr3f}]Term Definition:Dcr,, Dr2f, and Dr3f are the dead-time reduced count rate, second moment and third moments.Tcr, Tr2f, and Tr3f are the true, no-dead-time count rate, second moment, and third moments. The process to correct moment-based dead time is to use dead-time afflicted count distributions (Applicants theory or measurements) to observe (fit) the perturbation in time dependence. Time dependences created by this method may be used to fit observed measured data to infer the amount of dead-time D. Then one may sequentially compute corrections to the moments starting with the count rate: Dcr=Tcrexp(−DTcr−DLTr2f). Note the first iteration uses the observed data r2f. Then use Dr2f=Tr2fexp(−D[3Tcr−LTr2f+{2LTr3f/Tr2f}]. This next step uses the observed data r3f. Next, compute Dr3f=Tr3fexp[−D[5Tcr−LTr2f+{(2TcrTr2f2+3LTr4f)/Tr3f}]. Note this last step uses r4f which Applicants set equal to zero the first time through this process. Then one solves the three equations for the three unknowns. Now Applicants have the first estimate of Tcr, Tr2f, and Tr3f. Now Applicants feed them to the H-C algebra to get an estimate of mMeA. Next Applicants compute what Tr4f would be if the H-C algebra were correct. Then Applicants repeat the process started with the count rate data, now using the estimated Tr4f. Iteration continues until Tcr, Tr2f, Tr3f don't change from one iteration to the next. The final feed of Tcr, Tr2f, Tr3f into the H-C theory results in the true mMeA. The present invention provides a system that includes the effects of background. Background comes from cosmic ray interactions in the detector, surrounding structures, the unknowns' non-fissile mass, or fissioning uranium in terrestrial material. The basic idea is to use the generating function to reverse engineer the Λ's in background. The present invention provides a system that measures background with one of Applicants counters, in the presence of large masses of iron, lead, and polyethylene. Specifically, the process is to compute the natural log of the background count distribution generating function and solve for the Λ's. The present invention provides a system that utilizes the background as a free parameter in generating data to develop specific understanding, or to partition an unknown measurement into the fraction of background present at measurement time. This approach is technically superior since fission chains are created from the non-linear process and not simply additive environmental fissioning mass. The present invention provides a system that utilizes hundreds of time dependent gates T, is that a table of T versus L may be measured and used as a lookup to characterize the general state of moderation in an unknown object. The general method allows one to estimate the mass of hydrogenous moderator mixed with fissioning material. This knowledge is useful for waste barrels where hydrocarbons in the presence of alpha-emitting fissile material tend to liberate hazardous gases. The present invention provides a system that utilizes data visualization techniques that give insight into the physics and the impact of statistical fluctuations on derived quantities. The present invention comprises the steps of counting neutrons from the unknown source and providing an absolute nuclear material assay. In one embodiment the step of providing an absolute nuclear material assay comprises utilizing a sampling method to distribute theoretical count distributions over time. In one embodiment the step of providing an absolute nuclear material assay comprises utilizing a random sampling of a count distribution to generate a continuous time-evolving sequence of event-counts by spreading the count distribution in time. In one embodiment the step of providing an absolute nuclear material assay comprises altering time tagged data with “coincidence-sum limits” to create dead-time in time-tagged data or summed-count distributions. In one embodiment the step of providing an absolute nuclear material assay comprises observing fine resolution of T axis data to obtain modal structure. In one embodiment the step of providing an absolute nuclear material assay comprises H-C Point-model extension by using constrained sums of T dependence, to select best L to fit the data which includes T-cut approach to get long-mode asymptotes, multiple mode sums to get asymptotes, and single mode fits to see deviations from single mode behavior. In one embodiment the step of providing an absolute nuclear material assay comprises H-C Point-model extension by using constrained sums of T dependence, to select best L to fit the data which includes T-cut approach to get long-mode asymptotes, multiple mode sums to get asymptotes, and single mode fits to see deviations from single mode behavior and subsequently, use the best fit parameters from the model for analysis. In one embodiment the step of providing an absolute nuclear material assay comprises dead-time correction based on T dependence perturbations/shifts. In one embodiment the step of providing an absolute nuclear material assay comprises using L to estimate moderator mass around the fissioning material. In one embodiment the step of providing an absolute nuclear material assay comprises precomputing lookup tables of real-time computed count distributions for comparison to measured data. Referring to FIG. 1, one embodiment of a system of the present invention is illustrated. This embodiment of the system is designated generally by the reference numeral 100. The system 100 comprises a number of interconnected structural components. The structural components include a multigate neutron multiplicity counter 101, a processor that computes the time dependent moments 102, a processor that provides fits to deadtime, lifetime, biases, and allows the selection of the number of unknown parameters 103, a processor that solves for the unknown parameters 104, a processor that compares truncated data moments with untruncated and truncated theoretical moments 105, and a processor that checks for consistency and stability of solutions 106. The system 100 can be used to provide an assay of nuclear material and/or to provide the amount of moderator, neutron time constant, or other biases. Note that process 104 is described in tables 2 and 3, and process 105 depends on the process in table 1. Alternatively, count distributions may be generated from first principles. Table 3 includes a discussion and process ramp-up about BIGFIT. The present invention provides a system that relates, in detail, a correlated or uncorrelated chain of neutrons with what appears in an instrument (i.e., relates physical parameter to a measured quantity). How the chain of neutrons is used was traditionally related in a process that connects count sums to physical parameters of interest, such as multiplication. The limitations of prior art start with and are rooted in approximations in the detailed description of the neutron chain. These approximations, in the details of exactly how a chain is described and evolves in time, conspire to make the process of relating chains to physical parameters highly unstable. Prior art therefore relies on a process of calibration. For example, calibration means that of the five parameters needed to describe a physical system, four are determined independent of an assay measurement. The old assay process then proceeds by assuming the four parameters apply and are considered with a measurement of the fifth parameter, to be extracted from the assay measurement. In the present invention, the assay solution comes from the solution of a coupled set of equations where all five parameters are used to solve for a physical parameter of interest, such as multiplication. The present invention benefits from a complete understanding of an arbitrary chain and variously allows the extraction of all five parameters, or four parameters given only one, or three parameters given only two, etc. In the present invention neutrons are measured in a neutron detector and five parameters determine (mass, multiplication, alpha ratio, efficiency, and time constant) that describe the object that is being assayed. The present invention makes an assay for the purpose of determining these five parameters, given that one does not know these five parameters. A neutron is created by a physical process, either fission or an inducing nuclear reaction. The created neutron or neutrons then interact with the environment. If the environment contains more nuclear material (i.e., uranium), the first neutrons may create more neutrons by causing more fission or other nuclear reactions. The first and second and subsequent neutrons are the chain. A chain may start with an alpha particle creating a single neutron that subsequently creates hundreds of fissions. Another chain may start with a spontaneous fission creating three neutrons that go on to create hundreds of fissions. These chains evolve over time and some of the neutrons are absorbed or lost. Finally, some members of the chain are captured in a detector. The final captured neutrons may be counted as a simple sum or observed as a time dependent rate. What may start out as a chain of 1000 neutrons may result in a count of two neutrons during some snippet of time, in a detector. The specific numerical process of relating the relevant physical parameters (mass, multiplication, alpha ratio, efficiency, and time constant) to an observed quantity (how many 2's) is based on approximations in the prior art. Describing these chains, with all the numerical detail requires a way to relate the five physical parameters to how the chains are created. This procedure is summarized in Table 1 below. TABLE 1 x = ∫ 0 t ⁢ e - λ ⁡ ( t ′ - t f ) ⁢ λ ⁢ ⁢ ⅆ t ′ = e λt f ⁡ ( 1 - e - λt ) , y = ∫ t f t ⁢ e - λ ⁡ ( t ′ - t f ) ⁢ λ ⁢ ⁢ ⅆ t ′ = ( 1 - e - λ ⁡ ( t - t f ) ) , λ is lifetime. t is time ε is efficiency ⩓ f ⁢ = ⁢ { ∫ - ∞ 0 ⁢ [ ∑ v = f ∞ ⁢ P v ⁡ ( v j ) ⁢ ( εx ) j ⁢ ( 1 - εx ) v - j ] ⁢ F s ⁢ ⅆ t f + ⁢ ∫ 0 t ⁢ [ ∑ v = j ∞ ⁢ P v ⁡ ( v j ) ⁢ ( εy ) j ⁢ ( 1 - εy ) v - j ] ⁢ F s ⁢ ⅆ t f } . Fs is p/s (mass) Pv = f(M, v(snm)) A comes from a special case of a single neutron multiplyingFor example, the number of fives is: b 5 = ⁢ ( ⩓ 5 ⁢ + ⩓ 4 ⁢ ⩓ 1 ⁢ + ⩓ 3 ⁢ ⩓ 2 ⁢ + ⩓ 3 ⁢ ⩓ 1 2 2 ! + ⁢ ⩓ 2 2 2 ! ⁢ ⩓ 1 ⁢ + ⩓ 2 ⁢ ⩓ 1 3 3 ! + ⩓ 1 5 5 ! ) ⁢ exp ⁡ [ - ( ⩓ 1 ⁢ + ⩓ 2 ⁢ + … ) ] . Bn is the multiplet count in the measurement and is directly related to the five parameters with this calculation process. A multi-gate counter measures Bn as a function of lifetime and neutron number. Since degenerate use of the procedure of the present invention is possible, Applicants made a NMAC procedure. It is similar to the hage-Cifferelli moments approach, but the NMAC procedure extends that procedure by allowing solutions that may be truncated as all measurements are, allows detailed time dependent analysis to better understand time truncated measurements, allows for the inclusion of gamma-rays in the assay process, and allows for dead time correction for the second and higher moments. The NMAC procedure is summarized in Table 2 below. TABLE 2NMAC solves algebra solutions based on the first 3 moments.We always fit λ to determine neutron lifetime and therefore correct forasymptotic saturation.This leaves four unknowns to determine: m, M, A, ε.Case examples:Given one unknown and R1, R2, and R3, we solve for the remainingunknowns (e.g. Given A, we solve for m, M, and ε).Given two unknowns and R1 and R2, we solve for the remaining twounknowns (e.g. Given A and ε, we solve for m and M) A comparison of the NMAC procedure and the BigFit procedure is summarized in Table 3 below. TABLE 3Neutron Multiplicity Analysis Code (NMAC)Mass, Multiplication, Alpha, efficiency, Lambda are unknown.R2 = mass [ε2M2q2(D2s + M − 1(1 + A)D2] F(λt) and describes one of themoments of the count distribution, which is only a piece of the countdistribution information.NMAC solves algebra solutions based on the first 3 moments.We cannot know efficiency if we don't know the geometryWe cannot solve for five unknowns with three equations (e.g. y1, y2, y3)Higher moments algebra (y4, y5) depends too much on the tail, i.e. noisy.Algebra involves ratios of moments, where uncertainties in the momentscause large solution errors.BigFitAlternatively, count distributions may be generated from first principles.Count distributions are the complete realization of the fission chain,related to all of the measured physical parameters and therefore provideall the available information and therefore the most definitive connectionto the assay quantities that we want.As a process, template fitting searches for a match between an unknownmeasurement and a library of variations. It appears that a library of ~4,000variations and about 106 counts is sufficient to provide a goodmatch to the assay of the unknown. Referring to FIG. 2, an embodiment of a system utilizing the present invention is illustrated. This embodiment is designated generally by the reference numeral 200. The system 200 provides a system for absolute nuclear material assay of an unknown source. The system 200 comprises the steps of