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	abstract	A MEMS differential actuated nano probe includes four suspension beams arranged in parallel, a connecting base connecting to the suspension beams, a nano probe. Two of the suspension beams elongate due to thermal expansion to allow the deflection of the probe. By heating the suspension beams at different positions, the MEMS differential actuated nano probe can move in two directions with two degrees of freedom. The deflection of the MEMS differential actuated nano probe can be also achieved in piezoelectric or electrostatic way.
	summary
054426677	abstract	A reactor pressure vessel for a nuclear reactor includes an upper part and a lower part. At least the lower part has first subregions with a lesser wall thickness and second subregions with a greater wall thickness. The lesser wall thickness is selected for a rated operation. The first subregions are formed by recesses in the outer surface of the wall.
	summary
	description	The present invention relates to a particle beam therapy system that performs therapy by irradiating a particle beam onto a diseased site of a patient, and particularly to a particle beam therapy system that can perform irradiation with an irradiation shape conforming to a diseased site, by use of a multileaf collimator. Some particle beam therapy systems shape an irradiation field through a multileaf collimator so as to irradiate a particle beam in an irradiation shape based on a treatment plan. In such a particle beam therapy system, in the case where during irradiation of a particle beam, the opening shape of the multileaf collimator deviates from the shape specified by a treatment plan, the dose given to a diseased site, which is an irradiation subject, and the distribution of doses differ from the treatment plan; therefore, it is required to rapidly stop the beam irradiation. Accordingly, monitoring, (confirmation) of the leaf positions, which specify the opening shape, and monitoring of the patient position are important functions for realizing the dose set by the treatment plan and need to be of high redundancy and of high multiplicity; thus, a plurality of detection methods are utilized at the same time. The leaf position detection methods can roughly be divided into a built-in type detection mechanism integrated in each of the leaves or a driving unit corresponding to each leaf and an external (optical) detection method. As the built-in type detection mechanism, for example, there exists a method in which an encoder is provided in a driving motor for driving the leaf and from the number of revolution of the motor, the leaf position is detected. However, in the case of the built-in type detection mechanism, a detecting member needs to be mounted on each leaf or the driving unit corresponding, to each leaf; thus, the complexity of the system is raised. Accordingly, in the case where a plurality of detection methods is redundantly utilized, as one of them, there is adopted, in many cases, an optical (image) type detection method. For example, there has been proposed a particle beam therapy system (for example, refer to Patent Document 1) in which there are provided a built-in type detection mechanism such as an encoder, and there is further provided an optical detection mechanism that takes an image of the opening shape of a leaf by use of a mirror provided in the irradiation field and a camera provided outside the irradiation field and that measures the position of the leaf from the image, so as to monitor the opening shape of the multileaf collimator in a redundant and multiple manner. Moreover, as another optical detection method, there is proposed a particle beam therapy system (for example, refer to Patent Document 2) in which a screen is provided in the irradiation field and the opening shape projected on the screen is photographed by a camera. Furthermore, there is also proposed a particle beam therapy system (for example, refer to Patent Document 3) in which a collimated laser beam is irradiated onto outer side surface of a leaf and, from the irradiation position, the position of the leaf is detected. [Patent Document 1] Japanese Patent Application Laid-Open No. 2008-295860 (paragraph 0004 and 0021 through 0022, FIGS. 2 and 9) [Patent Document 2] International Publication No. WO2010/073318A1 (Paragraphs 0015 through 0018, FIG. 1) [Patent Document 3] Japanese Patent Application Laid-Open No. 2010-104452 (Paragraphs 0037 through 0041, FIGS. 6 through 8) However, in such a particle beam therapy system as disclosed in Patent Document 1 or 2, a foreign material such as a mirror or a screen is included in the irradiation field; therefore, in some cases, because a charged particle beam collides with the mirror or the screen and its energy is reduced, the range of the particle beam is changed or the particle beam is shut off to disperse, whereby irradiation conforming to a treatment plan cannot be performed. Moreover, in the case of such a particle beam therapy system as disclosed in Patent Document 3, it is required to mount a light source and an image-capturing unit outside the movable range of the leaf so that the light source and the image-capturing unit can face the outer side surface of the leaf; therefore, the image capturing device sticks out of the multileaf collimator main body, whereby the system upsizes. In particular, in the case where the light source and the image-capturing unit are provided in the rotating gantry, there has been a problem, for example, that because the moving body increases in volume, it is difficult for the rotating gantry to rotate smoothly and perform multi-port irradiation. The present invention has been implemented in order to solve the foregoing problems; the objective thereof is to obtain a particle beam therapy system that suppresses dispersion of a charged particle beam, reduction of the energy thereof, and upsizing of the system from being caused by the image capturing unit, and that can accurately monitor the opening shape of the multileaf collimator so as to perform high-accuracy particle beam therapy. A particle beam therapy system according to the present invention is provided with an irradiation nozzle that scans a particle beam supplied from an accelerator and irradiates the particle beam in such a way as to enlarge an irradiation field; a multileaf collimator in which a pair of leaf rows, each of them is composed of a plurality of leaf plates laminated in thickness direction, is arranged in such a way as to interpose a beam axis of the particle beam, in which respective side faces, of the plurality of leaf plates, that face the beam axis are driven in approaching or in departing direction with respect to the beam axis so that a predetermined opening shape is formed, and that forms the particle beam emitted from the irradiation nozzle in such a way that the particle beam conforms to an irradiation subject and then emits the particle beam; and an image-capturing unit that takes images of outer ends, of respective downstream side faces of the plurality of leaf plates in irradiation direction of the particle beam, that are distal with respect to the beam axis, wherein the image-capturing unit is provided for each of the pair of leaf rows, in such a way as to be situated at a position that is at outer side of the irradiation field of the particle beam that has passed through the multileaf collimator and at downstream of the multileaf collimator; and the image-capturing unit is adjusted in such a way that foot of a perpendicular from a viewpoint to the downstream side face, is situated at a position that is at inner side of the position of the outer end when the leaf plate is maximally driven in the departing direction. In a particle beam therapy system according to the present invention, the arrangement of devices for taking an image of a multileaf collimator is appropriately adjusted; thus, the irradiation field is not blocked off, and the system is not upsized. As a result, there can be obtained a particle beam therapy system that accurately monitors the opening shape of the multileaf collimator and can perform high-accuracy particle beam therapy. FIGS. 1 and 2 are diagrams for explaining the configuration of a particle beam therapy system according to Embodiment 1 of the present invention; FIG. 1 is a diagram illustrating the configuration of the whole irradiation system in a particle beam therapy system; FIG. 2 is a diagram illustrating the positional relationship between a multileaf collimator provided in a particle beam therapy system and an image-capturing unit for taking an image of the multileaf collimator. FIG. 3 is a set of diagrams illustrating examples of adjustment of the viewpoint of an image capturing unit (camera) for accurately detecting the position of a leaf, in a particle beam therapy system according to Embodiment 1 of the present invention. FIG. 4 is a set of diagrams for explaining the positional relationship between the leaf plate of a multileaf collimator and the viewpoint of an image-capturing unit in a variant example of particle beam therapy system according to Embodiment 1 of the present invention. The largest characteristics of a particle beam therapy system according to Embodiment 1 of the present invention are the installation and the configuration of a shape monitoring device (the image-capturing unit thereof) for detecting the opening shape (positions of respective leaves) of a multileaf collimator. However, before the detailed explanation for the characteristics, there will be explained the overall configuration of the irradiation system in a particle beam therapy system provided with a multileaf collimator. As illustrated in FIG. 1, a particle beam therapy system 2 is provided with a dose monitor 3 that measures the dose of a charged particle beam B supplied from an unillustrated accelerator, a set of wobbler electromagnets 4 that functions as an irradiation nozzle for enlarging an irradiation field RB by circularly scanning the supplied charged particle beam B, a scatterer 5 that is formed of lead or the like and scatters the charged particle beam B, a ridge filter 6 that is formed of aluminum or the like and enlarges the width of a Bragg peak in accordance with the thickness of the irradiation subject, a range shifter 7 that is formed of an acrylic resin or the like and changes the energy (range) of the charged particle beam B in accordance with the depth (irradiation depth) from the body surface of an irradiation subject Ka, a multileaf collimator 8 that is configured with a leaf unit formed of a plurality of leaf plates and a leaf drive mechanism for driving each of leaf plates and that limits an irradiation field in such a way that the irradiation field coincides with the shape of a diseased site, and a bolus 9 that limits the range of the charged particle beam B in such a way that the range coincides with the depth-direction shape of an irradiation subject. The particle beam therapy system 2 is further provided with video cameras 11a and 11b (collectively referred to as a video camera 11) that are arranged at both respective outer sides that are situated opposite to the opening portion of the multileaf collimator 8 and serve as a shape monitoring device for monitoring an opening shape PS of the multileaf collimator 8, and image processing units 13a and 13b (collectively referred to as an image processing unit 13) that perform processing on respective images taken by the video cameras 11a and 11b; the processed images are displayed, as shape images Da and Db (collectively referred to as a shape image D), by display devices 14a and 14b (collectively referred to as a display device 14), respectively. In FIG. 1, a patient K who undergoes a particle beam therapy is illustrated as a cross section thereof; in the cross section, there are illustrated a diseased site Ka, as an irradiation subject, and a patient position marker Kb for tracking the position of the diseased site Ka being treated. Next, the operation of enlarging and shaping the irradiation field RB in the irradiation system will be explained. The charged particle beam B, accelerated by an unillustrated accelerator and supplied by way of a transport system, enters the dose monitor 3, and then the irradiation dose thereof is counted. At this moment, the charged particle beam B that has entered the irradiation system is a so-called pencil beam having a diameter the same as or smaller than several millimeters; the charged particle beam B is scanned by the wobbler electromagnet 4 in such a way as to draw a circular orbit and is scattered by the scatterer 5, so that the irradiation field RB is enlarged. The wobbler electromagnet 4 is usually provided with an x-direction electromagnet and a y-direction electromagnet; the two electromagnets are arranged in such a way as to lie in series along the center axis XB of the charged particle beam B. Here, for clarity of description, the x direction and the y direction will be defined. The direction (XB direction) in which the charged particle beam B travels is defined as the positive direction of the z axis. The x axis and the y axis are axes that are perpendicular to the z axis; the x axis and the y axis are also perpendicular to each other. In FIGS. 1, 2, and thereafter, the x direction is defined as the driving direction of a leaf 8L of the multileaf collimator 8; the y direction is defined as the thickness (laminate) direction of the leaf plate 8L. Then, the irradiation field RB is expanded in the xy direction (in the plane direction) by the wobbler electromagnet 4 and the scatterer 5. The charged particle beam B whose irradiation field RB has been enlarged passes through the ridge filter 6. The ridge filter is formed, for example, in such a way that a great number of cones or plates whose cross sections are triangles are arranged on a plane; assuming that, for example, the irradiation field RB is divided into a great number of sub-areas, there exist beams B that pass through different thicknesses from one another. For easier understanding, FIG. 1 or 5 illustrates triangular prisms that are arranged in such as way as to be in a row. The Bragg peak is enlarged in such a manner as described above, so that the irradiation field has a predetermined width of SOBP (Spread-Out Bragg Peak). That is to say, the ridge filter 6 enlarges the irradiation field also in the z direction. Next, the charged particle beam B whose irradiation field has been enlarged passes through the range shifter 7. The range shifter 7 is a device that changes the energy of the charged particle beam B. Adjusting the range of the charged particle beam B by the range shifter 7, irradiation of the charged particle beam B (dose delivery) can be performed onto a position of a desired inner-body depth. As described above, the irradiation field is enlarged not only in the plane direction (x, y) but also in the thickness direction (z); however, in this DESCRIPTION, only the enlargement in the plane direction (x, y) is illustrated as the irradiation field RB. As an example of the orbit created by wobbler electromagnet, a circular orbit has been described; however, the orbit is not limited to the foregoing circular orbit but may a spiral orbit or another orbit, as long as it enlarges the irradiation field RB. Next, the charged particle beam B passes through the multileaf collimator 8. The multileaf collimator 8 is configured with two leaf rows 8C, obtained by laminating tabular leaf plates 8L in the thickness direction (y direction), that are arranged so as to interpose the beam axis XB and to face each other (in FIG. 1, the right row is indicated by 8Ca, and the left row is indicated by 8Cb). Each of the leaf plates 8L is positioned at a predetermined position along a direction that approaches to or departs from the beam axis XB within the direction (x) perpendicular to the thickness direction, so that the desired opening shape PS is formed. After that, the