counting neutrons from the unknown source and providing an absolute nuclear material assay. In one embodiment, the step of providing an absolute nuclear material assay comprises utilizing a sampling method to distribute theoretical count distributions over time. In another embodiment, the step of providing an absolute nuclear material assay comprises utilizing a random sampling of a count distribution to generate a continuous time-evolving sequence of event-counts by spreading the count distribution in time. In another embodiment, the step of providing an absolute nuclear material assay comprises altering time tagged data with “coincidence-sum limits” to create dead-time in time-tagged data or summed-count distributions. In another embodiment, the step of providing an absolute nuclear material assay comprises observing fine resolution of T axis data to obtain modal structure. In another embodiment, the step of providing an absolute nuclear material assay comprises H-C Point-model extension by using constrained sums of T dependence, to select best L to fit the data which includes T-cut approach to get long-mode asymptotes, multiple mode sums to get asymptotes, and single mode fits to see deviations from single mode behavior. In another embodiment, the step of providing an absolute nuclear material assay comprises H-C Point-model extension by using constrained sums of T dependence, to select best L to fit the data which includes T-cut approach to get long-mode asymptotes, multiple mode sums to get asymptotes, and single mode fits to see deviations from single mode behavior and subsequently, use the best fit parameters from the model for analysis. In another embodiment, the step of providing an absolute nuclear material assay comprises dead-time correction based on T dependence perturbations/shifts. In another embodiment, the step of providing an absolute nuclear material assay comprises using L to estimate moderator mass around the fissioning material. In another embodiment, the step of providing an absolute nuclear material assay comprises precomputing lookup tables of real-time computed count distributions for comparison to measured data. Sources of fission neutrons can be statistically distinguished from random neutron sources. A random source produces a Poisson distribution, b n = C n n ! ⁢ ⁢ ⅇ - C Equation ⁢ ⁢ ( 1 ) for the probability to detect a particular number, n, during a counting window, where C is the average number of counts during that counting time. A fission source produces a distribution with a larger width. Since fission chains produce multiple neutrons in bursts, the larger width, or larger fluctuation, is related to the probability to detect more than one neutron from the same fission chain. The form of the counting distribution for a fission source is a generalized Poisson distribution. Unlike the Poisson distribution that depends on only a single time dependent parameter, C═R t, where R is the count rate, the generalized Poisson distribution depends on many, in principle even an infinite number, of time dependent parameters, Ak(t), k=1, 2, 3, . . . . If bn(t) is the probability to get n neuron counts in a time gate of length t, then, b n = b 0 ⁢ ⁢ ∑ i 1 + 2 ⁢ ⁢ i 2 + 3 ⁢ ⁢ i 3 + … + n ⁢ ⁢ i n = n ⁢ Λ 1 i 1 ⁢ Λ 2 i 2 ⁢ ⁢ … ⁢ ⁢ Λ n i n i 1 ! ⁢ ⁢ i 2 ! ⁢ ⁢ … ⁢ ⁢ i n ! Equation ⁢ ⁢ ( 2 ) where ik is the number of independent chains contributing k counts (for k=n, in=0 or 1, while for k=1, i1=0, 1, 2, . . . , n), andb0=exp[−(Λ1+Λ2+ . . . +Λn+ . . . )]. Equation (3) For example, the probability to get 5 counts is b 5 = ( Λ 5 + Λ 4 ⁢ Λ 1 + Λ 3 ⁢ Λ 2 + Λ 3 ⁢ Λ 1 2 2 ! + Λ 2 2 2 ! ⁢ Λ 1 + Λ 2 ⁢ Λ 1 3 3 ! + Λ 1 5 5 ! ) ⁢ exp ⁡ [ - ( Λ 1 + Λ 2 + … ⁢ ) ] . Equation ⁢ ⁢ ( 4 ) If all the Lk but Λ1 are zero, then b5→Λ15e−Λ1/5!, a Poisson distribution. The term Λ15e−(Λ1+Λ2+ . . . )/5! represents the probability that each of the 5 counts was due to an independent random source, where only a single neutron is counted from each independent chain. The term Λ5e−(Λ+Λ2+ . . . )L is the probability that all 5 counts arise from a common source, a single chain. The term Λ22Λ1e−(Λ1+Λ2+ . . . )/2!, for example, is the probability that the 5 counted neutrons arise from 3 independent random sources, two pairs of counts each have a different common ancestor, and an additional count arises from a third source. For a weak neutron source in a system of high multiplication, it is likely to get multiple counts from the same chain, but the chains are few and far between. For a strong source in a system of low multiplication, the probability of getting multiple counts from a single chain is small, while the probability of getting many counts, most from independent chains, is high. So clearly information about the source strength and multiplication are encoded in the counting distribution. Applicants would like to have a complete theory relating the material and detector properties to the time dependent counting distribution. This requires a more complete theory of fission chains. In this paper Applicants develop an analytic formula for the t→∞ fission chain, and, in the approximation that at most two neutrons are emitted in an induced fission, a closed form expression for the time evolving fission chain. These formulas apply in the point model approximation, in which spatial dependence and neutron spectrum are neglected. Referring again to FIG. 2, the system 200, comprises step 201 measurements with multigate neutron multiplicity counter, step 202 solve three moment equations, step 203, use fit to actual time dependence of the moments to get proper asymptotic moments, step 204 use the estimated parameters to compute the full count distribution, step 205 compare truncated data moments with untruncated and truncated theoretical moments, and step 206 making adjustments to reduce bias. Measurements are made with the multigate neutron multiplicity counter 101. Three moment equations are solved with the truncated asymptotes to estimate three of the unknowns (MmeA), given one parameter. A fit to the actual time dependence of the moments is used to get the proper asymptotic moments. Since the estimates are biased (wrong) because of the finite sampling problem, Applicants use the estimated parameters to compute the full count distribution that would have been measured if there was no finite sampling error. Then Applicants compare the truncated data moments (measurement) with untruncated and truncated theoretical moments. Adjustments to reduce bias in the moments or count distributions are then possible via a data entry window. While the invention may be susceptible to various modifications and alternative forms, specific embodiments have been shown by way of example in the drawings and have been described in detail herein. However, it should be understood that the invention is not intended to be limited to the particular forms disclosed. Rather, the invention is to cover all modifications, equivalents, and alternatives falling within the spirit and scope of the invention as defined by the following appended claims.
	summary
	claims	1. A system for enhancing electron screening comprising:an electrically conductive base structure, the base structure including light element atoms and containing free electrons; anda source of electromagnetic (EM) radiation applied to the base structure, the EM radiation having an excitation frequency, wherein the base structure is configured such that:in response to the EM radiation, the free electrons oscillate between at least two localized regions of the base structure; andthe oscillation generates periodic charge density variations around a portion of the light element atoms that are disposed in the at least two localized regions. 2. The system of claim 1, wherein the oscillation includes a plasmon oscillation. 3. The system of claim 1, wherein the source of EM radiation comprises one or more of a laser, a diode, an electron generator, a voltage generator, a microwave generator, a radio wave generator, or a magnetron. 4. The system of claim 1, wherein the excitation frequency is at least about 1 GHz. 5. The system of claim 1, wherein the base structure comprises a nanostructure. 6. The system of claim 1, wherein the source of EM radiation is configured to generate EM radiation having a wavelength of from about 10 microns to about 0.1 micron. 7. The system of claim 1, wherein the source of EM radiation is configured to generate EM radiation having X-ray or gamma ray wavelengths. 8. The system of claim 1, wherein at least a portion of the base structure includes at least one electrode having a tapering section and a tip, the tip being proximate to a discontinuity. 9. The system of claim 8, wherein the discontinuity is configured as a knife edge, an annular knife edge, disposed at the tip. 10. The system of claim 8, wherein the source of EM radiation is configured to operate at a power that is generally less than about 1 mW per discontinuity. 11. The system of claim 1, wherein at least a portion of the base structure comprises a discontinuity. 12. The system of claim 11, wherein at least a portion of the discontinuity comprises a shape that is at least partially generally circular, square, rectangular, elliptical, tubular, or pointed. 13. The system of claim 11, wherein at least a portion of the discontinuity comprises a nanostructure. 14. The system of claim 13, wherein the discontinuity is coated with a coating material to enhance or facilitate a flow of electrons along its surface. 15. The system of claim 14, wherein the coating material is a gold, copper, silver or other conducting material. 16. The system of claim 11, wherein the discontinuity has an area of about of about 10 nm2 or less. 17. The system of claim 11, wherein the base structure comprises two or more discontinuities. 18. The system of claim 1, wherein at least a portion of the base structure comprises an array of discontinuities. 19. The system of claim 18, wherein the array of discontinuities is coupled to at least one substrate arranged on a support structure. 20. The system of claim 1, wherein at least a portion of the base structure is coated with a coating material to enhance or facilitate a flow of electrons along its surface. 21. The system of claim 20, wherein the coating material is a gold, copper, silver or other conducting material. 22. The system of claim 1, wherein the light element atoms have an atomic mass of 62 or less. 23. The system of claim 1, wherein at least a portion of the base structure is configured to exhibit a ratio of light element atoms to other atoms of at least 2 to 1. 24. The system of claim 1, wherein light element atoms include one or more of hydrogen-1, deuterium, boron-11, lithium-6, lithium-7, deuterium, tritium, helium-3, nitrogen-15 and tritium. 25. The system of claim 1, wherein at least a portion of the base structure comprises one or more of palladium, tungsten, boron hydride, titanium, tantalum. 26. The system of claim 1, wherein at least a portion of the base structure comprises one or more getter materials. 27. The system of claim 1, further comprising one or more antenna structures. 28. The system of claim 27, wherein said one or more antenna structures comprise a first structure and a second metal structure respectively located on opposite sides of said base structure. 29. The system of claim 1, wherein the source of EM radiation is configured to operate at a power between about 1 mW and 10 mW. 30. The system of claim 1, wherein electron screening substantially offsets or reduces the effect of the Coulomb barrier between nuclei of the light element atoms. 31. The system of claim 30, wherein the oscillating free electrons, create an electric field greater than about 108 volts/meter at a location proximate to the two localized regions and provides localized compression by ponderomotive forces that induces a fusion reaction between nuclei of at least a portion of the light element atoms. 32. An apparatus for enhancing electron screening comprising:an electrically conductive base structure, the base structure including light element atoms and free electrons;the base structure being configured such that, when subjected to applied electromagnetic (EM) radiation, having an excitation frequency:in response to the applied EM radiation, the free electrons oscillate between at least two localized regions of the base structure; andthe oscillation generates periodic charge density variations around a portion of the light element atoms that are disposed in the at least two localized regions. 33. The apparatus of claim 32, wherein the base structure comprises a nanostructure. 34. The apparatus of claim 32, wherein electron screening substantially offsets or reduces the effect of the Coulomb barrier between nuclei of the light element atoms. 35. The apparatus of claim 34, wherein the oscillating free electrons creates an electric field greater than about 108 volts/meter at a location proximate to the two localized regions and provides localized compression by ponderomotive forces that induces a fusion reaction between nuclei of at least a portion of the light element atoms.
	claims	1. A UV curing irradiator comprising:a body including a housing;a UV lamp source extending longitudinally within the housing for producing UV radiation;a handle on the body configured to permit holding the irradiator and to permit moving of the irradiator to a desired orientation;at least one wall defining at least one opening in the housing configured relative to the lamp source to permit UV radiation from the lamp source to pass through the opening; andat least one reflector in the housing, the at least one reflector having a portion extending longitudinally between first and second end portions, and wherein the at least one reflector is positioned relative to the UV lamp source to have a first portion parallel to a portion of the UV lamp source and to have a second portion non-parallel to the UV lamp source and wherein the first end portion of the reflector is open and an adjacent portion of the housing is open and extends in a plane non-parallel to the longitudinal portion of the lamp source. 2. The UV curing irradiator of claim 1 wherein the reflector has a middle portion having a middle cross section profile, and wherein the first end portion of the reflector has no reflective surfaces other than identical to the middle cross section profile. 3. The UV curing irradiator of claim 2 wherein the middle cross section profile is a partial ellipse and the first end portion profile is a partial ellipse. 4. The UV curing irradiator of claim 2 wherein the first end portion profile is configured to permit radiation passing parallel to the reflector to pass beyond the first end portion and outside the housing. 5. The UV curing irradiator of claim 1 wherein the reflector has a middle portion having a middle cross section profile, wherein the first end portion of the reflector terminates in a first end cross section profile substantially identical to the middle cross section profile and wherein the housing includes at least a second wall defining an opening at least as large as the first end cross section profile. 6. The UV curing irradiator of claim 1 wherein the handle includes a portion adjacent a second end portion of the body substantially opposite the first end portion of the reflector. 7. The UV curing irradiator of claim 1 wherein the at least one reflector extends substantially longitudinally of the body and wherein the housing includes at least one wall extending adjacent the at least one reflector and the at least one housing wall includes a plurality of aperture wall defining openings for allowing air to pass through the openings. 8. The UV curing irradiator of claim 7 wherein the at least one reflector includes an upper portion and a lower edge and wherein the at least one wall extends upwardly and the openings are configured to allow air to pass along a portion of the at least one reflector between the upper portion and the lower edge. 9. The UV curing irradiator of claim 7 further including at least one powered air flow device above the upper portion of the at least one reflector. 10. The UV curing irradiator of claim 1 wherein the first portion of the at least one reflector is formed from at least third and fourth reflector portions and wherein the third and fourth reflector portions are mirror images of each other along a plane. 11. A UV curing irradiator comprising:a body including a housing:a UV lamp extending linearly and configured to produce UV radiation when energized and positioned within the housing;a handle on the body and positioned outside the housing;a reflector having a linear portion extending linearly substantially parallel to the UV lamp and positioned inside the housing for reflecting UV radiation; anda first wall defining a first opening in a bottom portion of the housing and a second wall defining a second opening in a side of the housing non-parallel to the linear portion of the UV lamp and wherein the housing and the reflector are configured in such a way that UV radiation can exit the housing through the first opening or through the second opening. 12. A UV curing irradiator comprising:a body including a housing;a UV lamp source within the housing for producing UV radiation;a handle on the body configured to permit holding the irradiator and to permit moving of the irradiator to a desired orientation;at least one wall defining at least one opening in the housing configured relative to the lamp source to permit UV radiation from the lamp source to pass through the opening;at least one reflector in the housing, the at least one reflector having first and second end portions, and wherein the at least one reflector is positioned relative to the UV lamp source to have a first portion parallel to a portion of the UV lamp source and to have a second portion non-parallel to the UV lamp source and wherein the first end portion of the reflector is open and an adjacent portion of the housing is open; andwherein the UV lamp source includes a linear portion having a first electrode extending substantially parallel to the linear portion and a second electrode substantially perpendicular to the linear portion. 13. The UV curing irradiator of claim 12 wherein the second electrode is substantially adjacent the first end portion of the reflector. 14. The UV curing irradiator of claim 11 wherein the reflector has a portion at least in part facing the second opening and wherein the reflector portion is at an end of the UV lamp substantially opposite the second opening. 15. The UV curing irradiator of claim 11 wherein the reflector linear portion is substantially elliptical in transverse cross section. 16. The UV curing irradiator of claim 15 wherein the reflector linear portion is formed from two symmetric reflector portions. 17. The UV curing irradiator of claim 15 wherein the second wall defines the second opening at least as large as the elliptical transverse cross section. 18. The UV curing irradiator of claim 17 wherein the second wall defines a substantially elliptical profile. 19. The UV curing irradiator of claim 11 wherein the first opening is substantially perpendicular to the second opening. 20. The UV curing irradiator of claim 11 wherein the first opening is substantially rectangular. 21. The UV curing irradiator of claim 11 wherein the UV lamp includes a first electrode extending in a first direction and a second electrode extending in a second direction. 22. The UV curing irradiator of claim 21 wherein the first and second electrode directions are substantially perpendicular. 23. The UV curing irradiator of claim 11 wherein the housing includes first and second sidewalls extending away from the first opening and wherein each of the first and second sidewalls includes respective walls defining passageways allowing air to pass through the respective sidewalls. 24. The UV curing irradiator of claim 11 wherein the UV lamp includes an electrode at an end of the UV lamp adjacent to the second opening, and wherein the electrode is perpendicular to a longitudinal axis of the UV lamp. 25. The UV curing irradiator of claim 11 wherein a portion of the UV lamp extending along a longitudinal axis of the UV lamp passes through the second wall. 26. The UV curing irradiator of claim 1 wherein the housing is closed at an end portion opposite the first portion of the reflector and further including a reflector portion adjacent the closed end portion of the housing. 27. A UV curing irradiator comprising:a body;a housing on the body and including a top portion and first and second sidewalls;a UV irradiation lamp extending within the housing;a reflector extending within the housing and between part of the lamp and part of the housing, the reflector including a reflector portion having a first end portion and wherein the reflector portion extends from the first end portion and between the first housing side wall and part of the lamp to an upper reflector portion; anda plurality of walls forming a plurality of openings in the first sidewall configured to allow fluid to flow through the openings and between the reflector portion and the first sidewall. 28. The UV curing irradiator of claim 27 wherein the first sidewall is relatively flat and the plurality of walls define a plurality of rectangular openings. 29. The UV curing irradiator of claim 28 wherein one or more of the rectangular openings combine to form open space extends at least half a height of the first sidewall. 30. The UV curing irradiator of claim 27 wherein the openings include a first set of openings on an upper portion of the first sidewall and a second set of openings on a lower portion of the first sidewall. 31. The UV curing irradiator of claim 27 wherein the reflector extends a first length within the housing from a second reflector end portion to a third reflector end portion and wherein a first opening in the first side wall is adjacent the second reflector end portion and a second opening is adjacent the third reflector end portion. 32. The UV curing irradiator of claim 27 wherein adjacent openings are separated by substantially equal spacing. 33. The UV curing irradiator of claim 27 wherein the reflector is a first reflector and further including a second reflector at an end portion of the first reflector and wherein the openings in the first sidewall are formed only on a side of the second reflector that the first reflector is found. 34. The UV curing irradiator of claim 27 further including a powered fluid flow device on the body above the first sidewall. 35. The UV curing irradiator of claim 34 wherein the fluid flow device forces air through the plurality of openings and between the first wall and the reflector portion. 36. The UV curing irradiator of claim 35 wherein the reflector is a first reflector and further including a second reflector at an end portion of the first reflector and wherein the fluid flow device forces air on a side of the second reflector opposite the first reflector. 37. The UV curing irradiator of claim 36 wherein the fluid flow device forces air on two sides of the second reflector. 38. The UV curing irradiator of claim 27 further including a plurality of walls forming a plurality of openings in the second sidewall configured to allow fluid to flow through the openings and between the reflector portion and the second sidewall. 39. The UV curing irradiator of claim 38 wherein the openings in the first sidewall are substantially the same as the openings in the second sidewall. 40. A UV curing irradiator comprising:a body having a housing for enclosing a portion of the irradiator and having a first wall defining an opening facing a first direction and a second wall defining a second opening facing in a second direction different from the first direction;a handle supported on the body accessible to an operator's hand for orienting the body;a reflector element within at least a portion of the housing; anda UV radiation lamp positioned to illuminate the reflector element and having first and second electrodes for producing UV radiation in an envelope, the first and second electrodes being oriented at an angle with respect to each other. 41. The UV curing irradiator of claim 40 wherein the first and second electrodes are oriented at right angles to each other. 42. The UV curing irradiator of claim 40 wherein part of the lamp extends parallel to a longitudinal axis and wherein the first electrode is oriented parallel to the longitudinal axis and the second electrode is adjacent the second opening. 43. The UV curing irradiator of claim 42 wherein the second opening is positioned in the housing at an end opposite the first electrode. 44. The UV curing irradiator of claim 43 wherein the second electrode extends substantially parallel to a plane of the second opening. 45. The UV curing irradiator of claim 40 wherein the second electrode is positioned less than an inch from the second opening. 46. The UV curing irradiator of claim 45 wherein the second electrode extends substantially parallel to a plane of the second opening. 47. A method of curing a UV curable material using a handheld UV irradiation lamp assembly, the method comprising:positioning a handheld UV irradiation lamp assembly so as to direct UV radiation toward a surface from a first portion of the assembly;irradiating a surface having a UV curable material with radiation emitted from the first portion of the assembly;positioning the assembly so as to direct UV radiation toward a surface from a second portion of the assembly from which radiation had not previously been directed at the time of directing radiation from the first portion of the assembly; andirradiating a surface with radiation emitted from the second portion of the assembly. 48. The method of claim 47 further including irradiating the surface with radiation emitted from the first portion of the assembly and passing the assembly across the surface while irradiating the surface with radiation emitted from the first portion of the assembly. 49. The method of claim 48 further including irradiating a surface with radiation emitted from the second portion of the assembly and passing the assembly across the surface while irradiating the surface with radiation emitted from the second portion of the assembly. 50. The method of claim 49 wherein passing the assembly across the surface while irradiating the surface with radiation emitted from the second portion of the assembly includes irradiating a surface adjacent a corner. 51. The method of claim 50 wherein irradiating a surface adjacent a corner includes irradiating a surface adjacent a corner of three intersecting surfaces. 52. The method of claim 47 further including irradiating a surface with radiation emitted from the second portion of the assembly and passing the assembly across the surface while irradiating the surface with radiation emitted from the second portion of the assembly. 53. The method of claim 47 wherein irradiating a surface with radiation emitted from the first portion of the assembly includes irradiating a surface with radiation emitted from the first portion of the assembly after reflection from at least a partially elliptically-profiled reflector. 54. The method of claim 47 wherein irradiating a surface with radiation emitted from the second portion of the assembly includes irradiating a surface with radiation emitted from the second portion of the assembly after reflection from a substantially flat reflector. 55. The method of claim 54 wherein irradiating a surface with radiation emitted from the first portion of the assembly includes irradiating a surface with radiation emitted from the first portion of the assembly after reflection from at least a partially elliptically-profiled reflector.
051679108	claims	1. A fuel assembly comprising: an upper tie plate having a first nuclear fuel identification code and a second nuclear fuel identification code of a different expression from that of the first nuclear fuel identification code, marked adjacently to each other; a lower tie plate; and a plurality of fuel rods having opposite ends thereof held by said upper tie plate and said lower tie plate. 2. A fuel assembly according to claim 1 wherein said upper tie plate includes a handle having the first nuclear fuel identification code expressed by characters and the second nuclear fuel identification code of the corresponding content expressed by a plurality of separate recesses formed in adjacent to the first nuclear fuel identification code. 3. A fuel assembly according to claim 2 wherein said handle has an additional recess formed on the top thereof, said additional recess having a different width than the width of the recesses constituting the second nuclear fuel identification code and, serving as a read reference for the second nuclear fuel identification code.