irradiation field RB of the charged particle beam B that has passed through the multileaf collimator 8 is limited in accordance with the shape of the diseased site Ka, by means of the opening shape PS conforming to the shape of the diseased site Ka. That is to say, the multileaf collimator 8 performs limitation and formation of the irradiation field RB in the xy direction. The multileaf collimator 8 is provided with at least a cluster of leaf plates 8L and an unillustrated leaf drive mechanism. However, if the leaf drive mechanism itself is drawn in a figure, it becomes difficult to illustrate the arrangement of the leaf plates 8L and the like; therefore, in FIGS. 1, 2, and thereafter, for the sake of simplicity, a cluster of leaf plates 8L is illustrated or a discrete leaf plate 8L is extracted and illustrated. Lastly, the charged particle beam B passes through the bolus 9. The bolus 9 is a limiter that is formed of resin or the like; it is formed in such a shape as to compensate the depth-direction shape of the diseased site, for example, the distal shape of the diseased site. The distal shape denotes the depression-protrusion contour of the deepest portion KaD. In this situation, the energy of the irradiation field is limited (formed in the z direction) to have a shape the same as the distal shape. That is to say, the bolus 9 performs limitation and formation of the irradiation field in the z direction. In the case where irradiation utilizing the foregoing irradiation system is performed according to the multi-layer conformal irradiation method, special dose delivery is divided in the depth (z) direction and the dose is given. When irradiation is started, the wobbler electromagnet 4, the range shifter 7, and the multileaf collimator 8 (the opening shape of the multileaf collimator) are set in accordance with the dose delivery for the layers (slices) including the deepest portion KaD, and the charged particle beam B is irradiated onto the diseased site Ka. After the irradiation onto the deepest layer (slice) is completed, the range is automatically adjusted by the range shifter 7 so as to be situated at a position that is shallower (closer to the irradiation source) by a depth corresponding to the Bragg peak width, and the setting of the wobbler electromagnet 4 and the multileaf collimator 8 are also changed; then, irradiation onto the next layer is performed. After that, similarly, the range is adjusted by the range shifter 7, and while the setting of the wobbler electromagnet 4 and the multileaf collimator 8 are changed, the dose optimized for the shape of the diseased site Ka is delivered as a whole. In order to perform a particle beam therapy in which the level of coincidence (conformity) between the therapy volume and the shape of the diseased site Ka is high, i.e., a high-accuracy particle beam therapy, through such multi-layer conformal irradiation as described above, it is required to confirm and monitor whether or not the opening shape PS of the multileaf collimator 8 is the same as the preliminarily set shape for each irradiation layer (slice). Thus, in the particle beam therapy system 2 according to Embodiment 1, in order to securely monitor the opening shape PS of the multileaf collimator 8, the video camera 11 is disposed as illustrated in FIG. 2. FIG. 2 illustrates the position of the video camera 11 (the case itself and the viewpoint PV) and the irradiation coverage (irradiation field RB), of the charged particle beam B, that is specified by the driving range of the leaf plate 8L and the leaf plate 8L in the right side of the beam axis XB, when the multileaf collimator 8 is viewed in the y direction. The leaf plate 8L basically has four side faces, i.e., an inner side face AI that is approximately parallel to the traveling direction of the charged particle beam B and faces the beam axis XB, an upstream side face AU that is approximately perpendicular to the traveling direction of the charged particle beam B and is situated at the upstream side (incident side) of the traveling direction of the charged particle beam B, an outer side face AX that is approximately parallel to the traveling direction of the charged particle beam B and is situated at the distal side of the beam axis XB, and a downstream side face AD that is approximately perpendicular to the traveling direction of the charged particle beam B and is situated at the downstream side (emitting side) of the traveling direction of the charged particle beam B. In Embodiment 1, the leaf plate 8L has a rectangular tabular shape, as illustrated in FIG. 2; each leaf plate 8L is driven in such a way that the position thereof can freely be set in the direction (x) that is perpendicular to the beam axis XB (z) and the lamination (thickness) direction (y) of the leaf plate 8L and is parallel to the side faces AU and AD. The inner side face AI is driven up to LOI when it departs most from the beam axis XB; at this moment, the irradiation coverage of the charged particle beam B becomes maximum (RBMAX). The case of the video camera 11 is provided at a position that is outside the maximum irradiation coverage RBMAX and through which the charged particle beam B does not pass, whatever opening shape PS is formed. Furthermore, the visual field FV of the video camera 11 has the range covering the driving range of the outer side face AX of the leaf plate 8L, i.e., the range from the most approaching position LCX to the most departing position LOX; the video camera 11 is disposed in such a way that when the outer side face AX (strictly speaking, the outer end C (the corner between AD and AX) of the downstream side face AD) is situated at least at outer side of the most approaching position LCX, the line connecting the viewpoint PV with the outer end C is slanted by “α” from the outer side face X. In addition, the video camera 11 is disposed in such a way that at a certain point within the driving range of the outer end C, the angle between the line connecting the viewpoint PV with the outer end C and the downstream side face AD becomes a right angle. The left-hand camera 11b is disposed in the same manner as the camera 11a. The video camera 11 provided as described above takes an image in such a way as to include the outer end C, within the downstream side face AD of the leaf plate 8L, that is situated at the opposite side of the inner side face AI that forms the opening shape PS. As illustrated in FIG. 1, the image taken by the right-hand video camera 11a is processed by the right-hand image processing unit 13a and becomes a right-hand collimator shape image Da. Also in the left-hand side of the beam axis XB, the image taken by the left-hand video camera 11b is processed by the left-hand image processing unit 13b and becomes a left-hand collimator shape image Db. In this situation, when the distance between the inner side face AI and the outer side face AX (outer end C) is offset, there is obtained the position of the inner side face AI; when the space (the portion of the downstream side face AD) and the shielding material (the other portion) are reversed on the outer end C as the boundary, there is obtained the opening shape PS (the position of the inner side face AI of each leaf plate 8L) of the multileaf collimator 8. As described above, the video camera 11 takes an image of the outer end C, within the downstream side face AD of the leaf plate 8L, that is situated at the opposite side of the opening portion; therefore, it is made possible that without interrupting the irradiation field RB of the charged particle beam B, the visual field FV of the video camera 11 can perpendicularly face (β becomes a right angle) the downstream side face AD of the multileaf collimator 8. Accordingly, in the case where, for example, as illustrated in FIG. 3(a), while it is prioritized that β becomes a right angle (90), adjustment is performed in such a way that the foot PF of a perpendicular line PP from the viewpoint PV to the downstream side face AD is located in the vicinity (in the range closer to the middle LMX than to LCX or LOX) of a position LMX of the outer side face AX (strictly speaking, the outer end C of the downstream side face AD) when the leaf plate 8L is moved to the middle of the driving range thereof, the angle β swings within an angle that is the same as or smaller than the half of the visual field FV, with respect to 90°; thus, regardless of the driving position of the leaf plate 8L, a uniform contrast can be obtained, whereby the visibility is enhanced. In addition, in the case where the outer end C is located at a position with which the angle β largely deviate from a right angle, the image processing unit 13 corrects distortion in an image, so that there can be created a multileaf collimator shape image D that is equivalent to that at a time when the opening shape PS is directly viewed along the beam axis XB. In the case where, for example, as illustrated in FIG. 3(b), while it is prioritized that a range in which the slant α between a line from the viewpoint PV to the outer end C and the outer side face AX is positive is made wide, adjustment is performed in such a way that the foot PF of the perpendicular line PP from the viewpoint PV to the downstream side face AD is located in the vicinity (in the range closer to LCX than to the middle LMX) of the position LOX of the outer end C when the leaf plate 8L is moved to the most approaching position, a taken image does not include the outer side face AX or even when the taken image includes the outer side face AX, the angle is small; thus, the distinction between the downstream side face AD and the outer side face AX can readily be performed, whereby the position of the outer end (edge) C can accurately be detected. In the case where the viewpoint PV is situated at a position of inner side (the beam axis XB side) of LCX, there exists no portion that perpendicularly faces the downstream side face AD; however, no taken image includes the outer side face AX. In each of FIGS. 3(a) and 3(b), there are extracted and illustrated only the viewpoint PV of the video camera 11 and the portion, of the leaf plate 8L, that is in the vicinity of the outer side face AX. In FIG. 1, there is illustrated an example in which the display devices 14a and 14b for the right and left shape images Da and Db are separately arranged; however, the present invention is not limited thereto. For example, it may be allowed that image data pieces of the video cameras 11a and 11b are collectively image-processed, converted into the form of the opening shape PS, and displayed by a single display device. Moreover, it may also be allowed that not only the shape image D is simply displayed, but also, for example, a preliminarily stored pattern and the shape image are compared with each other so that the image data is converted into numerical data such as a driving distance or a driving angle that indicates the position of the leaf. In this situation, for example, there may be provided a determination device that performs comparison between data of such a built-in position detection device as explained in “Background Art” herein and data obtained from an image and then determines whether or not there exists an abnormality in the position of the leaf, the position detection device, or the driving mechanism, based on the level of coincidence between the data pieces (for example, based on whether the difference is smaller or not smaller than a threshold value). With regard to the relationship between the viewpoint PV and the leaf plate 8L, illustrated in FIG. 3, even in the case where, in addition to Embodiment 1, it is assumed that the orbit is a circumference orbit and the multileaf collimator has a shape such as the outer portion of a sector obtained by simply dividing a ring in the radial direction, there is assumed a basic form in which neighboring side faces (or the tangential lines thereof) cross each other at an angle of 90° and there exists a restriction of α+β=90. Accordingly, in the case where a position where β becomes a right angle is prioritized and, as illustrated in FIG. 3(a), there is performed setting in which the foot PF of the perpendicular line PP drawn from the viewpoint PV is situated in the vicinity of the middle LMX in the driving range of the outer end C, the range where α is positive becomes as narrow as less than the half of the driving range. However, in the case where, as leaf plates 8Lv1 through 8Lv3 (collectively referred to as 8Lv) in a variant example illustrated in FIGS. 4(a) through 4(c), the angle of the outer side face AX is changed from a basic form to an acute angle from the downstream side face, the range where a is kept to be a positive value can be expanded, even when the position at which β becomes a right angle is set to be in the vicinity of the middle LMX of the driving range of the outer side face AX (strictly speaking, the outer end C of the downstream side face AD) of the leaf plate 8Lv. As is the case in FIG. 3, in each of FIGS. 4(a) through 4(c), there are extracted and illustrated only the viewpoint PV of the video camera 11 and the portion, of the leaf plate 8Lv, that is in the vicinity of the outer side face AX. In this situation, in the case where, for example, as the leaf plate 8Lv1 illustrated in FIG. 4(a), the overall outer side face AX has a constant gradient with respect to the downstream side face AD, α is positive in the whole driving range of the leaf plate. Therefore, because the video camera 11 does not take any image of the outer side face AX, the outer end C can securely be recognized. The dotted line indicates the side faces of the rectangular leaf plate 8L. Alternatively, in the case where as the leaf plate 8Lv2 illustrated in FIG. 4(b), only the portion (the portion close to the downstream side face AD), of the outer side face AX, that is in the vicinity of the outer end C is made to have an acute angle, the video camera 11 does not take an image of the portion. In this case, the video camera 11 may take an image of the portion, of the outer side face AX, that is apart from the outer end C; however, because compared to the side face AD, the portion is away from the viewpoint PV, the outer end C can readily be recognized. Moreover, in the case where as the leaf plate 8Lv3 illustrated in FIG. 4(c), the portion (the portion close to the downstream side face AD), of the outer side face AX, that is in the vicinity of the outer end C is curved surface, the tangential line of the portion that makes contact with the outer end C may have an acute angle with respect to the downstream side face AD. As described above, the video camera 11 for monitoring the opening shape PS is disposed at a position that is at outer side of the irradiation field RB; thus, there occurs no case where the charged particle beam B collides with the image-capturing unit and hence the dispersion or the energy reduction is caused. Furthermore, the foot PF of the perpendicular line PP from the viewpoint PV of the video camera 11 to the downstream side face AD is situated at a position that is at inner side (the irradiation field RB side) of the position LOX of the outer end C of the downstream side face AD when the leaf plate 8L is driven to the most departing position; therefore, the image capturing device is suppressed from projecting from the main body of the multileaf collimator 8, whereby the irradiation system can be made compact. Accordingly, even when the image capturing device is disposed in the rotating gantry, the size of the moving body is suppressed, whereby rotating movement can readily be performed. As a result, it is made possible that even when irradiation is being implemented, the opening shape PS of the multileaf collimator 8 is confirmed and monitored, without interrupting irradiation of a charged particle beam; for example, by redundantly performing monitoring along with a built-in position detection device, the possibility of erroneous irradiation is reduced, whereby there can be configured a particle beam therapy system that can perform high-accuracy particle beam therapy the above example, as the image-capturing unit, a video camera is utilized; however, the monitoring can also be performed by a still camera. The perpendicular line PP and the foot PF thereof, the angles α and β, and the like in the foregoing explanation or the explanation, described later denote positions and angles when viewed from the thickness direction (y) of the leaf plate 8L, i.e., positions and angles in the xz plane; positions and angles in the lamination direction (y) in the leaf row 8C are not specified. As described above, the particle beam therapy system 2 according to Embodiment 1 is provided with the wobbler electromagnet 4 and the scatterer 5 that function as an irradiation nozzle for scanning the charged particle beam B supplied from an accelerator and irradiating the charged particle beam B in such a way as to enlarge the irradiation field RB; a multileaf collimator 8 in which a pair of leaf rows 8C, each of them is composed of a plurality of leaf plates 8L laminated in the thickness, is arranged in such a way as to interpose a beam axis XB of the charged particle beam B, in which respective side faces AI, of the plurality of leaf plates 8L, that face the beam axis XB are driven in approaching or in departing direction with respect to the beam axis XB so that the predetermined opening shape PS is formed, and that forms the particle beam B emitted from an irradiation nozzle in such a way that it conforms to an irradiation subject and then emits it; and video camera 11 that is an image-capturing unit that takes an image of each of the outer ends C, of respective downstream side faces AD of the plurality of leaf plates 8L in irradiation direction of the particle beam B, that are distal with respect to the beam axis XB, in order to monitor the opening shape PS. The image-capturing unit 11 is provided for each of the pair of leaf rows 8C, in such a way as to be situated at a position that is at outer side of the irradiation field RB of the particle beam B that has passed through the multileaf collimator 8, and at downstream of the multileaf collimator 8; furthermore, the image-capturing unit 11 is adjusted in such a way that foot PF of a perpendicular PP from a viewpoint PV to the downstream side face AD, is situated at a position that is at inner side of the position LOX of the outer end C when the leaf plate 8L is maximally driven in the departing direction. As a result, dispersion of a charged particle beam, reduction of the energy thereof, and upsizing of the system is suppressed from being caused and the outer end C of the leaf plate 8L is securely recognized, so that the opening shape PS of the multileaf collimator 8 can accurately be monitored. Thus, it is made possible that even when irradiation is being implemented, the opening shape PS of the multileaf collimator is confirmed and monitored, without interrupting irradiation of the charged particle beam B; for example, by redundantly performing monitoring along with a built-in position detection device, the possibility of erroneous irradiation is reduced, whereby there can be configured a particle beam therapy system that can perform high-accuracy particle beam therapy. Moreover, the shape monitoring device 10 is configured in such a way as to include the image processing unit 13 that determines the position of the outer end C, based on an image taken by the image-capturing unit 11, and performs conversion processing in which the position of the outer end C is converted into data that indicates the opening shape PS; therefore, the opening shape PS of the multileaf collimator 8 can quantitatively be evaluated, and by making comparison with data from another monitoring device, the opening shape can further securely be monitored. In this situation, in the case where foot PF of a perpendicular PP from the viewpoint PV to the downstream side face AD is located at a position that is closer to a position LMX of the outer end C at a time when the leaf plate 8L is moved to a middle of driving range than to a position LOX of the outer end C at a time when the leaf plate 8L is maximally moved in the approaching direction, the angle β swings within an angle that is the same as or smaller than the half of the visual field FV, with respect to 90°; thus, there is expanded the range where the image of the outer end C is approximately perpendicularly taken, and hence regardless of the driving position of the leaf plate 8L, a uniform contrast can be obtained, whereby the visibility is enhanced. Furthermore, in the case where at least the portion, of an outer side face AX, that is adjacent to the downstream side face AD, has an acute angle with respect to the downstream side face AD, α can be positive in the whole driving range of the leaf plate, even when the foot PF of perpendicular PP is situated in the vicinity of the middle LMX and the angle β is prioritized. Therefore, because the video camera 11 does not take any image of the outer side face AX, the outer end C can further securely be recognized based on an image including less distortion and having a high contrast. Alternatively, in the case where foot PF of perpendicular PP from the viewpoint PV to the downstream side face AD is located at a position that is closer to a position LCX of the outer end C at a time when the leaf plate 8L is maximally moved in the approaching direction than to a position of the outer end C at a time when the leaf plate 8L is moved to a middle of driving range, there is expanded the range where the gradient α of the line from the viewpoint PV to the outer end C with respect to the outer side face AX, becomes positive; thus, the taken image does not include the outer side face AX, or even when the outer side face AX is included, the angle becomes small. As a result, the distinction between the downstream side face AD and the outer side face AX can readily be performed, whereby the position of the outer end C can accurately be detected. In Embodiment 1, there has been described a case where an image is obtained by directly viewing the edge portion C of the downstream side face AD of the leaf plate through the video camera 11; however, in Embodiment 2, there is further provided a mirror; through the mirror, there is taken an image of the edge portion C of the downstream side face AD of the leaf plate 8L. FIGS. 5 and 6 are diagrams for explaining the configuration of a particle beam therapy system according to Embodiment 2 of the present invention; FIG. 5 is a diagram illustrating the configuration of the whole irradiation system in a particle beam therapy system; FIG. 6 is a diagram illustrating the positional relationship between a multileaf collimator provided in a particle beam therapy system and a shape monitoring device of the multileaf collimator. FIG. 7 is a set of diagrams for explaining the positional relationship between a leaf plate of a multileaf collimator and a shape monitoring device in a variant example of particle beam therapy system according to Embodiment 2 of the present invention. As illustrated in FIG. 5, the right visual field shape monitoring mirror 12a and the left visual field shape monitoring mirror 12b (collectively, referred to as a mirror 12) are arranged at respective positions that are at the downstream side of the multileaf collimator 8 in the irradiation direction of the charged particle beam B and are right-outer side and left-outer side with respect to the irradiation field RB; the image of the outer end C (edge portion) of the downstream side face AD of each leaf plate 8L is reflected by the right visual field shape monitoring mirror 12a and the left visual field shape monitoring mirror 12b, and the right visual field video camera 11a and the left visual field video camera 11b each take a picture of that image from the downstream side of the multileaf collimator. In this situation, in the case where as illustrated in FIG. 6, instead of the viewpoint PV, the virtual image Q, of the viewpoint PV, that is formed by the mirror 12 is disposed at the position described in Embodiment 1, the same effect as that in Embodiment 1 can be demonstrated. Moreover, the flexibility of arranging the image-capturing unit is raised and the distance between the viewpoint PV and the end C, which is an imaging subject, can be made long; thus, the angle of the visual field FV can be reduced and hence the angle for performing image-capturing in a wide region out of the driving range becomes small, whereby the accuracy of determining the position is raised. In each of FIGS. 5 and 6, there has been illustrated an example where the video camera 11 is disposed at a position that is inner side of the mirror 12; however, the present invention is not limited thereto. For example, as illustrated in FIG. 7, the video camera 11 may be disposed at a position that is outer side of the mirror 12. When the video camera 11 is disposed at a position that is outer side of the mirror 12, the positional adjustment of the visual point (the virtual image Q thereof) can readily be performed with the image-capturing unit (the camera 11 and the mirror 12) unincluded in the irradiation field RB, even in the case where the distance between the irradiation field RB and the position LOX of the outer end C at a time when the leaf plate 8L approach each other most closely is small. Still moreover, it is not necessarily required that only a single mirror 12 is utilized; it may be allowed that a plurality of mirrors fold back the optical path. As described above, in the particle beam therapy system according to Embodiment 2, the image-capturing unit includes a mirror 12 that reflects the image of the outer end C of the downstream side face AD of the leaf plate 8L and a video camera 11 that is an image-capturing device disposed at a position that is away from the mirror 12, and the arrangement of the mirror 12 and the image-capturing device 11 is adjusted, based on a virtual image Q, of the viewpoint PV of the image-capturing device 11, that is caused by the reflection of the mirror 12; therefore, the flexibility of arranging the image-capturing unit is raised and the distance between the viewpoint PV and the outer end C, which is an imaging subject, can be made long; thus, the angle of the visual field FV can be reduced and hence the angle for performing image-capturing in a wide region out of the driving range of the leaf plate 8L becomes small, whereby the accuracy of determining the position is raised. 2: particle beam therapy system 3: dose meter 4: wobbler electromagnet (irradiation nozzle) 5: scatterer (irradiation nozzle) 6: ridge filter 7: range shifter 8: multileaf collimator 8C: leaf row 8C: leaf plate 9: bolus 10: shape monitoring device 11: camera (image-capturing unit) 12: mirror (image-capturing unit) 13: image processing unit 14: image display device AI: inner side face of leaf plate AD: downstream side face of leaf plate AU: upstream side face of leaf plate AX: outer side face of leaf plate B: charged particle beam (particle beam) C: outer end (edge portion) of downstream side face of leaf plate D: processing image FV: visual field of image-capturing unit LCX: position of outer end at the timing of closest approach LMX: position of outer end at the timing of being at middle of driving range LOI: position of inner side face at the timing of being apart most LOX: position of outer end at the timing of being apart most PF: foot of perpendicular line between viewpoint and downstream side face PP: perpendicular line between viewpoint and downstream side face PS: opening shape PV: viewpoint of image-capturing unit Q: optical viewpoint (virtual image of viewpoint) of image-capturing unit RB: irradiation field α: angle between outer side face and line from viewpoint to outer end of downstream side face β: angle between downstream side face and line from viewpoint to outer end of downstream side face
048062782	claims	1. A method of segregating radioactive iodine isotopes present in fluid samples, especially in those taken from the primary coolant circuit of a pressurized water nuclear reactor comprising the steps of: driving bubbles of a suitably chosen inert gas through a sample taken in a continuous process from a fluid medium to be investigated, expelling thereby the dissolved gases and among them radioactive noble gases from said sample, adding, if necessary, a reagent to said sample for adjusting its pH-value to neutral or an alkaline level, transporting said sample to a column filled with a high active surface material in order to bind quantitatively the cations, fluoride anions and colloidal-state corrosion products from said sample, letting said sample leaving said column to clear and permitting thereby the decay of short and very short half-period radioisotopes which may have remained in said sample during the previous steps, making the effluent leaving said column--containing between radioactive isotopes in considerable percentage iodine isotopes only--pass in continuous flow through an energy-selective gamma detector and continuously plotting the signals generated in said detector by the radioactive iodine isotopes. 2. A method according to claim 1, wherein for quantitative binding of the undesired components said sample is transported through said column from about 5 minutes to about 15 minutes. 3. A method according to claim 1, wherein said sample is cleared over about 20 minutes. 4. A method according to claim 1, wherein the signals generated by the radioactive iodine isotopes are plotted in integrated form by calibrated counters operated in gated mode and adjusted according to the different gamma-energy levels assigned to the various radioactive iodine isotopes. 5. A method according to claim 1, wherein the signals generated by the radioactive iodine isotopes are plotted selectively by means of at least one ratemeter adjusted according to the different gamma-energy levels assigned to the various radioactive iodine isotopes. 6. An apparatus for segregating radioactive iodine isotopes present in fluid sample, especially for use with samples taken from the primary coolant circuit of a pressurized water nuclear reactor, comprising a sampling unit for continuously taking and providing samples of suitable pressure and temperature, a degassing vessel connected through a piping with said sampling unit and equipped with a gas inlet for transporting appropriate neutral gas for agitating said sample, for expelling therefrom the undesired gaseous contaminations and for permitting level control of said sample, a transfer pump for forwarding said sample to a bubble removing cell, a segregation column filled with high specific surface active material for binding quantitatively the cations, fluoride anions and colloidal-state corrosion products present in said sample, a piping connected to an output of said segregation column to contain said sample during a predetermined time for letting the short and very short half-period radio-isotopes to decay before detecting said sample, a continuously operating flow-type energy-selective gamma-radiation detecting system for investigating said clear sample and a signal processing and recording unit for receiving the signals emitted by said detecting system. 7. An apparatus according to claim 6, further comprising a cock arranged between said transfer pump and said segregation column and connected to a pump for forwarding appropriate reagent, e.g. for adjusting the pH-value of said sample to neutral or an alkaline level, from a container to said cock. 8. An apparatus according to claim 6, wherein said segregation column is filled with amorphous zirconium phosphate. 9. An apparatus according to claim 6, comprising in said detecting system scintillation detectors connected to said signal processing and recording unit being capable of differentiating at least three specific gamma-energy levels of radioactive iodine isotopes.