	description	This application claims the benefit of Japanese Patent Application No. 2009-244585 filed Oct. 23, 2009, which is hereby incorporated by reference in its entirety. This invention relates to collimator modules for removing scattered X-rays in X-ray detectors and an assembling method thereof. Also, this invention relates to X-ray detectors and X-ray CT (Computed Tomography) devices. X-ray CT devices arrange a detector unit comprising a scintillator for detecting X-rays with respect to an X-ray tube. This detector unit includes a plurality of detector units arrayed in a circular-arc fashion. In the detector unit, an X-ray collimator is arranged between the scintillator and a subject such that only X-rays passed through the subject, not X-rays reflected by the subject, reaches the scintillator. The X-ray CT device disclosed in U.S. Pat. No. 6,587,538 has a base member arranged along a channel direction, a plurality of collimator modules arranged along the channel direction, and a detector unit corresponding to the plurality of collimator modules. The collimator module holds only long sides of a collimator single plate that has a rectangle shape extending in a slice direction by a support member. The collimator module disclosed in U.S. Pat. No. 6,587,538 only holds the long sides of the collimator single plates so that the short sides of the collimator module are not held strongly. With desired improvement of the quality of X-ray CT devices, the holding positions of the collimator single plate must be accurate, but the supporting member disclosed in U.S. Pat. No. 6,587,538 only inserts the long sides of the collimator single plate into grooves. Recently, X-ray detectors used for X-ray CT devices are grown to be multi-detectors having detector-rows of 16 or 64, but the device having detector-rows of 128 or 320 has been proposed. If collimator modules are used for such multi-detector CT having more than 100 detector-rows, the length of the collimator modules in a slice direction (a row direction of the X-ray detector) becomes longer. Thus, if the X-ray detector having collimator modules rotates at a high speed, the collimator single plates are easily bent. Therefore, if the collimator modules disclosed in U.S. Pat. No. 6,587,538 are used, the collimator modules cannot resist the high-speed rotation. The present invention solves the above-mentioned problem by providing collimator modules, X-ray detectors, and a X-ray CT device that are not easily bent at the high-speed rotation of the X-ray detector having the collimator module, and also by providing assembling methods of such collimator modules. A collimator module of a first aspect comprises a plurality of collimator single plates including a rectangle shape having a pair of long sides and a pair of short sides shorter than of the pair of long sides, a pair of blocks including a plurality of first grooves extending along an irradiation direction of the X-rays to which the short sides are inserted and supporting the plurality of collimator single plates which are inserted to the grooves so as to be vertically installed along the irradiation direction of X-rays, and a supporting member configured to cover the long side of the plurality of the collimator single plates from an incident side and an emission side of the X-rays and having X-ray transmission property, wherein the supporting member comprising an incident side fixing sheet and an emission side fixing sheet each having a plurality of second grooves to which the long sides of the collimator single plates are inserted and supporting the plurality of collimator single plates by covering from the incident side and the emission side of the X-rays the plurality of first grooves of the pair of blocks and at least a portion of each of the long sides of the plurality of collimator single plates adjacent to the first grooves with the short sides inserted to the first grooves and the long sides inserted to the second grooves. An X-ray detector of a second aspect comprises a plurality of collimator single plates including a rectangle shape having a pair of long sides and a pair of short sides shorter than of the pair of long sides, a pair of blocks including a plurality of first grooves extending along an irradiation direction of the X-rays to which the short sides are inserted and supporting the plurality of collimator single plates which are inserted to the grooves so as to be vertically installed along the irradiation direction of X-rays, a supporting member configured to cover the long side of the plurality of the collimator single plates from an incident side and an emission side of the X-rays and having X-ray transmission property, wherein the supporting member comprising an incident side fixing sheet and an emission side fixing sheet each having a plurality of second grooves to which the long sides of the collimator single plates are inserted and supporting the plurality of collimator single plates by covering from the incident side and the emission side of the X-rays the plurality of first grooves of the pair of blocks and at least a portion of each of the long sides of the plurality of collimator single plates adjacent to the first grooves with the short sides inserted to the first grooves and the long sides inserted to the second grooves, a detection element configured to detect X-rays which pass through spaces defined between the collimator single plates and fixed to the pair of blocks, and a base member having a plurality of positioning portions placed along an circular-arc fashion for fixing the pair of blocks to a reference position. An X-ray CT device of a third aspect comprises a plurality of collimator single plates including a rectangle shape having a pair of long sides and a pair of short sides shorter than of the pair of long sides, a pair of blocks including a plurality of first grooves extending along an irradiation direction of the X-rays to which the short sides are inserted and supporting the plurality of collimator single plates which are inserted to the grooves so as to be vertically installed along the irradiation direction of X-rays, and a supporting member configured to cover the long side of the plurality of the collimator single plates from an incident side and an emission side of the X-rays and having X-ray transmission property, wherein the supporting member comprising an incident side fixing sheet and an emission side fixing sheet each having a plurality of second grooves to which the long sides of the collimator single plates are inserted and supporting the plurality of collimator single plates by covering from the incident side and the emission side of the X-rays the plurality of first grooves of the pair of blocks and at least a portion of each of the long sides of the plurality of collimator single plates adjacent to the first grooves with the short sides inserted to the first grooves and the long sides inserted to the second grooves, a detection element configured to detect X-rays which pass through spaces defined between the collimator single plates and fixed to the pair of blocks, and a base member having a plurality of positioning portions placed along an circular-arc fashion for fixing the pair of blocks to a reference position. A method for assembling of a collimator module which has a plurality of collimator single plates including a rectangle shape having a pair of long sides and a pair of short sides shorter than of the pair of long sides, a pair of blocks including a plurality of first grooves extending along an irradiation direction of the X-rays to which the short sides are inserted and supporting the plurality of collimator single plates which are inserted to the grooves so as to be vertically installed along the irradiation direction of X-rays, and a supporting member configured to cover the long side of the plurality of the collimator single plates from an incident side and an emission side of the X-rays and having X-ray transmission property, wherein the supporting member comprising a first incident side fixing sheet and a first emission side fixing sheet each having a plurality of second grooves to which the long sides of the collimator single plates are inserted and supporting the plurality of collimator single plates by covering from the incident side and the emission side of the X-rays the plurality of first grooves of the pair of blocks and at least a portion of each of the long sides of the plurality of collimator single plates adjacent to the first grooves with the short sides inserted to the first grooves and the long sides inserted to the second grooves, and a second incident side fixing sheet and a second emission side fixing sheet each having a plurality of second grooves to which the long sides of the collimator single plates are inserted and covering the long sides of the collimator single plates with gaps between those and the first incident side fixing sheet and the first emission side fixing sheet respectively with the long sides of the collimator single plates inserted to the second grooves, the method comprising steps of inserting the short sides of the collimator single plates to the plurality of first grooves of the pair of blocks, arraying the collimator single plates by pushing the collimator single plates to one wall surface of the first groove by a pressing component placed so as to sandwich the long sides of each of the plurality of collimator single plates, fixing the long sides of the collimator single plates to the first incident side fixing sheet and the first emission side fixing sheet, and the second incident side fixing sheet and the second emission side fixing sheet, by placing the first incident side fixing sheet and the first emission side fixing sheet such that the long sides on the incident side and the emission side of the arrayed collimator single plates being inserted to the first grooves of the first incident side fixing sheet and the first emission side fixing sheet, and the second grooves of the second incident side fixing sheet and the second emission side fixing sheet and that the pressing component are arranged to a gap between the first incident side fixing sheet and the first emission side fixing sheet, and between the second incident side fixing sheet and the second emission side fixing sheet, and removing the pressing component from the collimator single plates. A plurality of first grooves of a pair of blocks and a portion of each of the long sides of a plurality of collimator single plates adjacent to the plurality of first grooves of the pair of blocks are covered from an incident side and an emission side of X-rays where short sides of the collimator single plates are inserted to the first grooves and long sides of the collimator single plates are inserted to the second grooves. Therefore, the strength of the collimator module of present invention is increased and the collimator single plates are not easily bent. FIG. 1 is a perspective view of an X-ray CT (computer tomography) device 100. As shown in FIG. 1, the X-ray CT device has a scanning gantry 101 for scanning a subject S (shown in FIG. 2) and acquiring a projection data and a cradle 102 on which the subject is placed going in to and out of a bore 104 of the scanning gantry 101 which is a scanning area. The X-ray CT device 100 further has an operating console 103 to operate the X-ray CT device 100 and to reconstruct images based on the acquired projection data. In FIG. 1, a body axis direction of the subject S on the cradle 102 is a Z-axis direction, a direction perpendicular to the ground (to the Z-axis) is Y-axis direction, and a direction orthogonal to the Z-axis and Y-axis is X-direction. The cradle 102 contains a motor to elevate and to move the cradle 102 horizontally. The subject S is placed on the cradle 102 and the cradle 102 goes in to and out of the bore 104 of the scanning gantry 101. The operating console 103 is provided with an input device receiving inputs from an operator and a monitor for displaying images. Also, the operating console 103 has a central processing device for controlling each member to acquire the projection data of the subject or processing three-dimensional image reconstruction, a data acquisition buffer for acquiring obtained data by the scanning gantry 101, and a memory device for memorizing programs, data, and the like. FIG. 2 is a schematic view for explaining the scanning gantry 101. As shown in FIG. 2, the scanning gantry 101 has an X-ray tube 30 for scanning the subject S and an X-ray detector 40. In the X-ray CT device 100, as shown with a solid line and a dotted line shown in FIG. 2, the X-ray tube 30 rotates at high speed along a direction of an arrow AR (channel direction) and irradiates X-rays to the subject S, and then the X-ray detector 40 detects the X-rays. The X-ray tube 30 can be a structure that a housing contains a cathode sleeve and a target electrode supported in opposed position of the cathode sleeve in X-axis direction. The target electrode comprises disk-shaped tungsten, for example. The target electrode is a rotary type that rotates about an axis and also a reflection type that generates X-rays on the same surface that an electron beam collides. The X-ray detector 40 has a collimator 20 and a plurality of X-ray detection elements 50 for detecting X-rays. The collimator 20 has a plurality of collimator modules 10 for collimating X-rays from the X-ray tube 30 and an arc-shaped base 60 fixing the plurality of the collimator modules 10 in reference positions. One X-ray detection element 50 is fixed to one collimator module 10 at the opposite side of the X-ray tube 30 so as to be situated between the X-ray tube 30 and the collimator module 10. That means, a plurality of X-ray detection elements 50 are arrayed on XY surfaces of the collimator module 10 along the channel direction. Then, the X-ray detection element 50 detects the X-rays passed through the subject S, which is put on the cradle 120 and carried to the bore 104. The X-ray detection element 50 has a photodiode chip (not shown) having photoelectric conversion devices arrayed in the channel direction and a slicing direction and a scintillator block (not shown) that emits visible light by receiving X-rays. The X-ray detection element 50 further comprises a semiconductor chip (not shown) having functions to estimate outputs from the photodiode chip and to switch outputs for changing a slicing thickness. The X-ray CT device 100 further includes an X-ray controller for controlling a width of X-ray beams from the X-ray tube 30 and a rotation controller for controlling rotations of the X-ray tube 30 and the X-ray detector 40. Overview of Collimator 20 FIG. 3 is a perspective view for explaining a configuration of the collimator 20. As shown in FIG. 3, the collimator 20 includes a plurality of collimator modules 10 for collimating X-rays from the X-ray tube 30 and an arc-shaped base 60 for fixing the plurality of collimator modules 10 in reference positions. For convenience of explanation the X-ray tube 30 is illustrated, but it is not included in the collimator 20. The base 60 includes a rectangular frame having a pair of circular-arc base members 61 and a pair of linear base members 62 connecting the distal ends of the base members 61. Also positioning pins or positioning holes on the base side for positioning the plurality of collimator modules 10 are provided on the base member 61. In FIG. 3, although the positioning pins or holes are illustrated only on the base side at the reference position corresponding to the collimator module 10 which is surrounded by a dotted line BB, the positioning pins or holes are formed on all reference positions corresponding to collimator modules 10. At the base 60, a length L1 is in a range of 350 mm to 400 mm for example, a thickness H is about in a range of 35 mm to 40 mm, and a length L2 comprising the base members 61 and 62 is in about a range of 300 mm to 350 mm. A width D of each collimator module 10 is 50 mm, for example. The collimator module 10 will be described below by referring FIG. 4. For a material of the base 60, a carbon fiber reinforced plastic (CFRP) which is a composite material of aluminum alloy or carbon fiber and a thermoset resin is used because aluminum alloy or CFRP is light in weight and strong, and also has a characteristic of high durability. Because of the light weight and a strong characteristic of aluminum alloy or CFRP, the base 60 can be rotated at high speed in the scanning gantry 101 of the X-ray CT device 100 without generating unnecessary centrifugal forces. Additionally, because of the high durability of aluminum alloy or CFRP, the base 60 hardly strain or bend and so do the collimator modules 10 fixed thereon. In FIG. 3, thirteen collimator modules 10 are fixed to one base 60, but dozens of collimator modules 10 are fixed for an actual usage. The collimator module 10 has a plurality of collimator single plates 11, which collimates X-ray beams from the X-ray tube 30. Note that this application discloses a first collimator module 10A through a fifth collimator module 10E as the collimator module 10. The collimator module 10 illustrated in FIG. 3 is a second collimator module 10B which will be described by referring FIG. 8 later on. FIGS. 4A and 4B show the first collimator module 10A of a first embodiment. FIG. 4A is a plan view of the first collimator module 10A. FIG. 4B is a side view of the first collimator module 10A. As shown in FIGS. 4A and 4B, the first collimator module 10A includes dozens of rectangle collimator single plates 11, a pair of blocks 12 formed at both distal ends in Z-axis direction of the collimator single plate 11, and a first incident side fixing sheet 13 and a first emission side fixing sheet 15 formed at both distal ends of the collimator single plate 11 and the blocks 12. In the first embodiment, the collimator single plate 11 is bonded to the blocks 12 by an adhesive, and the first incident side fixing sheet 13 and the first emission side fixing sheet 15 are bonded to the collimator single plate 11 and the blocks 12 by an adhesive. The bonding manner will be explained later by referring FIG. 12. The first collimator module 10A has a size to be placed on the base 60 shown in FIG. 3. That is, a total length L1 in Z-axis direction of the first collimator module 10A is about 350 mm, which is the same length of the length L1 of the base 60, a thickness in Y-axis direction of the first collimator module 10A is about 30 mm, which is the same thickness of the thickness H of the base 60, and a width D in X-axis direction of the first collimator module 10A is about 50 mm. A length of long sides of the collimator single plate 11 is about 300 mm and is almost the same length of the length L2 of the inner space of the base 60, and a length HH of short sides of the collimator single plate 11 is about 30 mm and is slightly longer than a thickness WW (see FIG. 7) of the block 12. The collimator single plate 11 is a rectangle shape comprising a pair of long sides 11LS and a pair of short sides 11SS, and the four corners preferably are chamfered or round chamfered. Note that a board thickness t1 (see FIG. 5 and FIG. 6) of the collimator single plate 11 is about 0.2 mm. The collimator single plate 11 is made of a heavy metal having a high X-ray absorption rate, e.g., molybdenum, tungsten, or lead. When the first collimator module 10A is attached to the base 60 shown in FIG. 3, the X-ray tube 30 is positioned on the extensions of the short side of the collimator