	summary
	claims	1. A gamma radiation detecting apparatus comprising:a gamma radiation collimator which receives and collimates gamma radiation;a scintillation crystal which receives the gamma radiation from the gamma radiation collimator and converts the gamma radiation into visible light;a light-collecting mirror which collects and focuses the visible light received from the scintillation crystal;a beam splitter which splits and directs the visible light from the light-collecting mirror into at least two beams;an optical lens which focuses a first beam of the visible light from the beam splitter;a charge coupled device which receives the visible light from the optical lens and converts the visible light into an electrical charge;shielding which prevents the charge coupled device and the scintillation crystal from being exposed to unwanted gamma radiation; andat least one light detector which receives a second beam of the visible light from the beam splitter and measures total energy of the second beam;wherein the charge coupled device is positioned so that it is not directly exposed to gamma radiation; andwherein the light-collecting mirror splits the visible light from the gamma radiation flux received through the scintillation crystal, preventing exposure of the charge coupled device to gamma radiation. 2. The gamma radiation detecting apparatus of claim 1, wherein the charge coupled device is an electron multiplying charge coupled device. 3. The gamma radiation detecting apparatus of claim 1, wherein the charge coupled device is an intensified charge coupled device. 4. The gamma radiation detecting apparatus of claim 1, wherein the gamma radiation collimator is a parallel-hole collimator. 5. The gamma radiation detecting apparatus of claim 1, wherein the gamma radiation collimator is a coded aperture collimator. 6. The gamma radiation detecting apparatus of claim 1,wherein the beam splitter comprises a second mirror that is partially silvered to split and direct the visible light. 7. The gamma radiation detecting apparatus of claim 1 further comprising:a first electronic device which reads and processes the output of the charge coupled device to create digital data of an image, wherein the first electronic device analyzes the image and determines whether the gamma radiation was scattered. 8. The gamma radiation detecting apparatus of claim 7, wherein the first electronic device determines the position of the gamma radiation interaction within the scintillation crystal by analyzing the pixels of the charge coupled device. 9. The gamma radiation detecting apparatus of claim 7 further comprising:a second electronic device connected to the first electronic device and running software to process the digital data and display the image, wherein the software extracts the position of a single gamma ray interaction from the image. 10. The gamma radiation detecting apparatus of claim 1, wherein the optical lens comprises a zoom lens which can vary the spatial resolution anywhere in between and including a high spatial resolution for imaging a portion of the scintillation crystal and a low resolution for imaging substantially the entire scintillation crystal. 11. The gamma radiation detecting apparatus of claim 1 further comprising:a cooling device which cools the charge coupled device. 12. The gamma radiation detecting apparatus of claim 1,wherein the at least one light detector measures energy deposited by gamma rays during one gamma event by integrating the amount of visible light emitted by the scintillation crystal during one gamma event; andwherein the at least one light detector triggers acquisition of an image by the charge coupled device on detection of a gamma interaction. 13. The gamma radiation detecting apparatus of claim 12, wherein the at least one light detector comprises a photomultiplier tube. 14. The gamma radiation detecting apparatus of claim 1, wherein the charge coupled device comprises a plurality of charge coupled devices. 15. A method comprising the steps of:(a) collimating gamma radiation with a gamma radiation collimator;(b) positioning a charge coupled device so that it is not directly exposed to the gamma radiation;(c) shielding the charge coupled device and a scintillation crystal from being exposed to unwanted gamma radiation;(d) converting the collimated gamma radiation into visible light using the scintillation crystal;(e) collecting and focusing the visible light using a light-collecting mirror such that the visible light is split from the gamma radiation flux received through the scintillation crystal;(f) splitting and directing the visible light into at least two beams using a beam splitter;(g) focusing a first beam of the visible light on the charge coupled device using an optical lens;(h) directing a second beam of the visible light onto at least one light detector;(i) measuring total energy of the second beam using the at least one light detector;(j)aconverting the first beam of the visible light into an electrical signal using the charge coupled device; and(k) processing the electrical signal from the charge coupled device into an image using an electronic device electrically coupled to the charge coupled device. 16. The method of claim 15 further comprising the steps of:determining the position of the gamma radiation interaction within the scintillation crystal by analyzing the pixels of the image produced by the charge coupled device. 17. The method of claim 16, wherein step (i) further comprises:integrating the amount of visible light emitted by the scintillation crystal during one gamma event, and eliminating scattered gammas from the image of step (k). 18. The method of claim 17, wherein step (k) is triggered by the at least one light detector detecting a gamma interaction. 19. A gamma radiation detecting apparatus comprising:a gamma radiation collimator which receives and collimates gamma radiation;a scintillation crystal which receives the gamma radiation from the gamma radiation collimator and converts the gamma radiation into visible light;a light-collecting mirror which collects and focuses the visible light received from the scintillation crystal;an optical lens which focuses the visible light from the light-collecting mirror;an electron multiplying charge coupled device which receives the visible light from the optical lens and converts the visible light into an electrical charge;a gamma-transparent apparatus, located between the gamma radiation collimator and the scintillation crystal, which collects visible light from at least one side of the scintillation crystal;at least one light detector which receives the visible light from the gamma-transparent apparatus and measures total energy of the received light;shielding which prevents the electron multiplying charge coupled device and the scintillation crystal from being exposed to unwanted gamma radiation; anda first electronic device which reads and processes the output of the electron multiplying charge coupled device to create digital data of an image, wherein the first electronic device analyzes the image and determines whether the gamma radiation was scattered; and wherein the first electronic device determines the position of the gamma radiation interaction within the scintillation crystal by analyzing the pixels of the charge coupled device;wherein the electron multiplying charge coupled device is positioned so that it is not directly exposed to gamma radiation;wherein the light-collecting mirror splits the visible light from the gamma radiation flux received through the scintillation crystal, preventing exposure of the electron multiplying charge coupled device to gamma radiation;wherein the at least one light detector measures energy deposited by gamma rays during one gamma event by integrating the amount of visible light received from the gamma-transparent apparatus during one gamma event; andwherein the at least one light detector triggers acquisition of the image on detection of a gamma interaction. 20. The gamma radiation detecting apparatus of claim 19, further comprising:a cooling device which cools the electron multiplying charge coupled device; anda second electronic device connected to the first electronic device and running software to process the digital data and display the image, wherein the software extracts the position of a single gamma ray interaction from the image.
	summary
	claims	1. An ion implantation system having an ion beam mass analyzer beam guide comprising: a waveguide associated with the ion beam mass analyzer beam guide; wherein the waveguide is adapted to couple RF or microwave power to the ion beam mass analyzer beam guide; and wherein the waveguide comprises a plurality of unobstructed openings for coupling the RF or microwave power to the ion beam mass analyzer beam guide. 2. The system of claim 1 , wherein the waveguide is configured to facilitate a standing wave resonance condition therein, and the plurality of unobstructed openings are positioned along the waveguide to enable an efficient coupling of the RF or microwave power to the ion beam mass analyzer beam guide. claim 1 3. The system of claim 2 , wherein the plurality of unobstructed openings are spaced from each other be a distance of nxcex/2, and wherein n is an integer. claim 2 4. A waveguide for coupling microwave energy from a power source with a beam plasma in a passageway of an ion beam mass analyzer beam guide, comprising: a first layer adapted to propagate microwave energy from the power source with a top side and a bottom side, the first layer extending longitudinally along an arcuate path in a first plane from an entrance end to an exit end and laterally between an inner radial side and an outer radial side; and second and third layers mounted respectively on the top and bottom sides of the first layer and extending longitudinally in second and third planes respectively along the path from the entrance end to the exit end and laterally between the inner and outer radial sides, wherein the second and third planes are generally parallel with the first plane, wherein at least one of the second and third layers faces the interior of the passageway and comprises a plurality of laterally extending longitudinally spaced unobstructed slots therethrough between the passageway and the first layers and wherein the other of the second and third layers faces away from the interior of the passageway. 5. The waveguide of claim 4 , further comprising a plurality of laterally extending longitudinally spaced magnets adapted to provide a multi-cusped magnetic field in the beam guide passageway, whereby the multi-cusped magnetic field and the microwave energy from the power source cooperatively interact to create an electron cyclotron resonance condition along at least a portion of the passageway. claim 4 6. The waveguide of claim 5 , further comprising a beam guide cover located between the plurality of magnets and the interior of the passageway and adapted to cover the plurality of magnets and to expose the plurality of unobstructed slots. claim 5 7. The waveguide of claim 6 , further comprising a plurality of O-rings individually encircling the laterally extending longitudinally spaced unobstructed slots. claim 6 8. The waveguide of claim 7 , wherein the first layer is made from quartz, the second and third layers are made from aluminum, the O-rings are made from elastomer material, and the beam guide cover is made from aluminum. claim 7 9. The waveguide of claim 8 , wherein the plurality of laterally extending longitudinally spaced unobstructed slots through the at least one of the second and third layers are located at an angular pitch of about 5.326 degrees with a lateral slot length of about 50 mm and a longitudinal slot width of about 5 mm, wherein the waveguide extends through an arcuate angle of about 135 degrees, and wherein the inner and outer radial sides have a radiuses of about 370 mm and 430 mm, respectively. claim 8 10. The waveguide of claim 5 , wherein the plurality of laterally extending longitudinally spaced unobstructed slots through the at least one of the second and third layers are located at an angular pitch of about 5.326 degrees with a lateral slot length of about 50 mm and a longitudinal slot width of about 5 mm, wherein the waveguide extends through an arcuate angle of about 135 degrees, and wherein the inner and outer radial sides have a radiuses of about 370 mm and 430 mm, respectively. claim 5 11. The waveguide of claim 4 , wherein the plurality of laterally extending longitudinally spaced unobstructed slots through the at least one of the second and third layers are located at an angular pitch of about 5.326 degrees with a lateral slot length of about 50 mm and a longitudinal slot width of about 5 mm, wherein the waveguide extends through an arcuate angle of about 135 degrees, and wherein the inner and outer radial sides have a radiuses of about 370 mm and 430 mm, respectively. claim 4 12. The waveguide of claim 11 , wherein the other of the second and third layers is mounted on a dipole magnet in the beam guide passageway. claim 11 13. The waveguide of claim 4 , wherein the other of the second and third layers is mounted on a dipole magnet in the beam guide passageway. claim 4 14. The waveguide of claim 4 , wherein the first layer is made from quartz and the second and third layers are made from aluminum. claim 4 15. A mass analyzer beam guide apparatus for conditioning an ion beam along a path in an ion implantation system, comprising: a mass analysis magnet mounted along a passageway along the path; a power source adapted to provide an electric field in the passageway; a waveguide to couple the electric field with a beam plasma associated with the ion beam; and a magnetic device adapted to provide a multi-cusped magnetic field in the passageway; wherein the power source, the waveguide, and the magnetic device are cooperatively adapted to provide containment of the ion beam in at least a portion of the passageway; wherein the beam guide further comprises a top, a bottom, and laterally opposite first and second sides defining the passageway, wherein the top, bottom, first and second sides extend longitudinally along the path between an entrance end and an exit end; wherein the waveguide further comprises: a first layer adapted to propagate microwave energy from the power source with a top side and a bottom side, the first layer extending longitudinally along an arcuate path in a first plane from an entrance end to an exit end and laterally between an inner radial side and an outer radial side; and second and third layers mounted respectively on the top and bottom sides of the first layer and extending longitudinally in second and third planes respectively along the path from the entrance end to the exit end and laterally between the inner and outer radial sides, wherein the second and third planes are generally parallel with the first plane, and wherein at least one of the second and third layers faces the interior of the passageway and comprises a plurality of laterally extending longitudinally spaced unobstructed slots therethrough between the passageway and the first layer and the other of the second and third layers faces away from the interior of the passageway; wherein the magnetic device comprises a plurality of laterally extending longitudinally spaced magnets adapted to provide a multi-cusped magnetic field in the beam guide passageway; and wherein the multi-cusped magnetic field and the electric field from the power source cooperatively interact to create an electron cyclotron resonance condition along at least a portion of the passageway. 16. The apparatus of claim 15 , wherein the first layer is made from quartz and the second and third layers are made from aluminum. claim 15 17. The apparatus of claim 16 , wherein the plurality of laterally extending longitudinally spaced unobstructed slots through the at least one of the second and third layers are located at an angular pitch of about 5.326 degrees with a lateral slot length of about 50 mm and a longitudinal slot width of about 5 mm, wherein the waveguide extends through an arcuate angle of about 135 degrees, and wherein the inner and outer radial sides have a radiuses of about 370 mm and 430 mm, respectively. claim 16 18. A waveguide for coupling an electric field with a plasma in an ion beam mass analyzer passageway, comprising: a base layer located in a first plane adapted to propagate microwave energy from a power source, and having a top side and a bottom side extending longitudinally along an arcuate path from an entrance end to an exit end and laterally between an inner radial side and an outer radial side; a second layer mounted on one of the top and bottom sides facing the interior of the passageway and having a plurality of laterally extending longitudinally spaced unobstructed slots therethrough between the interior of the passageway and the base layer; whereby microwave energy from the power source propagating along the base layer is coupled with the plasma in the interior of the passageway near the laterally extending longitudinally spaced unobstructed slots. 19. The waveguide of claim 18 , further comprising means for providing the microwave energy from the power source to the waveguide. claim 18 20. The waveguide of claim 18 , wherein the base layer is made from quartz, the second layer is made from aluminum. claim 18 21. The waveguide of claim 18 , wherein the plurality of laterally extending longitudinally spaced unobstructed slots through the second layer is located at an angular pitch of about 5.326 degrees with a lateral slot length of about 50 mm and a longitudinal slot width of about 5 mm, wherein the waveguide extends through an arcuate angle of about 135 degrees, and wherein the inner and outer radial sides have a radiuses of about 370 mm and 430 mm, respectively. claim 18 22. A method of delivering RF or microwave power to an ion beam mass analyzer beam guide, comprising: providing a waveguide in the beam guide comprising a plurality of unobstructed slots; coupling an RF or microwave power source to the waveguide; and delivering the RF or microwave power to one or more regions proximate at least one of the unobstructed slots in the ion beam mass analyzer beam guide via the waveguide. 23. The method of claim 22 , wherein delivering the RF or microwave power to one or more regions in the ion beam mass analyzer beam guide via the waveguide comprises generating a standing wave in the waveguide. claim 22 24. The method of claim 23 , wherein delivering the RF or microwave power to one or more regions in the ion beam mass analyzer beam guide via the waveguide comprises coupling the RF or microwave power at regions spaced from each other in the beam guide by a distance of nxcex/2, and wherein n is an integer. claim 23 25. The method of claim 22 , wherein delivering the RF or microwave power to one or more regions in the ion beam mass analyzer beam guide via the waveguide comprises coupling the RF or microwave power at regions spaced from each other in the beam guide by a distance of nxcex/2, and wherein n is an integer. claim 22
	description	This application is a Continuation-In-Part of applicant's Parent application Ser. No. 14/243,302, Filed 2014 Apr. 2, now U.S. Pat. No. 9,852,819, granted 2017 Dec. 26. This Parent Application claims priority of applicant's provisional patent application, Ser. No. 61/930,369, filed 2014 Jan. 22. The following abbreviations are used in this patent and are defined in situ at the first use but are also all defined here for ready reference. CAES Compressed Air Energy StorageCG Compressed GasCGES Compressed Gas over Water Emergency Reactor Cooling SystemCGECCS compressed gas emergency core cooling systemCO2 Carbon DioxideDPCS Prior-Art Depressurization and Cooling SystemGJ GigaJoulesHOE Heat Of Expansion For Compressed AirMJ MegaJoulesMW MegaWattsMWH MW HoursNCAP Nuclear Compressed Air Power (combined thermal power plantand a CAES facility)PCV Primary Containment VesselPSI Pounds Per Square InchRES Renewable Energy StorageRPV Reactor Pressure VesselRRTC Railroad Tank CarSCRAM An emergency shutdown of an operating nuclear power plant(“Safety Control Rod Axe Man”)SFP Spent Fuel PoolTPP Thermal Power PlantWHF Waste heat fluid from a thermal power plant Climate warming of the planet has become a serious problem in recent years. Such warming has caused glaciers to melt, raising the ocean levels and flooding coastal areas and ocean islands. The warming has been attributed, inter alia, to the increase of carbon dioxide (CO2) levels in the atmosphere. Thus moderating climate warming will require the replacement of many fossil fuel power plants with renewable energy sources that emit no CO2. However, most of the world's available wind, solar, and tidal energy cannot be utilized because there is no way to store most of the available renewable energy until it is needed on a power grid. Hence, an important physical infrastructure needed to moderate climate warming is an inexpensive, utility-scale, renewable energy storage (RES) design that can be installed on electrical grids all over the world. A desirable “utility-scale” design would be an affordable facility which can store at least 2000 megawatt-hours (MWH) of energy and can generate at least 200 MW of power on short notice. The largest utility-scale compressed air energy storage (CAES) facilities in existence are compressed air energy storage (CAES) facilities in McIntosh, Ala., and Huntorf, Germany. These CAES systems have been in operation for decades. Although desirable for storing energy, these old-style CAES facilities have not been copied around the world because it is expensive to recover and utilize their stored energy. Specifically when the stored compressed air is to be used, it must be expanded from a highly compressed level, e.g., 1000 psi or 70 bar (1 bar=atmospheric pressure or 14.7 psi) to ambient pressure. But when isothermal gas is partially expanded in an air turbine or motor its temperature drops substantially such that it will freeze equipment in contact. Complete expansion requires the addition of heat energy to the compressed air. This is called the “heat of expansion” (HOE). In the present old-style CAES facilities, large amounts of natural gas are burned to provide the HOE so that the stored compressed air can expand to a lower pressure and drive a turbine or air motor to produce electrical power. The amount of HOE needed by a CAES system can be as much as 30% of the energy output of the CAES facility. In addition to the expense of the natural gas, these old style CAES systems produce substantial CO2 emissions when they burn the natural gas or any other fossil fuel. An energy storage facility that puts large amounts of CO2 in the atmosphere defeats the purpose of generating and capturing renewable wind and solar energy. The emergency reactor core cooling system (ECCS) system and method described overcomes one or more of the deficiencies of prior-art systems while satisfying many requirements to prevent more nuclear accidents under the worst conditions such as happened at Fukushima in 2011. All embodiments are designed to absorb the decay heat from a large reactor after an emergency shutdown (called “SCRAM” in the industry for “Safety Control Rod Axe Man”) when all internal ECCS systems have been disabled and none of the existing and prior art ECCS systems, passive or otherwise, can function. Some embodiments assume only that the reactor pressure vessel (RPV) can still receive and hold externally injected cooling fluid long enough to cool the nuclear fuel rods therein. Another embodiment cools a reactor under a very worst condition in which the RPV is so damaged (ruptured) that it can no longer hold water around the fuel rods therein. All embodiments are “passive safety systems” in that the only energy source needed or used for the reactor cooling operation performed is the potential energy stored in the compressed gas contained in the system. Various embodiments are installed outside nuclear plant buildings. The cooling fluid used is cooling water stored separately outside a nuclear plant and-or the compressed gas itself. The various embodiments can be installed immediately using existing proven structural components and technology from the oil and gas well drilling and the railroad industries. The system can be installed underground or safely above ground, outside the structures housing a nuclear power station so that this ECCS system is far less susceptible to damage by acts of nature or terrorism. It does not depend on or need any of the existing ECCS systems or operational cooling systems inside a nuclear plant or any emergency backup power systems. This system requires only a connection to the external cooling water input port(s) that exist on all nuclear reactors as a last resort means to cool a reactor when all internal cooling functions are inadequate. One embodiment, called the Compressed Gas Emergency Cooling System (CGES) employs a gas, such as air or nitrogen, that is safely stored at high pressure in a first tank and is used to pressurize cooling water in another, larger tank. Unlike other gas-over-water pressure tank schemes, the cooling water tank in this embodiment is pressurized only when it is needed during a nuclear emergency. This allows cooling water to be retained long-term in much less expensive water storage tanks. The present system can bring a reactor to “cold condition” with the reactor coolant held below 100° C. long-term. All embodiments of the present system do not need nor use continued primary coolant circulation in a SCRAMMED reactor. Another embodiment uses a compressed air energy storage (CAES) system to provide emergency reactor cooling by direct injection of large quantities of compressed gas. It is called here a Nuclear Compressed Air Power (NCAP) plant. The NCAP embodiment cools the fuel rods in a reactor under the worst case that the RPV is so damaged (ruptured) that it cannot hold liquid coolant. This is something that cannot be done by any other existing or approved ECCS system for water cooled nuclear reactors. The NCAP also combines a CAES with a nuclear power station in a unique manner that achieves great reductions in operating cost and production of climate warming CO2 during normal power generating operations. These reductions in cost and pollution more than pay for the addition of a CAES plant to a nuclear plant. The NCAP system utilizes the abundant (free) waste heat from any standard Thermal Power Plant (TPP) to replace the costly and polluting natural gas that is used to heat the expanding compressed air in an old-style CAES facility. The NCAP embodiment is a large improvement of the CGES design concept. The NCAP is very appropriate for new nuclear power stations. This heat energy transferred to the compressed air—the heat of expansion (HOE)—is necessary to supply the heat that expanding compressed air requires to prevent freezing. The operating cost reduction and elimination of CO2 emissions justify connecting a CAES facility and an existing TPP plant to implement the NCAP system. While all prior art CAES systems use some expensive additional heat source for the HOE, the NCAP system simply utilizes this free waste heat to satisfy the heat energy requirement of a companion CAES. Potential underground air reservoir sites exist near many TPP power plants at all latitudes around the world. It has been estimated that underground CAES air reservoirs could be located under 80% of the land area of the U.S. That means there can be CAES facilities near many of the thousands of thermal power plants around the world. And, as explained herein, the CAES storage facility can be many miles distant from its companion TPP power plant in the NCAP system. In the NCAP system, a TPP power plant and a CAES are connected by a relatively inexpensive heat exchanger so that some of the free waste heat generated by the TPP plant is transferred to the compressed air coming out of the CAES. The heat exchanger required is similar in design to the heat exchangers used at all thermal power plants and in other systems cited above. In the NCAP system, the additional heat exchanger simply transfers some of the TPP waste heat to the compressed air instead of throwing all of the TPP waste heat into the environment. The TPP power plant in the NCAP system can be a fossil fuel, nuclear, biomass, geothermal, or solar-thermal power plant. The waste heat generated and released to the environment by a TPP plant is conveyed by a waste heat fluid (WHF) such as steam or hot water, hot exhaust gases from a gas turbine, or molten metal that stores energy. Forty to sixty percent of the thermal energy generated in a TPP plant is waste heat. Ninety percent of the power plants in the world are steam generating (TPP) power plants. More thermal energy is thrown into the environment than is converted into useful power. In fact, the amount of waste heat generated by thermal power plants is many times the total amount of renewable energy being captured on the electrical grids of the world today. The NCAP system described herein simply combines these two huge untapped or lost sources of energy to provide significantly more useful power on electrical grids. The CGES and NCAP systems are able to cool a large nuclear reactor for many days after shutdown when all in-plant, prior art emergency cooling systems are disabled. The CGES and NCAP systems do not require any outside emergency power for their operation. They are low in cost, easily installed, and safe. As described, they can be installed immediately using existing proven structural components and technology from the oil-gas well drilling industry and the railroad industry. Advantages Various aspects of the present systems have one or more of the following advantages: They are passive safety systems that require no outside power and hence can be more reliable. They are or can be separate from and located away from the nuclear plant structures that contain the nuclear reactor and the in-plant emergency reactor cooling systems. They do not require external or backup power sources, other than their own stored energy. They can cool a nuclear reactor for several days after shutdown (SCRAM) when all existing in-plant cooling equipment, cooling fluid, emergency power systems are lost or disabled. They provide an add-on system that does not require major interruption of the operation of a nuclear power plant. They can be constructed with readily available and inexpensive standard components and they can be installed immediately at our aging nuclear plants. Further advantages of various aspects will be apparent from the ensuing description and accompanying drawings. The NCAP system provides very large, inexpensive, pollution-free