single plate 11. Also, the extensions of the long side of the collimator single plate 11 are parallel to the body axis of the subject S, which is the slicing direction. The block 12 is made of a plastic or a light-weight metal such as aluminum, and the block 12 has an incident side surface 121 and an emission side surface 12E contacting to the base 60. +Y side is a side facing the X-ray tube 30 and X-rays entered from +Y side travels to −Y side. A plurality of first grooves 125 (see FIG. 5) extending from the incident side surface 121 to the emission side surface 12E is formed on the surfaces 12S facing one another. The block 12 further comprises a flange 12F, so it is like a shape of letter “L” when it is seen from X-axis direction. Also, a hole is formed on the flange 12F of the block 12 for inserting a positioning pin 122 to fix the block 12 at a reference position of the base 60. Next, the block 12 is described by referring FIG. 5. FIG. 5 is an enlarged view of a part surrounding a dotted line CC of FIG. 4A. For a better understanding, twenty four collimator single plates 11 are illustrated in FIG. 5, but the fist collimator module 10A actually has dozens of them. For showing the first grooves 125, the first incident side fixing sheet 13 is illustrated transparent (and it is shown with a dotted line). As shown in FIG. 5, dozens of first grooves 125 corresponding to dozens of collimator single plates 11 are formed on the block 12. A width t2 in X-axis direction of the first groove 125 is formed wider than a width t1 of the collimator single plate 11 for inserting the collimator single plate 11 into the first groove 125 and the width t2 is 0.24 mm, for example. Thus, the short side 11SS of the collimator single plate 11 is inserted easily into the first groove 125. In the first embodiment, a depth W0 of the first groove 125 into which the collimator single plate 11 is inserted is 1.0 mm, for example. A side surface 125K in a −X direction of the first groove 125 of the block 12 is formed at a precise position corresponding to the position of the positioning pin 122. Thus, if the base block 12 is fixed to the right position of the base 60, the collimator single plate 11 is touched to the side surface 125K firmly so that a plurality of collimator single plates 11 are positioned at right positions. The first groove 125 is formed from the incident side surface 12I to the emission side surface 12E and it spreads like a fan shape when it is seen from Z-axis direction. That is, when the first collimator module 10A is attached to the base 60 shown in FIG. 3, the X-ray tube 30 is positioned on the extensions of the short side of the collimator single plate 11. Further, a hole is formed at the center of the flange 12F of the block 12 and a positioning pin 122 is inserted into the hole. This positioning pin 122 is for positioning and placing the collimator module 10 at the reference position of the base 60 as explained by referring FIG. 3. Four positioning holes 123 are formed around the positioning pin 122. Those four positioning holes 123 are configured to fix the X-ray detection element 50 shown in FIG. 2 to a right position. Next, the first incident side fixing sheet 13 and the first emission side fixing sheet 15 are explained by referring FIG. 6 and FIG. 7. FIGS. 6A and 6B is a figure for explaining the first incident side fixing sheet 13 and the first emission side fixing sheet 15. FIG. 6A is a perspective view of the first incident side fixing sheet 13 and the first emission side fixing sheet 15. FIG. 6B is a cross-sectional view along A-A line of FIG. 4A where the first incident side fixing sheet 13 and the first emission side fixing sheet 15 are bonded to the collimator single plate 11 by an adhesive (not shown). As shown in FIG. 6A, a plurality of second grooves 135 are formed parallel to one another on one surface of the first incident side fixing sheet 13. A plurality of second grooves 155 are formed on one surface of the first emission side fixing sheet 15 in the same fashion. In the first embodiment, the pitch of the second grooves 135 and the second grooves 155 is formed so as to be the same pitch of the first grooves 125 of the block 12 explained by referring FIG. 5. A thickness h3 of the first incident side fixing sheet 13 and the first emission side fixing sheet 15 is about in a range of 0.2 mm to 0.5 mm, and a depth h4 of the second groove 135 and 155 is in a range of 0.1 mm to 0.3 mm. Note that the second grooves 135 on the first incident side fixing sheet 13 and the second grooves 155 on the first emission side fixing sheet 15 are formed depending on the number of collimator single plates 11. Further, the first incident side fixing sheet 13 is situated on a side where X-rays come in, and the first emission side fixing sheet 15 is situated on a side where X-rays emit. For the first incident side fixing sheet 13 and the first emission side fixing sheet 15, carbon fiber reinforced plastic (CFRP) having X-ray transmission property is used. The long sides 11LS, which is a side X-rays come in (+Y side in FIG. 6B), of the collimator single plates 11 are inserted into the second grooves 135 of the first incident side fixing sheet 13 and bonded by an adhesive. Also, the long sides 11LS, which is a side X-rays emit (−Y side in FIG. 6N), of the collimator single plates 11 are inserted into the second grooves 155 of the first emission side fixing sheet 15 and bonded by an adhesive. As shown in FIG. 6B, a width t3 of the second grooves 135 and 155 is wider than a width t1 of the collimator single plate 11 so that respective long sides 11LS of the collimator single plates 11 can be easily inserted into the second grooves 135 and 155. FIG. 7 is a partial enlarged view of an area surrounded by a dotted line CC of FIG. 4B. As shown in FIG. 7, the short side 11SS of the collimator single plate 11 is inserted into the first groove 125 of the block 12 and the long side 11LS of the collimator single plate 11 is inserted into the second groove 135 of the first incident side fixing sheet 13 and the second groove 155 of the first emission side fixing sheet 15. The short side 11SS of the collimator single plate 11 is fixed to the first groove 125 by an adhesive, and a portion of each of the long sides 11LS of the collimator single plate 11 is fixed to the second grooves 135 and 155 by an adhesive as well. Note that a width HH of the collimator single plate 11 is wider than a thickness WW of the bock 12 for the depth of the second grooves 135 and 155 (h4×2). Parts of the first incident side fixing sheet 13 other than the second grooves 135 are fixed to the incident-side surface 12I of the block 12 by an adhesive. Also, parts of the first emission side fixing sheet 15 other than the second grooves 155 are fixed to the emission-side surface 12E of the block 12 by an adhesive. As a result, the short side 11SS of the collimator single plate 11 is firmly fixed to the first groove 125 of the block 12. Also, a portion of the long side 11LS adjacent to the short side 11SS is fixed to the block 12 by the first incident side fixing sheet 13 and the first emission side fixing sheet 15. Further, as shown in FIG. 5, the collimator single plate 11 is precisely positioned on the wall surface of the first groove 125K in −X direction. According to the collimator module of the first embodiment, in a condition that the short sides 11SS are inserted into the first grooves 125 and the long sides 11LS are inserted into the second grooves 135, the first grooves 125 of the block 12 and the portions of the collimator single plate 11 adjacent to the first grooves 125 are covered at four places by the first incident side fixing sheet 13 and the first emission side fixing sheet 15 respectively. Therefore, dozens of the collimator single plates 11 hardly bend even they are arrayed in more than a line of 300 mm in a long side direction when the X-ray detector 40 having the first collimator module 10A rotates at high speed. Then, the collimator single plate 11 collimates X-ray beams from the X-ray tube 30 accurately. Note that in the collimator module of the first embodiment, only some portions of the long side 11LS of the collimator single plate 11 are covered by the first incident side fixing sheet 13 and the first emission side fixing sheet 15, but entire portion of the long side 11LS can be covered. FIGS. 8A and 8B are figures for explaining a second collimator module 10B of a second embodiment. FIG. 8A is a flat view of the second collimator module 10B. FIG. 8B is a side view of the second collimator module 10B. As shown in FIGS. 8A and 8B, the second collimator module 10B has a configuration that four second incident side fixing sheets 14 and four second emission side fixing sheets 16 are fixed in addition to the first collimator module 10A of the first embodiment. Thus, the second collimator module 10B has the same configuration of the first embodiment except the second incident side fixing sheets 14 and the second emission side fixing sheets 16. On respective one surfaces of the second incident side fixing sheet 14 and the second emission side fixing sheet 16, dozens of second grooves are formed with designated pitches respectively same as the first incident side fixing sheet 13 and the first emission side fixing sheet 15. For the second incident side fixing sheet 14 and the second emission side fixing sheet 16, carbon fiber reinforced plastic (CFRP) having X-ray transmission property is also used. The second incident side fixing sheet 14 that only bonds to the long side 11LS of the collimator single plate 11 is arranged on a side of the collimator single plate 11 where X-ray beams come in (+Y side). The second emission side fixing sheet 16 that only bonds to the long side 11LS of the collimator single plate 11 is arranged on another side of the collimator single plate 11 where X-ray beams exit (−Y side). A space having a distance L3 is maintained between the first incident side fixing sheet 13 and the second incident side fixing sheet 14. The space (distance L3) is also maintained between adjacent second incident side fixing sheets 14, between the first emission side fixing sheet 15 and the second emission side fixing sheet 16, and between adjacent second emission side fixing sheets 16. The space (distance L3) corresponds to the pressing components 75 (see FIGS. 12 and 13) arraying a plurality of collimator single plates 11. According to the collimator module of the second embodiment, the first groove 125 of the block 12 and the portions of the collimator single plate 11 adjacent to the first groove 125 are covered at four places by the first incident side fixing sheet 13 and the first emission side fixing sheet 15 respectively when the short sides 11SS are inserted into the first grooves 125 and the long sides 11LS are inserted into the second grooves 135. And the second incident side fixing sheet 14 and the second emission side fixing sheet 16 further cover the long sides 11LS of the collimator single plates 11. Therefore, dozens of the collimator single plates 11 hardly bend even they are arrayed in more than a line of 300 mm in a long side direction when the X-ray detector 40 having the second collimator module 10B rotates at high speed. Then, the collimator single plate 11 collimates X-ray beams from the X-ray tube 30 accurately. Moreover, according to the collimator module of the second embodiment, a space (L3) is maintained between the first incident side fixing sheet 13 and the second incident side fixing sheet 14, between adjacent second fixing sheets on incident side 14, between the first emission side fixing sheet 15 and the second emission side fixing sheet 16, and between adjacent second emission side fixing sheets 16. Therefore, in the manufacturing process, a jig (i.e. pressing components 75 shown in FIG. 12 and FIG. 13) for arraying a plurality of the collimator single plates is inserted into the spaces to array the collimator single plates with a high degree of accuracy. Further, the collimator single plates can be observed or inspected by an imaging camera through the spaces to determine whether the collimator single plates are arrayed correctly. FIG. 9 is a figure for explaining a third collimator module 10C of the third embodiment. FIG. 9A is a flat view of the third collimator module 10C. FIG. 9B is a side view of the third collimator module 10C. As shown in FIGS. 9A and 9B, the third collimator module 10C has a configuration that five third fixing sheets on incident side 17 and five third fixing sheets on emission side 19 are fixed in addition to the second collimator module 10B of the second embodiment. Thus, the third collimator module 10C has the same configuration of the second embodiment except the third incident side fixing sheet 17 and the third emission side fixing sheet 19. The third incident side fixing sheet 17 and the third emission side fixing sheet 19 have dozens of third grooves on one surface with designated pitches respectively same as the first incident side fixing sheet 13 and the first emission side fixing sheet 15. CFRP having X-ray transmission property is also used for the third incident side fixing sheet 17 and the third emission side fixing sheet 19. As explained in the second embodiment, the space having distance L3 is maintained between the first incident side fixing sheet 13 and the second incident side fixing sheet 14, and also between adjacent second fixing sheets on incident side 14. The fixing sheets on emission side are arranged in the same fashion. The third incident side fixing sheet 17 and the third emission side fixing sheet 19 are fixed so as to fill the spaces (L3). The third collimator module 10C can fix a plurality of collimator single plates 11 more firmly by bonding the third incident side fixing sheet 17 and the third emission side fixing sheet 19 to the long sides of the collimator single plates 11 of the second collimator module 10B of the second embodiment. FIGS. 10A and 10B are figures for explaining a fourth collimator module 10D of the fourth embodiment. FIG. 10A is a top view of the fourth collimator module 10D. FIG. 10B is a side view of the fourth collimator module 10D. As shown in FIGS. 10A and 10B, the fourth collimator module 10D has a configuration that a supporting sheet on incident side 21 and a supporting sheet on emission side 22 are additionally bonded to the second collimator module 10B of the second embodiment. The fourth collimator module 10D has the same configuration of the second embodiment except the supporting sheet on incident side 21 and the supporting sheet on emission side 22. CFRP having X-ray transmission property is used for the supporting sheet on incident side 21 and the supporting sheet on emission side 22. The length L4 in Z-axis direction of the incident side supporting sheet 21 and the emission side supporting sheet 22 is about in a range of 310 mm to 360 mm. One incident side supporting sheet 21 is attached to the outer side of the first incident side fixing sheet 13 and the second incident side fixing sheet 14 so as to cover entire outer surface of the first incident side fixing sheet 13 and the second incident side fixing sheet 14. One emission side supporting sheet 22 is also attached to the outer side of the first emission side fixing sheet 15 and the second emission side fixing sheet 16 so as to cover entire outer surface of the first emission side fixing sheet 15 and the second emission side fixing sheet 16. The incident side supporting sheet 21 and the emission side supporting sheet 22 do not touch the long sides 11LS of a plurality of the collimator single plates 11. Therefore, the incident side supporting sheet 21 and the emission side supporting sheet 22 may have or may not have the second grooves 135 and 155. By fixing the supporting sheet on incident side 21 and the supporting sheet on emission side 22 at the incident side and the emission side further, the forth collimator module 10D can fix the collimator single plates 11 more firmly. It is not particularly shown, but the second collimator module 10B on which either the supporting sheet on incident side 21 or the supporting sheet on emission side 22 is attached can be provided. FIGS. 11A and 11B are figures for explaining a fifth collimator module 11E of the fifth embodiment. FIG. 11A is a top view of the fifth collimator module 10E. FIG. 11B is a side view of the fifth collimator module 10E. As shown in FIG. 11, the fifth collimator module 10E has a configuration that the incident side supporting sheet 21 and the emission side supporting sheet 22 are further attached to the third collimator module 10C of the third embodiment. The fifth collimator module 10E has the same configuration of the third embodiment except the incident side supporting sheet 21 and the emission side supporting sheet 22. In the third embodiment, the third incident side fixing sheets 17 are fixed to spaces between the first incident side fixing sheet 13 and the second incident side fixing sheet 14, and between adjacent second incident side fixing sheets 14. The incident side supporting sheet 21 is attached so as to cover the first incident side fixing sheet 13, the second incident side fixing sheet 14, and the third incident side fixing sheet 17. It is not particularly shown, but the emission side is in the same fashion. By fixing the supporting sheet on incident side 21 and the supporting sheet on emission side 22 to the third collimator module 10C of the third embodiment further, the fifth collimator module 10E can fix the collimator single plates 11 more firmly. It is not particularly shown, but the third collimator module 10C on