compressed air energy storage (CAES) facilities that can be installed quickly all over the world near existing power plants to store renewable energy. The NCAP system uses only existing equipment and infrastructure. It requires no new exotic technology. Where ever the underground strata near an existing TPP power plant will support a large reservoir for compressed air, the NCAP system can be installed using nothing more than oil well drilling techniques and technology. In the NCAP system, the renewable wind/solar energy being generated and the CAES facility that stores it do not have to be located very close to each other or to the TPP power plant. The compressed air stored in the CAES facility can be transported (piped) long distances to its companion TPP plant just as easily as piping natural gas to the CAES. There is no heat loss because the isothermal compressed air is at ambient temperature. (Oil well drillers in the Permian Basin in Texas transport liquid CO2 at 47.6 bar and ambient temperature over a hundred miles from New Mexico). Likewise, the renewable energy generation can be far away from either the CAES or the TPP plant. The renewable energy is normally transmitted as electrical power over the gird to drive the compressor that charges the CAES air reservoir at some other location on the grid. Whenever the CAES companion in a NCAP facility is not generating power, the TPP plant operates as if its CAES companion did not exist. There are other important benefits not provided by any other CAES design. The NCAP allows conversion of some of the TPP thermal plant waste heat into useful power, as described herein. Also, when the TPP plant is a nuclear plant, the NCAP system provides another measure of safety if the standard Emergency Core Cooling Systems fail (as happened at Fukushima). The large amount of stored compressed air in the companion CAES facility can be used to cool the fuel rods in the reactor pressure vessel. Testimony to the uniqueness of the NCAP system is the fact that some very expensive old-style CAES are being planned today for hundreds of millions of dollars because of the demand by government agencies for utilities to build renewable energy storage. The NCAP system is very different than the above systems. The main purpose of the NCAP is to eliminate the need for natural gas or any other expensive source of energy to retrieve the energy stored in an operational CAES facility. The NCAP does this by using the free waste heat generated by an existing thermal power plant (TPP). The NCAP system does not emit additional CO2 into the atmosphere in order to supply the HOE needed by the compressed air released from a CAES. In addition, the NCAP converts some of the waste heat from a thermal power plant into useful power by transferring it into the CAES compressed gas that then expands in an air turbine to generate electrical power. The NCAP system provides very large, inexpensive, pollution-free compressed air energy storage (CAES) facilities that can be installed quickly all over the world near existing power plants to store renewable energy. The NCAP system uses only existing equipment and infrastructure. It requires no new exotic technology. Where ever the underground strata near an existing TPP power plant will support a large reservoir for compressed air, the NCAP system can be installed using nothing more than oil well drilling techniques and technology. In the NCAP system, the renewable wind/solar energy being generated and the CAES facility that stores it do not have to be located very close to each other or to the TPP power plant. The compressed air stored in the CAES facility can be transported (piped) long distances to its companion TPP plant just as easily as piping natural gas to the CAES. There is no heat loss because the isothermal compressed air is at ambient temperature. (Oil well drillers in the Permian Basin in Texas transport liquid CO2 at 700 psi and ambient temperature over a hundred miles from New Mexico). Likewise, the renewable energy generation can be far away from either the CAES or the TPP plant. The renewable energy is normally transmitted as electrical power over the gird to drive the compressor that charges the CAES air reservoir at some other location on the grid. Whenever the CAES companion in a NCAP facility is not generating power, the TPP plant operates as if its CAES companion did not exist. Cooling a Nuclear Reactor A nuclear power plant generates electricity by using a nuclear reactor to boil feed water to create high-pressure steam. This steam in turn drives turbines, which in turn drive electrical generators that create the electricity. The feed water is called the “primary coolant” in the nuclear industry since it absorbs and carries away the heat generated by the nuclear fuel rods in a reactor pressure vessel (RPV) inside the reactor. After the steam drives the turbines it is condensed back to water and is recirculated back to the RPV to be heated again to continuously generate new steam for the turbines. Sometimes it is necessary to shut down a nuclear plant for operational reasons or because of an accident or terrorist attack, loss of outside electrical power, or a natural event such as an earthquake or fire. Thus some means must be provided to stop the nuclear chain reaction which is continuously occurring in the reactor. This is done by inserting control rods and-or chemical substances into the reactor to absorb the fission neutrons that drive the nuclear chain reaction and heat the primary coolant. However, stopping the nuclear reaction does not eliminate the danger or cool the reactor adequately. A major source of heat, called the “decay heat,” remains after the fission process is stopped. This decay heat is generated by radioactive decay of isotopes within the nuclear fuel rods. Even after shutdown this heat must be continually removed from the nuclear fuel rods for days or weeks in order to stabilize the reactor. Otherwise, a nuclear “meltdown” can occur, as happened in the Three Mile Island, Pa. (1979) and Fukushima Daiichi, Japan (2011) nuclear accidents. During normal power generation the primary coolant circulates through plumbing (termed the “primary boundary”) coupled to the RPV. The primary boundary includes an outlet pipe from the RPV (commonly called the “hot leg”) that carries the steam generated in the RPV the turbine(s). After condensation back to water the primary coolant exhaust from the turbine is returned to the RPV by a pipe called the “cold leg.” Existing Emergency Core Cooling Systems (ECCS) in water-cooled reactors all depend on continued circulation of the primary coolant around the fuel rods to absorb and dissipate the decay heat. The cooling fluid is circulated by large electrical or steam driven pumps that require backup power sources in case the nuclear plant is disconnected from the electrical grid that it serves. At least one newer and approved ECCS design, the AP1000 (see Non-Patent Literature below), called a “passive safety system,” does not require backup emergency power. The AP1000 utilizes additional water from a large gravity flow storage tank mounted above the reactor to cool the primary coolant circulating through the reactor. The AP1000 relies on continuing decay heat to circulate the primary coolant through a heat exchanger by convection. Hence, the AP1000 design by itself cannot bring a reactor to “cold shutdown condition” (less than 100° C.), as noted below and in the discussion of the cited patent to Sato, infra. Naval nuclear reactors, i.e., those used in submarines and ships, are of less concern because they are surrounded at all times by abundant ocean water to cool the fuel rods in a reactor in an emergency. This degree of safety has not been available for land-based nuclear plants. The Fukushima nuclear disaster presented a horrible reality to the world. It showed that all existing emergency reactor cooling equipment in the world's nuclear power plants today can be disabled by the forces of nature or by terrorists. The existing emergency cooling systems are fragile in many ways. The biggest weakness is that the essential emergency cooling apparatus in most nuclear plants is co-localized in the nuclear plant buildings. It is rather disturbing that after the Fukushima accident there were several nuclear plant shutdowns in the U.S. where some of the backup electrical generators failed and a last-ditch battery backup system had to be used. Water Flow Urged by Gravity for Emergency Core Cooling of a Nuclear Reactor An example of water delivery by gravity flow is found in the above-mentioned AP1000 ECCS nuclear power plant design by Westinghouse Corp. of Cranberry Township, Pa. This design employs a large tank of emergency cooling water that is placed atop a structure within a nuclear plant. When released in an emergency, gravity causes this secondary cooling water to flow through a heat exchanger where it extracts the decay heat from the hot primary coolant in the reactor. So long as the primary coolant circulation system is intact and operating, the secondary cooling water delivered by gravity is turned into steam. The steam is condensed by cooling by outside air flow over surfaces attached to the containment building. In this design, the gravity water circulation equipment is all inside the nuclear plant structure. Its large emergency water tank is mounted high above the reactor pressure vessel and thus can be disabled by severe earthquakes. Any assault on a nuclear plant that damages the internal plumbing in which the primary or gravity cooling water are circulating can disable the AP1000 ECCS. A dedicated terrorist attack that damages a few critical pipelines and valves inside the nuclear plant can disable all emergency cooling operations. This can lead to a disastrous meltdown of the fuel rods. Water Flow Urged by Compressed Gas Domestic water tanks that are pressurized by compressed air above the water are well known. These tanks are designed to provide water in a given pressure range without requiring a water source to be energized for each delivery of water into or out of the tank. These tanks generally employ an impermeable membrane between the compressed air and the water below in order to avoid loss of the air by absorption of the air into the water (aeration). Gas pressurized water tanks called “accumulators” are used in nuclear plants today. The water is commonly loaded with a borate solution or other “neutron poisons” that stop the fission reaction in a nuclear reactor during a nuclear emergency. The tanks maintain a high pressure nitrogen or water vapor bubble above the water. Because these accumulator tanks are placed inside the nuclear plant building and are under constant high pressure, their size is limited. They do not store sufficient cooling fluid to absorb the decay heat for many hours, let alone many days, from a nuclear reactor immediately after shutdown. Very high static gas pressures (up to 200 bar) cannot be used in the water tanks above unless the tanks are relatively small. If the tank dimensions are large or the gas pressures too high, the tank cannot withstand the high hydraulic forces on its walls of the tank without bursting. The following is a list of some possibly relevant prior art that shows prior-art emergency cooling systems for nuclear power plants. Following this list I provide a discussion of these references. U.S. Utility PatentsPatentee orPatent or Pub. Nr.Kind CodeIssue or Pub. DateApplicant5,085,825B11992 Feb. 4Gluntz et al.7,873,136B22011 Jan. 18Meseth8,045,671B22011 Oct. 25Meseth8,559,583B12013 Oct. 15Sato8,347,628B22013 Jan. 18Gerard5,537,822B21996 Jul. 23Shnaid et al4,100,745A1978 Jul. 18Gyamathy 1. AMERICAN NUCLEAR SOCIETY, Special Committee on Fukushima, Fukushima Accident 2011, March 2012, http://fukushima.ans.org/. 2. HANSEN, JAMES ET AL., Climate Change Experts Endorse Nuclear Power, World Nuclear News, Nov. 4, 2013, http://www.world-nuclear-news.org/EE-Nuclear-essential-for-climate-stability-0411137.html. 3. CAES Compressed Air Energy Systems, http://en.wikipedia.org/wiki/Compressed_air_energy_storage, Huntorf, Germany 290 MW CAES 1978, McIntosh, Ala. 110MW CAES 1991. 4. WESTINGHOUSE ELECTRIC CO, LLC, The AP1000 Nuclear Plant Design, http://ap1000.westinghousenuclear.com/index.html, Cranberry Township, Pa., USA. 5. U.S. NUCLEAR REGULATORY COMMISSION, AP1000 Safety Evaluation Report, NUREG-1793, September, 2011 (http://www.nrc.gov/reactors/new-reactors/design-cert/ap1000.html) 6. Thermal Power Station, https://en.wikipedia.org/wiki/Thermal_power_station#Superheater, Wikipedia Gluntz and Meseth both show “accumulator” tanks containing water and a well-known neutron absorbing or “poisoning” material to halt or slow a nuclear reaction. The primary purpose of the accumulators is to inject poison neutrons into a reactor vessel. This rapidly shuts down the fission reaction within the reactor in case the control rods do not function properly, or minimizes the fission reactions after a rapid shutdown of the reactor. The accumulator tanks are limited in size and internal pressure by their inherent bursting strength. Because of this size limitation, the amount of water stored in an accumulator tank is too small to perform any major cooling of a reactor. The operation of an accumulator tank is dependent on infrastructure and piping inside the nuclear plant and, as such, it will be disabled if critical elements of the infrastructure are disabled. In addition, these tanks cannot deliver enough cooling fluid to cool a reactor at peak decay heat production for more than a few minutes or hours. Sato describes the operational features of the above-mentioned Westinghouse AP1000 passive ECCS system. Then he describes an improvement over the AP1000. He shows a Depressurization and Cooling System (DPCS) that is attached to the AP1000 design to cool the reactor to “cold shutdown condition.” The DPCS uses compressed gas to propel water out of an auxiliary water tank. Sato also shows that the water tank for his DPCS can be outside the reactor Primary Containment Vessel (PCV) and optionally buried underground. However, this DPCS apparatus is connected to and part of the internal plumbing in a nuclear plant Like the AP1000, Sato's DPCS system requires primary coolant to be circulated through the reactor and associated plumbing attached to the reactor to cool the heated coolant and hence the reactor. Gerard describes a CAES system that utilizes a common multi-stage heat exchanger to transfer heat energy from an ambient temperature water reservoir into the compressed air. Such heat exchangers have been common for many years in thermal power plants where lower pressure steam is circulated through pipes inside the thermal plant furnace to raise the temperature of the steam before it goes into lower pressure stages of the steam turbine. Such a heat exchanger is called a “Reheater”. Shnaid claims to use waste heat for the heat of expansion (HOE) for the compressed air in his CAES system. But his waste heat comes from an additional natural gas turbine that is used to supply the HOE needed by the compressed air. Gamarthy uses steam generated by electrical power as the HOE. As an alternative, he bleeds process steam from a thermal plant to use as HOE for the compressed air. In both cases, this steals energy from the thermal plant and it costs more money. The Gamarthy design still generates additional CO2 pollution in order to supply the HOE needed by the compressed air released. The Thermal Power Station Wikipedia entry discusses power stations in general, including the types of such stations and their components. Many of the pressurized accumulator tanks used in current nuclear plants use heating elements to generate water vapor, which is used to provide a pressurizing fluid above the stored water. This is preferable to using stored compressed gas at high pressure inside a nuclear plant because a ruptured gas tank could present a significant safety issue. However, the electrical power required for such heating elements may not be available in a disabled nuclear power station. While each of the above systems may be suited for their particular use, all have one or more deficiencies as noted. 100Nuclear Power Plant101Power plant building102Reactor pressure vessel104Primary containment(RPV)vessel (PCV)106Tank108Water110Pressure112Tank114Pressure116Heat exchanger118Ground level120Pipe122Valve124Pipe126Valve128Expansion valve130Coil132Volume134, 136,Pipe142Valve138, 140144Spent-fuel pool146Pipe148Valve200Casing205Grout210Plug215Lid220Gasket225Blocks230Bolts235Brackets240Welds245Pipe255Valve260Handle265Pipe270Valve290Pipe295Pipe300Manifold305Valve310Valve315Pipe320Pipe325Valve330Valve335Valve350Manifold400CGES units405Pipes410Valves415Tanks420Tank425Valve500Railroad tank car505Track510Tanks515Manifold520Pipes600Railroad flat car700Housing705Trench710Fence715Roof720Tracks800CAES system805Cavern or Reservoir808Heat exchanger810Compressed Air Heat812CoilsExchanger814Pipe815Turbine820Generator822Conductors825Compressor830Pipe835Valve840Turbine845Generator850, 855Conductors870Valve875Pipe881Pipe900Fuel rods905Baffle910Manifold915Openings920Pool925Pipe930Valve935Pipe940Valve945Pipe950Valve1100Plate1105Plate Overview In one embodiment, CG is safely stored at high pressure in one tank and is used to pressurize cooling water in another, larger tank. The CG is admitted to pressurize the cooling water when it is needed during a nuclear emergency. In one aspect the heat required for rapid expansion of large volumes of the CG is extracted from the water that is pressurized by the expanding gas. Equal volumes of water and expanded gas pass through a heat exchanger so that there is always sufficient heat energy for the gas to expand. The temperature drop in the water due to contact with the expanding gas is small. In another aspect, very large quantities of CG from a CAES plant are used to cool a companion nuclear reactor by expansion of the compressed gas alone inside the reactor during emergencies. This aspect provides great cost savings to the CAES during normal power generation that pays for the CAES installation near a companion nuclear power station. A CAES is a relatively inexpensive power plant (3% or less of the cost of a nuclear plant) that can be installed outside existing or new nuclear power stations. The NCAP embodiment provides greatly enhanced safety for a nuclear plant and major cost and pollution reduction for a companion CAES plant that have heretofore not been available. The cost reduction and improved profitability for the CAES plant pay for combining the two power plants in one location as described in connection with this embodiment. The NCAP is a passive safety system in that it does not require any other external power supply or energy source other than the compressed gas stored in the CAES plant. The NCAP is a passive safety system that does not require any other external power supply or energy source. There is an additional advantage inherent in the NCAP embodiment. If the ECCS safety systems in a nuclear plant are operational but the plant has lost its emergency backup power (as happened at Fukushima), the companion CAES system can immediately supply the backup electrical power needed by the ECCS systems in the companion nuclear plant. The examples used herein are calculated to meet the emergency reactor cooling requirements for a 1300 MW electrical nuclear power station with a nuclear reactor that produces 3900 MW of thermal heat energy. The decay heat from the reactor fuel rods immediately after the reactor shutdown is assumed to be 7% of the rated 3900 MW thermal, equal to 273 MW immediately after shutdown, reducing to 1.5% after one hour. These values are used for explanatory purposes and are not intended to be limiting in any way. Those skilled in the art of nuclear power plant design will understand how to scale the emergency cooling system described herein to fit smaller and larger nuclear power plants. The NCAP system eliminates the major problems with prior CAES designs. The NCAP system uses some of the vast amount of (free) waste heat generated by a thermal power plant (TPP) to supply the heat of expansion (HOE) needed by the compressed air to generate power. The TPP waste heat replaces the costly natural gas and other energy sources used by other CAES designs to supply the HOE. The connection of a standard power plant and a nearby companion compressed air storage (CAES) facility provides a “new style” CAES that allows the capture and storage of a great deal more wind, solar, and other intermittent renewable energy on grids all over the world. The NCAP system also provides improved TPP plant efficiency. The NCAP system can be installed all over the world in a few years using only existing hardware and infrastructure not decades from now when it may be too late to moderate climate warming. More nighttime wind and daytime solar energy is available in the world than the total energy requirements of the world. Unfortunately, many wind turbines are idle at night because there is no place to store the energy that is not needed on a grid at the time. Solar power generation is also limited to what can be used during the day. The amount of waste heat thrown into the environment by existing thermal power plants in the world is almost twice as much as the electrical energy produced by all power plants. If even a small part of these immediately available untapped sources of energy, wind and solar energy and thermal plant waste heat, could be converted to useful power, the world could replace most of its CO2 emitting fossil fuel power plants and make a big reduction in climate warming. However, capturing enough additional renewable wind and solar energy to replace even one half of the existing fossil fuel power plants will require a thousand or more large renewable energy storage (RES) facilities installed on the electrical grids of the world. The NCAP system uses waste heat from conventional thermal power plants (TPP) to provide compressed air renewable energy storage for wind and solar energy by reducing the cost and pollution generated by compressed air stored in a CAES. The NCAP system provides an inexpensive and reliable utility-scale compressed air energy storage (CAES) facility that the world can install quickly on almost all electrical grids in order to capture and store a great deal more wind, solar, and other intermittent renewable energy. There are thousands of TPPs in the world where companion CAES facilities can be added to provide NCAP systems. As explained below, the NCAP combination provides a double benefit. Some of the waste heat from a TPP that is transferred to compressed air is converted to useful power in the compressed air turbine. As a result, this increases the net energy efficiency of the TPP plant. When the TPP plant in a NCAP system is a nuclear power plant, the CAES in the system provides an additional safety factor. In one aspect, very large quantities of CG from a CAES plant are used to cool the fuel rods in a companion nuclear reactor by expansion of the compressed gas alone inside the reactor during emergencies. The value of this feature alone pays for the CAES installation near a companion nuclear power station. A CAES is a relatively inexpensive power plant (3% or less of the cost of a nuclear plant) that can be installed outside existing or new nuclear power stations. The NCAP is a passive safety system in that it does not require any other external power supply or energy source other than the compressed gas stored in the CAES plant. The NCAP embodiment provides greatly enhanced safety for a nuclear plant and major cost and pollution reduction for a companion CAES plant that have heretofore not been available. There is an additional advantage inherent in the NCAP embodiment. If the ECCS safety systems in a nuclear plant are operational but the plant has lost its emergency backup power, as happened at Fukushima (1), the companion CAES system can immediately supply the backup electrical power needed by the ECCS systems in the companion nuclear plant. The examples used are calculated to meet the emergency reactor cooling requirements for a 1300 MW electrical nuclear power station with a nuclear reactor that produces 3900 MW of thermal heat energy. The decay heat from the reactor fuel rods immediately after the reactor shutdown is assumed to be 7% of the rated 3900 MW thermal, equal to 273 MW immediately after shutdown, reducing to 1.5% after one hour. These values are used for explanatory purposes and are not intended to be limiting in any way. Those skilled in the art of nuclear power plant design will understand how to scale the emergency cooling system described herein to fit smaller and larger nuclear power plants. FIGS. 1 to 4 show aspects of one CGES embodiment. In one aspect of this embodiment, unpressurized water in a large tank is pressurized by compressed gas that is kept at high pressure in a separate tank. The water is pressurized only when it is needed during a nuclear emergency. The heat required for rapid expansion of large volumes of the compressed gas is taken from the pressurized water as it circulates around an expansion valve and coil combination and through a heat exchanger. Since equal volumes of water and expanded gas pass through the heat exchanger in a given time, the temperature drop in the water is small. FIG. 1 shows a schematic diagram of one aspect of a CGES system, while FIG. 2 shows one physical realization of the system of FIG. 1. The system comprises a nuclear power plant with a reactor pressure vessel (RPV) 102 that is housed within a primary containment vessel (PCV) 104. A first storage reservoir tank 106 contains water 108 at ambient pressure 110. A second storage tank 112 contains CG at high-pressure 114, a heat exchanger 116, and a plurality of valves and interconnecting pipes. Tanks 106 and 112, heat exchanger 116, and a portion of interconnecting pipes are located beneath ground level 118. A pipe 120 extends from tank 112 to a valve 122 and then to a point above ground level 118. Another pipe 124 connects to pipe 120 at a point between valve 122 and tank 112. Pipe 124 connects pipe 120 to heat exchanger 116 via another valve 126. Heat exchanger 116 comprises an interior volume and has a pressure regulator and expansion valve 128 and a coil 130. Valve 128 is one of many standardized pressure regulator and expansion valve designs commonly used in the compressed gas industry. Valve 128 is connected to pipe 124 at the right-hand side of valve 126. Pipe 138 delivers pressurized water from the bottom of tank 106 to volume 132 in heat exchanger 116. Additional pipes 136, 138, and 140 extend from heat exchanger 116 to tank 106 and plant 100. Pipe 136 extends from coil 130 in heat exchanger 116 to the top of tank 106. Pipe 138 extends from a point near the bottom of tank 106 to volume 132 in heat exchanger 116. Pipe 140 extends from internal volume 132 of heat exchanger 116 to plant 100 via a valve 142. Expansion valve 128 is shown inside heat exchanger 116 to depict that it is constantly surrounded by water in heat exchanger 116. Valve 128 is of sufficient design and surface area to absorb the heat needed to keep its components at acceptable operating temperatures. Coil 130 depicts additional surface area for heat transfer that may be an integral part of valve 128. Pipes 120 and 134 extend above ground for access to CG for tank 112 and water for tank 106. After their usefulness in a nuclear reactor is over, nuclear fuel rods still generate heat from nuclear decay and are usually stored in a cooling water bath to prevent overheating. Spent fuel rods are removed from reactor 100 and stored in a spent-fuel pool (SFP) 144. Emergency cooling water is delivered to pool 144 by a pipe 146 when valve 148 is opened. Pipes 120, 124, 134, 136, 138, and 140 are securely and sealed to tanks 106 and 112, heat exchanger 116, and plant 100 at their points of entrance and exit. Operation—FIG. 1 In FIG. 1, the external ECCS system pumps cooling water into plant 100 during a nuclear emergency to prevent a meltdown when in-plant Emergency Core Cooling systems are disabled or damaged. It can inject large volumes of cooling fluid immediately into RPV 102 even if there is still pressure in the RPV, so long as the pressure in tank 106 is greater than the water-steam pressure in the RPV. The pressure in tank 106 can be as high as 20.7 bar (300 psi) for a tank 106 holding over 100 m3 of cooling water (see design example below). One of the most dangerous situations in a nuclear plant is sudden loss of cooling water in the RPV and its nuclear fuel rods are no longer covered by cooling fluid. The RPV pressure can drop well below 20.7 bar. The CGES system can immediately inject cooling water at a rate as high as 6500 l/m to absorb all the decay heat from a 1300 MWe (3900 MW thermal) nuclear reactor (273 MW initial rate after SCRAM). A full ECCS system to provide up to three days of reactor cooling will comprise many CGES modules as shown in FIG. 1. Valve 126 is first opened manually or automatically allowing high pressure gas (CG) from tank 112 to enter expansion and pressure regulator valve 128 in heat exchanger 116. Gas flows from valve 128, through coils 130, through pipe 136, and into tank 106. The pressure exerted on the surface of the water in tank 106 urges the water upward through pipe 138, into volume 132 of heat exchanger 116, and out into pipes 140 and-or 146. When valve 142 is manually or automatically opened, water flows outward through pipe 140 and onward to manifold 300 of plant 100. The water delivered under pressure to manifold 300 is sent to RPV 102 or primary containment vessel, 104, as needed, by selectively manually or automatically operating valves 305, 310, and 325 (FIG. 3). FIG. 2—Description FIG. 2 is a cross-sectional view that shows a physical example of one aspect of the system of FIG. 1. FIG. 2 shows additional components that strengthen the design in FIG. 1 in order to contain the high pressures during both storage and operation. The present aspect shown is an underground version of the CGES system. In this embodiment, CG tank 112 is mounted inside cooling water tank 106. This is an embodiment that allows both tanks to be contained in one bore hole that is constructed by standard oil-gas well drilling techniques. When valve 126 is open, CG from tank 112 flows into pressure reduction valve 128. Control device 260 adjusts the pressure of CG leaving valve 128 to keep the pressure in tank 106 below the maximum allowable level. Valve 128 is manually or automatically adjusted to set the pressure of CG from tank 112 entering coils 130 of heat exchanger 116. Safety relief valve 255 assures that the maximum allowable pressure in tank 106 is not exceeded. Adjustable flow rate valve 142 