which either the incident side supporting sheet 21 or the emission side supporting sheet 22 is attached can be provided. Assembling Processes of Collimator Module 10 Next, assembling processes of the collimator module 10 is explained. First of all, a jig 10 used for assembling the collimator module 10 is explained by referring FIG. 12 through FIG. 14. FIG. 12 is a flat view of the jig 70 (including a lower jig 70A and an upper jig 70B) for assembling the collimator module 10. The upper jig 70B is not illustrated for an explanation. In FIG. 12 and FIG. 13, the assembling steps of the second collimator module 10B explained in the second embodiment is explained. Thus, the first incident side fixing sheets 13 and the second incident side fixing sheets 14 are fixed to a plurality of collimator single plates 11 of the second collimator module 10B. However in FIG. 12, one of the second incident side fixing sheet 14 is removed for making the collimator single plate 11 more visible. As shown in FIG. 13, the jig 70 has the lower jig 70A and the upper jig 70B. A pair of pillar mounts 72, connecting pillars 74, and a pair of collimator module fixing blocks 79 that are explained later are attached to the lower jig 70A, which is the only difference between the lower jig 70A and the upper jig 70B. Now the lower jig 70A is mainly used for explanation. For members shown in FIG. 12, FIG. 13, and FIG. 17, “A” is suffixed to numberings of lower members and “B” is suffixed to numberings of upper members for distinguishing. The lower jig 70A has a frame 70 having a penetrating portion 78 formed at the center. A supporting block 90 explained by referring FIG. 15 is inserted into the penetrating portion 78. In FIG. 13, the supporting block 90 is shown with dotted lines. Five pairs of blocks for pressing component 73A are placed in X-axis direction on the upper surface of the frame 71. One pair of blocks for pressing component 73A positions one pressing component 75. The pressing component 75 lying next to each other are arrayed in Z-axis direction at a distance L11, such as 50 mm. The pressing components 75 align a plurality of collimator single plates 11 extending in Z-axis direction. The pressing component 75 has a reference board 751, a comb-shaped member 752, and a spring board member 753. The pressing component 75 has a width L4 in Z-axis direction. The width L4 of the pressing component 75 is narrower than the space L3 between the first incident side fixing sheet 13 and the second incident side fixing sheet 14, or between adjacent second incident side fixing sheets 14. This is because, in a condition that the pressing components 75 have aligned a plurality of collimator single plates 11, the first incident side fixing sheet 13 and the second incident side fixing sheet 14 are fixed at positions where the pressing components 75 are not arranged. A pair of pillar mounts 72 is formed at both distal ends in Z-axis direction on the upper surface of the frame 71. A space between the pair of pillar mounts 72 is longer than the entire length L1 in Z-axis direction of the second collimator module 10B. The pair of pillar mounts 72 is positioned by a pair of mounting blocks 76 (see FIG. 13) and positioning pins formed on the frame 71. Further, a pair of collimator module fixing blocks 79 are positioned and placed by the positioning pins on the pair of pillar mounts 72 inwardly. A pair of the blocks 12 of the second collimator module 10 is fixed to the collimator module fixing blocks 79. The pair of blocks 12 is correctly positioned to the collimator module fixing blocks 79 by the positioning pins 122 shown in FIG. 8. Four connecting pillars 74 are formed at four corners on the lower jig 70A of the jig 70 to support the upper jig 70B having the same configuration of the lower jig 70A. During assembling of the second collimator module 10B, the pair of pillar mounts 72 and the pair of collimator module blocks 79 are removed from the pair of blocks 12 in turn (this will be explained with a flow chart of FIG. 16). Therefore, the pillar mount 72 is formed about 0.1 mm to 3 mm shorter in Y-axis direction than the connecting pillar 74. The pressing component 75 is explained by referring FIGS. 14A, 14B and 14C. FIGS. 14A, 14B and 14C are figures showing the configuration of the pressing component 75. As shown in FIG. 12 and FIG. 13, the pressing component 75 includes a reference board 751, a comb-shaped member 752, and a spring board member 753. FIG. 14A is a front view of the reference board 751. FIG. 14B is a front view of the comb-shaped member 752. FIG. 14C is a front view of the spring board member 753. As shown in FIG. 14A, the reference board 751 is nearly a trapezoidal shape and an upper base 88 or an upper base 89 (dotted line) is formed a curved line. The reference board 751 touches the long sides 11LS of a plurality of collimator single plates 11. The reference board 751 of the pressing component 75 arranged on the lower jig 70A is a reference for the collimator single plate 11 on emission side of X-ray beams. Thus, the upper base 88 of the reference board 751 is a concave curved line. On the other hand, the reference board 751 of the pressing component 75 arranged on the upper jig 70B is a reference for the collimator single plate 11 on incident side of X-ray beams. Thus, the upper base 89 of the reference board 751 is a convex curved line. The reference board 751 has a rigidity for contacting the long sides 11LS of a plurality of collimator single plates 11. The reference board 751 is composed mostly of, for example, iron steel, stainless steel, or aluminum alloy, and the thickness is in a range of 0.5 mm to 3.0 mm. As shown in FIG. 14B, the comb-shaped member 752 is nearly a trapezoidal shape same as the reference board 751. Comb-shaped cutouts 81 into which a plurality of collimator single plates 11 is inserted are formed at the upper base 87. Dozens of cutouts 81 are formed. A width t4 of one cutout 81 is almost the same widths of the width t2 of the first grooves of the block 12 shown in FIG. 5, or the width t3 of the second grooves 135 of the first incident side fixing sheet 13, which is 0.24 mm, for example. The comb-shaped member 752 is made mostly of, for example, iron steel, stainless steel, or aluminum alloy, and the thickness is in a range of 0.5 mm to 3.0 mm. As shown in FIG. 14C, the spring board member 753 includes a spring portion 82 and a first backing plate 85 and a second backing plate 86. This spring portion 82 is a member for pressing a plurality of collimator single plates 11 in X-axis direction. Thus, cutouts 84 are formed on the spring portion 82 with the same pitch of the comb-shaped member 752, and a width t5 of the cutout 84 is same as the width t4 of the cutout 81. In order for the spring portion 82 to be elastically-deformed easily, a depth of the cutout 84 (Y-axis direction) is deeper than of the cutout 81 of the comb-shaped member 752 and a thickness of the spring portion 82 is in a range of 0.1 mm to 0.5 mm. The spring portion 82 is made of, for example, iron steel for spring, stainless steel for spring, or plastics. Since the thickness of the spring portion 82 is thin, the first backing plate 85 and the second backing plate 86 back the spring portion 82 from both surfaces. The pressing component 75 is set where a plurality of cutouts 81 of the comb-shaped member 752 and a plurality of cutouts 84 of the spring board member 753 are layered. In the condition that the cutouts 81 and the cutouts 84 are layered, the long sides 11LS of a plurality of collimator single plates 11 are inserted to the cutouts until the long sides 11LS touch the reference board 75. Although FIG. 14 does not show, the spring portion 82 can align a plurality of collimator single plates 11 at a time in X-axis direction by moving the spring board member 753 in X-axis direction with cams or enlarged holes to the comb-shaped member 752. FIG. 15 shows a supporting block 90 for supporting the collimator single plate 11 when the first incident side fixing sheet 13, the first emission side fixing sheet 15, the second incident side fixing sheet 14, and the second emission side fixing sheet 16 are fixed to the collimator single plate 11. As shown in FIG. 13, the supporting block 90 is attached to the frame 70 of the lower jig 70A. Also, it is not shown in FIG. 13, but when the jig 70 is turned upside down (Y-axis direction), the supporting block 90 is attached to the upper frame 71B of the upper jig 70B. As shown in FIG. 15, the supporting block 90 includes a first sheet mounting portion 91, a second sheet mounting portion 92, and a base portion 99. The first incident side fixing sheet 13 and the first emission side fixing sheet 15 are mounted on top surfaces 93 of the first sheet mounting portion 91. The second incident side fixing sheet 14 and the second emission side fixing sheet 16 are mounted on top surfaces 93 of the second sheet mounting portion 92. A width W11 of the first sheet mounting portion 91 is formed the same width of or relatively wider than the first incident side fixing sheet 13 and the first emission side fixing sheet 15. A width W12 of the second sheet mounting portion 92 is formed the same width of or relatively wider than the second incident side fixing sheet 14 and the second emission side fixing sheet 16. A width W15 of a concave portion 94 is formed wider than the width L4 of the pressing component 75 such that the first sheet mounting portion 91 and the second sheet mounting portion 92 do not touch the pressing component 75 when the supporting block 90 is attached to the frame 71. The width 15 of the concave portion 94 is nearly the same length or relatively narrower than the space L3 between the first incident side fixing sheet 13 and the second incident side fixing sheet 14, or between adjacent second incident side fixing sheets 14 explained in FIG. 13. Now, the assembling method of the collimator module 10 will be explained. The assembling steps of the second collimator module 10B in FIG. 8 are explained as an example of the collimator module 10. FIG. 16 is a flow chart showing the assembling steps of the second collimator module 10B. First of all, the upper jig 70B of the jig 70 (comprising the lower jig 70A and the upper jig 70B) is removed from the jig 70. This enables to assemble a pair of the blocks 12 or the collimator single plate 11 from an upper side. A plurality of cutouts 81 of the comb-shaped member 752 are layered on the cutouts 84 (see FIG. 14 or FIG. 17) of the spring board member 753 of the pressing component 75. In step S111, a pair of the blocks 12 is fixed to the collimator module fixing block 79 of the lower jig 70A. The pair of blocks 12 is positioned at a predetermined position by the positioning pins or the like. In step S112, dozens of the collimator single plates 11 are inserted into the first grooves 125 of the blocks 12 and the pressing components 75. More specifically, the long sides 11LS of the dozens of the collimator single plates 11 touch the upper base 88 of the reference board 751 and are inserted into the plurality of cutouts 81 of the comb-shaped member 752 and the cutouts 84 of the spring board member 753. The short sides 11SS of the dozens of the collimator single plates 11 are inserted into the first grooves 125 of the blocks 12. After the dozens of the collimator single plates 11 are inserted into the cutouts 81 and the cutouts 84, the first incident side fixing sheet 13 and the second incident side fixing sheet 14 are mounted on the long sides 11LS of the collimator single plates 11. In this condition, the long sides 11LS of the collimator single plates 11 are inserted into the second grooves 135, 155, but not yet bonded by an adhesive. Then, the upper jig 70B is attached to the top of the lower jig 70A by the connecting pillars 74. FIG. 17 is a cross-sectional view along the C-C line of FIG. 12 when the upper jig 70B is attached to the top of the lower jig 70A. Also FIG. 17 shows a condition that dozens of the collimator single plates 11 are inserted to the pressing members 75. As shown in FIG. 17, the plurality of the cutouts 81 of the comb-shaped member 752 and the plurality of the cutouts 84 of the spring board member 753 of the pressing component 75 (see FIG. 14 or FIG. 17) are layered. That is, the cutouts are fully opened (see FIG. 18A). In step S113, the dozens of the collimator single plates 11 are arrayed by the pressing components 75. The dozens of collimator single plates 11 are arrayed in X-axis direction by using the lower pressing components 75A and the upper pressing components 75B. The spring board member 753 of the lower pressing components 75A and the upper pressing components 75B move in X-axis direction so that the dozens of the collimator single plates 11 are pushed to the side surfaces 125K (see FIG. 5) on −X side of the first grooves 125 of the block 12. The moving condition of the spring board member 753 is explained with FIG. 18B. FIGS. 18A and 18B are figures for explaining two conditions of the pressing component 75. FIG. 18A shows that the pressing component 75 is fully opened. FIG. 18B shows after the pressing component 75 is moved. For a better understanding, the reference board 751 and the first backing plate 85 and the second backing plate 86 of the spring board member 753 are not shown. In FIG. 18A, the pressing component 75 is “fully opened” in that the cutouts 81 of the comb-shaped member 752 and the cutouts 84 of the spring portion 82 are completely layered. Thus, dozens of the long sides 11LS of the collimator single plates 11 are easily inserted into the pressing components 75. In a condition that the long sides 11LS of the collimator single plates 11 are inserted into the cutouts 81 and the cutouts 84, the spring portions 82 of the lower pressing components 75A and the upper pressing components 75B move to −X-axis direction so that the dozens of collimator single plates 11 inserted into the cutouts 81 and 84 are also moved to −X-axis direction at a time. As a result, the dozens of the collimator single plates 11 are pushed to the side surfaces 125K on −X side of the first grooves 125 of the block 12 as shown in FIG. 5. In step S114 of FIG. 16, the second incident side fixing sheets 14 are bonded to the dozens of the long sides 11LS of the collimator single plates 11. In this process, the second incident side fixing sheets 14 are supported by the supporting block 90 against the collimator single plates 11. Then, the jig 70 is turned upside down. In consequence, the excess adhesive is dropped by gravity and the dropped adhesive is removed by wiping with a cloth. After the adhesive is dried, the first emission side fixing sheets 15 and the second emission side fixing sheets 16 are mounted on the long sides 11LS of the collimator single plates 11. Then, the second emission side fixing sheets 16 are bonded to the dozens of the long sides 11LS of the collimator single plates 11. The jig 70 is again turned back to the original position. The excess adhesive applied on the second fixing sheets on incident side 16 is removed by wiping a cloth. In step S115, one of the pair of blocks 12 is removed from the collimator single plates 11. When the one of block 12 is removed, it is kept fixed to the pillar mount 72 and the collimator module fixing block 79. This is because the side surface 125K on −X side of the first groove 125 of the block 12 acts as a reference for arraying the collimator single plates 11 and it is required for returning the collimator single plates 11 to the original reference position by the pillar mount 72 when it is inserted again. In step S116, an adhesive is applied to the first grooves 125 of the removed block 12 and the block 12 is returned to the original position of the short sides 11SS of the collimator single plates 11. Then, the first grooves 125 of the block 12 and the short sides 11SS of the collimator single plates 11 are bonded. In step S117, the other block 12 is removed from the collimator single plates 11. In step S118, an adhesive is applied to the first grooves 125 of the other block 12 and then it is returned to the short sides 11SS of the collimator single plates 11. Then, the first grooves 125 of the other block 12 and the short sides 11SS of the collimator single plates 11 are bonded. In step S119, the first incident side fixing sheet 13 and the first emission side fixing sheet 15 are bonded to the incident side surface 121 and the emission side surface 12E of the block 12. In step S120, the second collimator module 10B is removed from the jig 70 and then inspected. That is, it is inspected to find whether the dozens of collimator single plates 11 are arrayed correctly and fixed to the blocks 12. The dozens of collimator single plates 11 of the second collimator module 10B are checked from the spaces L3 between the first incident side fixing sheet 13 and the second incident side fixing sheet 14, or between adjacent second incident side fixing sheets 14. By a general camera with a visible light, the dozens of the collimator single plates 11 are checked to determine whether they are arrayed properly. The third collimator module 10C, the forth collimator module 10D, and the fifth collimator module 10E can be made by attaching the third incident side fixing sheet 17 and the third emission side fixing sheet 18, or the supporting sheets 21 and 22 after the second collimator module 10B completed by the above-mentioned assembling steps. Representative embodiments are described above. It will be understood that these embodiments can be modified or changed while not departing from the spirit and scope of them and/or of the appended claims.
058928065	abstract	A spacer for maintaining a pressure tube in spaced relation with a caland tube of a nuclear reactor. The spacer comprising a split ring adapted to be disposed about the outer surface of said pressure tube. The ring has a central annular body portion with a raised bearing surface thereon adapted to contact the inner surface of said calandria tube and prevent contact between said outer surface and said inner surface. An annular land projecting from each side of said central body portion is adapted to receive a collar thereon effective create an interference fit between said ring and said pressure tube and thereby constrain axial movement of said spacer on said pressure tube. The spacer of the present invention maintains its location on the pressure tube and does not suffer the axial movement which characterizes some conventional spacers. The bearing surface can have a coating to reduce wear and heat transfer.
048088315	summary	BACKGROUND OF THE INVENTION I. Field of the Invention This invention relates generally to the measurement of radioactivity of a sample in a container and, more particularly to a new and improved sample container for either wet or dry samples. II. Background Description In the biological and medical sciences, certain radioisotopes are frequently used as tracers in tests and experiments in order to detect minute quantities of certain biochemicals present in test samples. For example, the radioisotope .sup.32 P, is commonly used by researchers in these fields to label genetic material (DNA/RNA) and proteins. Frequently, it is required to know the precise amount of a radioisotope contained in various test samples. Quantitative measurements of the amount of radioactive material present in a test sample usually are expressed as an activity in disintegrations per minute. These measurements provide valuable information both in preparing the radio-labelled chemicals and in measuring an amount of radio-labelled material recovered from a system under investigation. Measurements of the activity of a sample also are needed to limit exposure to personnel handling the radioisotopes. At the present time, most measurements of activity are obtained from scintillation counting or from Geiger counting. Scintillation counting uses photomultiplier tubes to detect photons produced in a scintillation medium in response to absorption by the medium of beta and gamma radiation. Many of the photons emitted from the scintillation medium are incident upon a photocathode of a multiplier phototube. These photons are converted to photoelectrons and are multiplied in number at a succession of phototube electrodes, called dynodes, the output of which is a measurable electrical pulse related to the incident radiation. Liquid scintillation counters operate on the same basic principle as scintillation counters, except that the scintillation medium is a liquid into which is dissolved, suspended or otherwise intermixed the radioactive sample being tested. Radioactive emissions of a sample are measured by collecting photons emitted from the scintillation medium and generating photoelectrons responsive thereto to produce electrical pulses related to the incident beta and gamma radiation. Scintillation and liquid scintillation counting require special sample preparation and the use of special sample containing vials in order to provide a quantitative measure of the amount of radioactive material present in a particular sample. Accordingly, an extra material handling step, involving a transfer of radioactive material into one of the special vials, is required when using these techniques. This transfer step is undesirable, for it is accompanied by an element of error in the measurement of material transferred to the vials. When this measurement error is added to the error inherent to the particular experiment or test technique being utilized, further uncertainty as to the accuracy of the quantitative data obtained from the sample results. Furthermore, the preparation of even a small amount of material for scintillation counting results in the loss of that material for further experimentation. In many cases, where only a very limited quantity of material is available this loss may be unacceptable. Geiger counters are generally used when counting small numbers of samples. These counters provide a simpler but much less reliable means for measuring an approximate activity of a radiation emitting sample. Geiger counters use gas filled tubes the contents of which are ionized by incident radiation to produce an electronic signal which registers on a meter or in an audio circuit. The magnitude of the electronic signal is proportional to the amount of radiation impinging upon the gas filled tubes. Commercial Geiger counters are generally hand held devices whose quantitative accuracy is limited by uncertainties in the geometrical positioning of the sample relative to the detector and the absence of careful calibration techniques. However, the instruments are very useful in determining the presence and/or location of radioactivity and in determining an approximate activity of the sample for safe handling considerations. Geiger counters are also helpful in assessing the progress of certain chemical reactions or experiments. At the current time, the vast majority of low energy radioisotope samples are counted using the technique of scintillation counting. The lowest energy samples such as tritium (.sup.3 H), carbon-14 (.sup.14 C), sulfur-35 (.sup.35 S), and phosphorus-32 (.sup.32 P) are counted using liquid scintillation counting (LSC). Liquid samples to be counted are placed in standard LSC vials, mixed with scintillation chemicals which fluoresce when excited by the low energy radiation, and counted in an instrument which detects the light flashes produced inside the vials. In the case of the highest energy of these isotopes (.sup.32 P), Cerenkov light is also produced by the interaction of the radiation with water or glass without the addition of a special scintillation chemical, and this light can also be counted by the LSC instrumentation. For higher energy radiation, such as gamma emitters iodine-125 (.sup.125 I) and technetium-99m (.sup.99m TC), samples can be counted with a solid scintillation crystal. The gamma radiation is much more penetrating than the beta radiation of the isotopes cited above and can penetrate the walls of a container and enter a crystal detector which produces light in response to the emitted radiation. Again, the light flashes are counted by the instrumentation. These techniques require samples in large standard vials (from approximately 10-25 cc volume). The vials are generally made from plastic or glass of sufficient thickness to absorb essentially all of the low energy beta emissions from carbon-14 and sulfur-35. For these low energy beta emitting isotopes, the samples must be mixed with a scintillation chemical to produce the desired light flashes. This chemical also destroys most biological activity, rendering the counted samples unsuitable for further biological experimentation. There are also problems of compatibility between sample and scintillant including phase separation and coloration which must be eliminated or measured to obtain accurate results. In addition, the samples once counted must be disposed of, and there are significant volumes of liquid radioactive waste generated. Another type of sample preparation in the prior art is the planchette used with a planchette counter, proportional and Geiger counters. In these devices, the liquid sample is spread and dried on a metal or paper surface. After drying, it can then be introduced into the sensitive gas volume of a proportional or Geiger counter which counts the radioactivity. A currently available detector is a compact, bench top radiation detection apparatus capable of measuring the radiation in samples placed in any of a plurality of different sized and shaped containers through the provision of removable sample holders configured to receive different sample containers. Thus, the apparatus can measure the radiation from sample containers. Thus, the apparatus can measure the radiation from samples in the form of a liquid in a vial when the energy of the emissions are sufficient to penetrate the wall of the vial, as well as from samples which have been deposited and dried on the surfaces of specially shaped disposable sample containers which can be inserted into an associated sample holder and positioned at a fixed distance from a radiation detector with the sample in direct communication with the detector. This latter arrangement enables an efficient and accurate measurement to be made due to the absence of a container wall between the sample and the detector. It also allows the apparatus to be used to detect and measure low energy emissions. The vials however, have the limitations mentioned above. The foregoing illustrates limitations known to exist in present devices. Thus, it is apparent that it would be advantageous to provide an alternative directed to overcoming one or more of the limitations set forth above. Accordingly, a suitable alternative is provided including features more fully disclosed hereinafter. OBJECTS AND SUMMARY OF THE INVENTION In view of the foregoing, it is an object of the present invention to reduce the amount of radioactive waste which must be disposed of. It is another object of the present invention to limit the amount of handling, mixing or addition of other toxic chemicals in the sample container. It is yet another object of the present invention to enhance retention of the biological activity of the samples and recovery for use in further biological experimentation. It is a further object of the present invention to provide a sample container with a sufficiently thin exit window for permitting significant amounts of radioactive emissions to penetrate out of the container and into a direct ionization detector. It is a still further object of the present invention to provide a sample holder in which either a wet or a dry sample can be sealed and analyzed. In one aspect of the present invention, this is accomplished by providing a container for radioactive samples including a carrier having an aperture formed therein. Means connected to one side of the carrier and forming a cavity therewith are provided for retaining a liquid or dry radioactive sample in the cavity and for simultaneously permitting radiation from the sample to exit therethrough. The cavity is accessible via the aperture in the carrier. Other means are provided to be removably attached to the carrier for sealing the aperture. The foregoing and other aspects will become apparent from the following detailed description of the invention when considered in conjunction with the accompanying drawing Figures. It is to be expressly understood, however, that the drawing is not intended as a definition of the invention but is for the purpose of illustration only.
	description	This invention relates generally to the field of energy devices and more specifically to energy conversion devices. Radioactive decay is the process in which an unstable atomic nucleus spontaneously loses energy by emitting ionizing particles and radiation. This decay, or loss of energy, results in an atom of one type, called the parent nuclide, transforming to an atom of a different type, named the daughter nuclide. A nuclide is an atomic species characterized by the specific constitution of its nucleus, i.e., by its number of protons, its number of neutrons, and its excited state. Isotopes are different types of atoms of the same chemical element, each having the same number of protons and a different number of neutrons. According to one embodiment, an energy conversion device comprises a nuclear battery, a light source coupled to the nuclear battery and operable to receive electric energy from the nuclear battery and radiate electromagnetic energy, and a photocell operable to receive the radiated electromagnetic energy and convert the received electromagnetic energy into electric energy. The nuclear battery comprises a radioactive substance and a collector operable to receive particles emitted by the radioactive substance. Certain embodiments of the invention may provide one or more technical advantages. A technical advantage of one embodiment may include the capability to provide a longer-life battery that can be used in roughly the same mechanical and electrical manner as a conventional electrochemical battery. Another technical advantage of one embodiment may include the capability to provide low-voltage power from a nuclear power source. Yet another technical advantage of one embodiment may include the capability to provide power over an extended lifetime and eliminate the need of an on/off switch from some electronic devices. Various embodiments of the invention may include none, some, or all of the above technical advantages. One or more other technical advantages may be readily apparent to one skilled in the art from the figures, descriptions, and claims included herein. It should be understood at the outset that, although example implementations of embodiments of the invention are illustrated below, the present invention may be implemented using any number of techniques, whether currently known or not. The present invention should in no way be limited to the example implementations, drawings, and techniques illustrated below. Additionally, the drawings are not necessarily drawn to scale. Electrochemical Cell Batteries Electric power in batteries may be provided through electrochemical reactions inside an enclosure. Examples include AA and AAA batteries. Once a resistive load is attached to one or more of these batteries, they will discharge their electrochemical potential into that load. As that energy is dissipated into that load, the energy stored in the battery will deplete over an amount of time proportional to the load placed on the battery. A typical AA battery may have a maximum energy storage capacity of perhaps 10,000 Joules, which will be consumed over some period of time. Once depleted, the electrochemical cell battery must be either replaced or recharged. This process is very costly. Accordingly, teachings of certain embodiments recognize the capability to provide a longer-life battery that can be used in roughly the same mechanical and electrical manner as a conventional electrochemical battery. Radioactive Decay Process A radioactive substance is an unstable isotope which will be permanently transformed by some nuclear decay process. Example decay processes include alpha decay and beta decay. When that transformation occurs, an unstable isotope (in alpha and beta decay) will emit radiation in the form of a particle and in the process lose some energy, which will be taken out of that nucleus by the emitted particle. The nucleus will tend towards a more stable state; although the new nucleus may still be radioactive it is one step closer in the decay chain toward a stable isotope. Alpha radiation is a form of nuclear fission in which an atom splits into two smaller atoms and releases kinetic and electrical energy. Alpha particles (the emitted particle involved in alpha radiation) are identical to the nucleus of a helium atom (two protons and two neutrons) with no attached electrons. There are many known sources of alpha radiation. One example is Americium-241, which is used in many types of smoke detectors. Beta radiation is a form of nuclear decay in which an atom releases a beta particle. Beta radiation comes in two forms, beta minus which is the release of electrons, and beta plus which is the release of positrons. There are many natural sources of beta radiation in nature. One example is as tritium, an isotope of hydrogen. Nuclear Power Sources FIG. 1 shows an energy conversion system 100 according to one embodiment. Energy conversion system 100 converts nuclear energy into electrical energy. Energy conversion system 100 features an energy source 110 and an energy collector 120. Energy source 110 includes a substrate 112, a coating 114, and an emitting layer 116. In some examples, substrate 112 may be either a conductor, such as a metallic substrate, or an insulator, such as a plastic substrate. In the illustrated example, substrate 112 is a 250 μm-thick, stainless steel substrate, coating 114 is a 5 μm-thick polyimide coating, and emitting layer 116 is a 3 μm-thick nuclear source layer. In this example, emitting layer 116 is a radioactive substance that undergoes a nuclear decay process. Energy collector 120 includes a substrate 122 and a coating 124. In the illustrated example, substrate 122 is a 250 μm-thick, stainless steel disk, and coating 124 is a 25 μm-thick polyimide coating. In some embodiments, energy source 110 and energy collector 120 are separated by some distance. Teachings of certain embodiments recognize that energy source 110 may emit particles that travel across this distance before striking energy collector 120. The distance may be of any suitable length. For example, teachings of certain embodiments recognize that the length of the distance may be optimized based on a desired capacitance across the distance. Also, teachings of certain embodiments recognize that the maximum length of the distance may be limited to the distance that particles can travel and still strike energy collector 120. In operation, emitting layer 116 emits charged particles that strike a conductor, such as substrate 122. In the illustrated embodiment, the emitting layer 116 is applied to a second conductor, substrate 112. In some examples, substrate 112 acts as a support structure only; in other embodiments, substrate 112 may act as a second collector. As a result of electrical charge transfer, substrate 122 will acquire an electrical charge, and energy conversion system 100 will generate electric power 150. In this example, energy source 110, which has an electrical charge, emits a particle such as a beta particle, and energy collector 120 struck by the particle acquires the opposite charge. This opposite charge may be opposite in both sign and magnitude. Teachings of certain embodiments recognize that the two electrodes may maintain a difference of potential for as long as charged particles are emitted from energy source 110 and these emissions strike energy collector 120. An electrical load may be attached to the two electrodes so that the separated charge from one electrode may pass through the load and recombine with the charge on its companion electrode. In this manner, energy source 110 that spontaneously emits charged particles may be made to provide an electric power source. Accordingly, teachings of certain embodiments recognize that a nuclear ‘battery’ may have many applications including the replacement of many disposable and rechargeable batteries now used in commercial and military applications. However, for optimal efficiency, the voltage difference across the electrodes of a typical nuclear battery should be on the order of the energy of the particles emitted from the nuclear material. This voltage is typically very high, and such nuclear batteries tend to provide extremely high output voltage and extremely low output current. For example, Tritium is a nuclear isotope that emits beta particles and photons with the beta particle carrying away roughly 5700 electron volts of kinetic energy. In order to obtain the maximum conversion efficiency of a nuclear battery based on the isotope of Tritium, the electrodes in this battery should be spaced so that they may be charged, via the absorption of the beta emissions, to slightly less than 5700 volts. If this is not the case, the particle will retain excess kinetic energy when it strikes the collection electrode; this excess kinetic energy will be lost as heat, light, or another form of energy. In addition to optimizing the voltage difference based on the order of the energy of the particles emitted from the nuclear material, one may improve nuclear battery efficiency by matching the resistance of the load connected to the nuclear battery to the internal resistance of the battery. For example, if a layer of beta-emitting tritium material generates an assumed 1 milliampere of current, the internal resistance of an optimized tritium battery would be 5700 volts/0.001 amperes=5.7 megohms. Thus, optimal efficiency for power transfer from the battery would require the load to have a resistance of the same value (a matched load) of 5.7 megohms. However, many electrical devices require a much lower operating voltage and have a much lower load resistance. A typical AA battery has an output voltage of 1.5 volts and an internal resistance of less than 1 ohm. Accordingly, teachings of certain embodiments recognize the capability to provide an efficient battery based on nuclear emissions with an output voltage commensurate with the requirements of modern battery-powered devices. Thus, teachings of certain embodiments recognize the capability to efficiently convert the high voltage output of a nuclear power source into lower voltage and lower effective internal resistance. FIG. 2 shows an energy conversion system 200 according to one embodiment. Energy conversion system 200 converts nuclear energy into electrical energy 250. Energy conversion system 200 features an energy source 210, one or more energy collectors 220, a light source 230, and a photocell 240. In this example, energy source 210 is a radioactive substance that undergoes a nuclear decay process. Energy collectors 220 may represent any conductors configured to detect particles emitted by energy source 210. Examples of energy source 210 and collectors 220 may include energy source 110 and energy collector 120 of FIG. 1. Light source 230 may represent any device operable to receive electric power from energy source 210 and/or energy collectors 220 and radiate electromagnetic energy, such as light. Examples of light source 230 may include light-emitting diodes (LEDs); laser diodes, such as vertical-cavity surface-emitting lasers (VCSELs); transverse electric lasers; and electroluminescence devices. Electroluminescence is an optical phenomenon and electrical phenomenon in which a material emits light as the result of radiative recombination of electrons and holes in a material, such as a semiconductor. Light source 230 may include one or more individual light sources. For example, light source 230 may include multiple LEDs connected in a series such that the forward voltage of the series of diodes is equal to the electrical output voltage of energy source 210. Photocell 240 may represent any device suitable to receive the radiated electromagnetic energy and convert the radiated electromagnetic energy into electric power 250. One example of photocell 240 may include a photovoltaic cell, such as a solar cell. In operation, energy source 210 emits charged particles that strike energy collectors 220. As a result of electrical charge transfer, energy collectors 220 will acquire an electrical charge generate electric power. Light source 230 will receive this electric power and radiate electromagnetic energy. Photocell 240 receives the radiated electromagnetic energy and converts the electromagnetic energy into electric power 250. Energy conversion system 200 may be constructed in any suitable manner. For example, FIGS. 3A, 3B, and 3C show a cylindrical energy conversion system 300 according to one embodiment. Energy conversion system 300 converts nuclear energy into electrical energy 350. Energy conversion system 300 features an energy source 310, one or more energy collectors 320, a light source 330, and a photocell 340. In this example embodiment, energy conversion system 300 is housed in a cylindrical body 360. Body 360 may be formed out of any suitable material. For example, in some embodiments, body 360 is a conductor, such as a metal body, or an insulator, such as a plastic body. In some embodiments, body 360 may act as an outer conductor and shield, absorbing particles from a radioactive material such as energy source 310 and shielding those particles from escaping outside energy conversion system 300. In this example embodiment, energy source 310 and energy collectors 320 are shown as layers of material wrapped inside of body 360. For example, energy source 310 and energy collectors 320 may be rolled around each other and fit within body 360, as shown in FIG. 3A. However, teachings of certain embodiments recognize that energy source 310 and energy collectors 320 may be disposed within body 360 in any suitable manner. Energy sources 310 and energy collectors 320 may be paired and arranged in any suitable manner. For example, the embodiment shown in FIG. 3B features energy source 310 disposed between two layers of energy collectors 320. In this example, when the energy source 310 and energy collectors 320 are rolled inside body 360, a cross-section of the roll would appear as follows: energy collector 320, energy source 310, energy collector 320, energy collector 320, energy source 310, energy collector 320, etc. Each energy collector 320 is situated between an energy source 310 and another energy collector 320, and each energy collector 320 only absorbs particles from the one adjacent energy source 310. In another example, the embodiment shown in FIG. 3C features energy source 310 disposed next to one layer of energy collectors 320. In this example, when the energy source 310 and energy collector 320 are rolled inside body 360, a cross-section of the roll would appear as follows: energy collector 320, energy source 310, energy collector 320, energy source 310, energy collector 320, energy source 310, etc. Each energy collector 320 is situated between two energy sources 310, and each energy collector 320 absorbs particles from both adjacent energy sources 310. FIGS. 3A, 3B, and 3C show example embodiments of layers of energy sources 310 and energy collectors 320 rolled inside of a cylindrical body 360. However, teachings of certain embodiments recognize that embodiments are not limited to a cylindrical body 360, but rather energy conversion device 300 may be of any suitable shape and dimensions. Additionally, teachings of certain embodiments recognize that energy sources 310 and energy collectors 320 are not limited to being rolled together. Rather, energy sources 310 and energy collectors 320 may be arranged in any suitable manner. For example, FIG. 4 shows stacked energy sources 410 and energy collectors 420 according to one embodiment. Unlike the examples shown in FIGS. 3A, 3B, and 3C, which features a single energy source 310 wrapped inside body 360, FIG. 4 shows multiple, discrete energy sources 410 stacked between multiple, discrete energy collectors 420. Energy sources 410 and energy collectors 420 may be of any suitable shape or size. For example, in one embodiment, energy sources 410 and energy collectors 420 may be circular so as to fit inside the cylindrical body 360 of FIG. 3A. In another example embodiment, energy sources 410 and energy collectors 420 may be square or rectangular so as to fit inside a square or rectangular energy conversion device. FIG. 5 shows an energy conversion system 500 according to one embodiment. Energy conversion system 500 features energy source 310, one or more energy collectors 320, a light source 330, and a switch 510. Switch 510 may represent any switch configured to turn on and/or turn off energy conversion system 500. For example, in one embodiment, switch 510 may turn on and/or turn off light source 330. In some embodiments, energy conversion system 500 does not feature a switch 510. Teachings of certain embodiments recognize that energy conversion system 500 may provide electromagnetic light energy for longer periods of time than traditional light sources, such as flashlights powered by alkaline, zinc-carbon, or lithium batteries. However, teachings of certain embodiments are not limited to flashlights; rather, energy conversion system 300 may be used to transmit light in any suitable environment for any suitable use. In the embodiment of FIG. 5, energy conversion system 500 is cylindrical. However, teachings of certain embodiments recognize that energy conversion system 500 may be any suitable shape. Modifications, additions, or omissions may be made to the systems and apparatuses described herein without departing from the scope of the invention. The components of the systems and apparatuses may be integrated or separated. Moreover, the operations of the systems and apparatuses may be performed by more, fewer, or other components. The methods may include more, fewer, or other steps. Additionally, steps may be performed in any suitable order. Additionally, operations of the systems and apparatuses may be performed using any suitable logic. As used in this document, “each” refers to each member of a set or each member of a subset of a set. Although several embodiments have been illustrated and described in detail, it will be recognized that substitutions and alterations are possible without departing from the spirit and scope of the present invention, as defined by the appended claims. To aid the Patent Office, and any readers of any patent issued on this application in interpreting the claims appended hereto, applicants wish to note that they do not intend any of the appended claims to invoke paragraph 6 of 35 U.S.C. §112 as it exists on the date of filing hereof unless the words “means for” or “step for” are explicitly used in the particular claim.
	summary
	abstract	A method for automatically identifying defects in turbine engine blades is provided. The method comprises acquiring one or more radiographic images corresponding to one or more turbine engine blades and identifying one or more regions of interest from the one or more radiographic images. The method then comprises extracting one or more geometric features based on the one or more regions of interest and analyzing the one or more geometric features to identify one or more defects in the turbine engine blades.

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