controls the rate of pressurized water leaving tank 106 which determines the rate of expanded CG allowed into tank 106 when valve 126 is open. Water flowing from tank 106 through volume 132 of heat exchanger 116 provides heat needed to prevent freezing of components in 116 or water in tank 106 by the expanding gas from tank 112. A tank 106 is formed within a borehole of predetermined depth and diameter in the earth. The tank has a cylindrical steel casing 200 that is open at both ends and has a diameter less than that of the borehole. During installation, casing 200 is lowered into the borehole and securely held a short distance above the bottom. Standard grouting procedures from in the oil-gas well drilling industry are used to insert concrete grouting material 205 between casing 200 and the bottom interior of the borehole to form a concrete plug 210 at the bottom of the borehole and grouting 205 around casing 200. Grouting 205 and plug 210 are of sufficient size and composition to prevent significant leaking of fluids contained therein or damage to casing 200 at the operating fluid pressures utilized. A lid 215 closes tank 106 at the upper end of casing 200. A gasket 220 or other sealant means is placed between the top of casing 200 and the bottom surface of lid 215 to prevent gas leakage when the present system is activated. A number of concrete blocks 225 are placed atop lid 215 to further weigh it down against the top of casing 200 and gasket 220, thereby forming a tight seal for tank 106. A plurality of bolts 230 further secure lid 215 to casing 200. In this aspect, bolts 230 bear against brackets 235 that are secured to casing 200 by a plurality of welds 240. When bolts 230 are tightened, lid 215 is clamped securely in place on tank 106. At the upper right of FIG. 2, an additional inlet pipe 245 is connected to pipe 134A via a valve 250. Inlet pipe 245 is connected to an outside source such as RPV 102, primary containment vessel (PCV) 104, spent-fuel pool (SFP) 144 or another source for return or delivery of water to tank 106. Water is also delivered to tank 106 through pipe 134A via valve 136A, heat exchanger 116, and pipe 138. A vent line 290 and a valve 295 are used to vent air from tank 106 when fluid is added via pipes 134 and 138 or pipe 245. Valve 295 is opened to vent displaced air from tank 106; otherwise it is closed. The fluid returned from plant 100 to tank 106 can be the hot fluid generated in the RPV from the cooling water initially injected by this tank 106 or other tanks 106 into plant 100. In this manner, the reactor cooling capacity of water initially stored in tanks 106 can be extended. A few days after SCRAM, the decay heat rate falls below 0.4% of rated reactor thermal power. When the heat transfer rate through the casing 200 of one of more tanks 106 equals or exceeds the decay heat rate in a reactor, a CGES system comprising several tanks 106 can cool the reactor indefinitely by recycling the cooling water in tanks 106. Underground tank 106 that is in a deep bore hole as shown in FIG. 2 can maintain a substantial heat transfer rate through its steel casing 200 and concrete grout 205, especially if the bore hole passes through underground aquifers where water is in constant contact with the outside grouting 205. The above cooling water recycling capability of the present CGES system is important in comparison to the AP1000 system, supra. The AP1000 cannot return a reactor to cold condition where the coolant is less than 100° C. The AP1000 must have continuous hot fluid in the reactor vessel to maintain circulation of the primary coolant through a heat exchanger that uses the gravity water from the AP1000 system tank to absorb decay heat from the hot primary coolant. In contrast to the AP1000, the CGES, by recycling its cooling water as described, can cool a reactor below 100° C. and maintain that condition long-term. Moreover the CGES does not require Sato's DPCS apparatus, supra, to bring the reactor to cold shutdown condition. Both the AP1000 and Sato's apparatus use reactor primary coolant circulation to cool the reactor. They are critically dependent on plumbing internal to the nuclear plant, but such plumbing can be disabled by attacks by nature or terrorists. In contrast, the present CGES cooling operation does not use the primary boundary plumbing connected to the reactor pressure vessel (RPV). The CGES system is effective even if the internal plumbing is damaged and cannot retain coolant in the primary boundary. The CGES system injects new coolant into a reactor through the external coolant input port and constantly replaces the coolant in the reactor anytime it is not covering the fuel rods in the reactor. The CGES coolant is constantly evaporated to absorb maximum decay heat and then vented out of the reactor pressure vessel (RPV) as water vapor, which is then replaced by more CGES coolant. The CGES system works best when any hot primary coolant is vented out of the RPV after a SCRAM so that the injected CGES coolant is not wasted absorbing the heat from the left-over primary coolant. The safety pressure release valve 255 below lid 215 allows venting of excess pressure within tank 106 whenever the pressure exceeds a pre-set maximum allowed pressure (usually less than 21 bar for a tank 106 greater than 1 meter in diameter). A pipe 260 extends from valve 255 to the air above ground level for venting. In various aspects an embodiment can have two or more safety relief valves 255 mounted on tank 106 for redundancy in this critical component that protects tank 106 from excessive pressures. In the present aspect, gas expansion valve 128A includes a handle 260 for adjustment of the output pressure of valve 128A. Valves 126, 142, 148, and 270 all are one-way (check) valves that allow flow only in the direction of the arrows shown in the respective lines. These check valves prevent backflow into the lines in case of excessive pressures on the other side of the valve, i.e., as in line 140 where excessive pressure in the RPV or PCV could force water back into tank 106 and damage it. Pipe 265 allows the CG in tank 112 to be directed elsewhere when valves 126 and 270 are open. For example, when the CG in tank 112 is nitrogen, the nitrogen can be used to recharge nitrogen supplies that are used to operate valves in a nuclear plant or to flood a reactor vessel with inert nitrogen to reduce known hazards of oxidation of metals in the fuel rods and to prevent an explosion of hydrogen gas. Exemplary Design Specifications and Dimensions for Tanks 106 and 112. The following are specifications for tanks 106 and 112 in an exemplary CGES system. 1. Tank 106 has a volume of 100 m3 (cubic meters). 2. Tank 112 has an initial CG pressure of 200 bar (3000 psi) and a maximum pressure of 20.7 bar (300 psi) in tank 106 during a nuclear emergency to limit internal hydraulic forces in tank 106 (as controlled by valve 128A in FIG. 2). 3. The volume ratio of tank 106 to tank 112 is 10:1 (100 m3 of water to 10 m3 of CG). 4. Cooling water must be delivered from tank 106 at a rate of 65521/m (1731 gpm) to handle the maximum decay heat rate of 273 kJ per sec (7%) from a 3900 MW thermal reactor immediately after SCRAM (assuming 2,500 MJ per m3 heat absorbed by injected cooling water raised from 20° C. to full evaporation in the reactor). 5. Tank 106 has an inside diameter of 1.19 m (47 in) and is 100 meters deep with a gross volume of 112 m3. The wall thickness of tank 106 is 0.95 cm (0.375 in). The critical stress in the wall of tank 106 with internal gas pressure 20.7 bar (300 psi) is 1,296 bar (18,800 psi). (This assumes no support from the concrete grouting outside the well casing in the case that tank 106 is in a borehole.) 6. Tank 112 has an inside diameter of 0.357 m (14.06 in) and is 100 meters deep to yield a volume of 10 m3. The wall thickness of steel pipe for tank 112 must be at least 2.68 cm (1.054 in) to keep the critical stress in the wall of this tank no greater than 1,379 bar (20,000 psi) for internal gas pressure of 20.7 bar (3000 psi). 7. The maximum expansion of the CG from tank 112 will be 11:1 when tank 106 is empty. The final pressure in tank 106 will be 200 bar divided by 11=18.2 bar (264 psi). This is enough remaining pressure to lift the last water in tank 106 up to a height of 186 m (609 ft). 8. At a pressure of 20.7 bar (300 psi) in tank 106 with a diameter of 1.19 m (47 in), the upward force on lid 215 (FIG. 2) is 2,315,300 N (520,500 lb). The 1.19 m diameter casing 200 of tank 106 with 0.95 cm thick steel wall can resist an upward force of 4,525,542 N (1,017,400 lb) with a maximum 1,379 bar (20,000 psi) stress in the casing wall when lid 215 is securely attached to casing 200. This is 2.1 times the actual load force in the casing wall due to the upward force of 2,315,300 N, for a safety factor of 2.1. The pressure in tank 106 could increase to 2.1 times 20.7 bar=43.5 bar and not exceed the stress limit of 1,379 bar in the wall of tank 106. 9. Concrete weight 225 on tank 106 should be 444,800 N (100,000 lb) as an added safety factor to resist the upward hydraulic force on the top of tank 106 and reduce the load on the outer part of lid 214. Fifty tons of concrete requires 19.12 m3 (25 cu yd) of concrete at 11,518 kg per m3 (4,000 lb/cu yd). A heat transfer rate of 10. 25 KW into the CG is required for it to expand by 10:1 at a rate of 0.1092 m3/s (1800 gpm) to propel water out of tank 106 at the same volume rate. 25 kW is 25 kJ/s taken from the water leaving tank 106 at the rate of 0.1092 m3/s. The specific heat of water is 4.18 MJ/m3/deg C. The CG takes only 25 kJ from each 0.1092 m3 of water that flows through heat exchanger 116. This is a rate of 228 kJ/m3 of water. The temperature drop per m3 of water is 228 kJ/m3 divided by 4.18 MJ/m3/° C.=0.05° C. Thus the water temperature drops by far less than 1° C. as it flows out of tank 106. A person skilled in the art of physics or engineering can scale the above design example to other workable dimensions for tanks 106 and 112 and CG pressures as appropriate. All metal components shown in FIG. 2 are steel or another alloy that is sufficiently strong to meet the design requirements of the CGES and are not susceptible to corrosion in the environment shown. Connection of CGES to Plant 100—FIGS. 2 and 3. FIG. 3 shows a schematic diagram of a nuclear power plant that illustrates paths that cooling fluids take after SCRAM. When a CGES is activated, water leaves tank 106 (FIG. 2) and is delivered to plant 100 via pipe 140 or inlet manifold of plant 100. A shut-off valve 305 on manifold 140 is opened to admit water into plant 100. A pipe 315 conducts water to a valve 310. A pipe 320 branches from pipe 315 and conducts water to a valve 325. When valve 310 is open, water is directed into RPV 102. When valve 325 is open water is directed into primary containment vessel 104. Either or both of valves 310 and 325 can be opened, depending on cooling requirements after a SCRAM in plant 100. Additional valves 330 and 335 are opened as appropriate to conduct water and steam from RPV 102 and primary containment vessel 104 to a manifold 350 within plant 100. Manifold 350 is connected to pipe 245 (FIG. 2) so that water and steam from plant 100 can be returned to tank 106 if desired. In the following example a CGES module as shown in FIG. 2 has been installed and connected to a plant 100 (FIG. 3). All aboveground connections to the CGES are sealed to prevent unwanted matter from entering the system. All valves in the CGES system are initially closed. Preparing CGES for Use. A source of water (not shown) is connected to pipe 134A and valve 136A is opened. A predetermined amount of water flows through pipe 134A and volume 132 of heat exchanger 116 and into tank 106. When the amount of water supplied fills approximately 95% of tank 106, valve 136A is closed, stopping the delivery of water to tank 106. A source of high-pressure gas (air, nitrogen, or other gas) is connected to pipe 120 and valve 122 is opened. Gas is urged from the source into tank 112. When a predetermined pressure is reached, valve 122 is closed, stopping the delivery of gas into tank 112. When the delivery of gas and water to the CGES system is complete, the sources of gas and water are optionally removed or left in place. Activating CGES after SCRAM. As with the system of FIG. 1, assume that a disaster occurs so that the reactor must be shut down under emergency conditions (SCRAM). To prevent a meltdown, the present system pumps cooling water into plant 100. Valve 126 is first opened manually or automatically allowing gas from tank 112 to enter expansion and pressure regulator valve 128A in heat exchanger 116. Gas flows from valve 128A, through coils 130, through pipe 136, and into tank 106. The pressure exerted on the surface of the water in tank 106 urges the water upward through pipe 138, into volume 132 of heat exchanger 116, and out into pipe 134A. When valve 142 is opened manually or automatically, cooling water flows outward through pipe 140 and onward to manifold 300 of plant 100. The water delivered under pressure to manifold 300 is sent to RPV 102 or primary containment vessel, 104, as needed, by selectively operating valves 305, 310, and 325 (FIG. 3) manually or automatically. Valve 128A control device 260 is manually or automatically adjusted to control the pressure of CG leaving valve 128A so that it does not exceed the maximum pressure allowed in tank 106. Other CGES Activities—Delivering Water to a SFP A pipe 146 (FIG. 3) is connected to a SFP (not shown) for delivery of cooling water to any spent fuel rods that still give off heat and must be cooled when normal plant 100 cooling water circulation is disabled. Starting with all valves closed and sufficient gas in tank 112 and water in tank 106, valves 126 and 260 are opened and valve 260 is adjusted so that pressure inside tank 106 is sufficient to urge water to flow out pipe 146. Valve 148 is then opened and water flows to SFP from tank 106 until either tank 106 empties or sufficient water is delivered. Valve 148 is then closed and gas and water supplies to the CGES are restored, if necessary. Recovery and Recycling of Water and Steam from Plant 100 In FIG. 2, a pipe 245 is connected to outflow manifold 350 of plant 100. Initially, all valves are closed. When it is desired to return water from plant 100 to a tank 106, valve 250 is opened and water and-or steam flows from plant 100 back into tank 106. This feature allows the venting of the initial high pressure water and steam in the SCRAMMED nuclear reactor back into empty tanks 106 instead of venting the radioactive steam into the environment. This is extremely important when it is necessary to inject cooling water into the reactor that has reached its maximum heat content and internal pressures because it cannot circulate and cool its existing cooling fluid. By this means, the pressure and water-steam in the reactor can be reduced to allow injection of new cooling water from tanks 106. Return pipe 245 also allows the CGES embodiment in FIGS. 2-5 to perform long-term cooling of a the reactor beyond the critical first days after SCRAM. Cooling water from a tank 106 that is heated in and then evaporated out from the reactor can be returned to other empty tanks 106 for later reuse as cooling water. In this manner, the cooling capability of an initial amount of water in a set of tanks 106 can be extended indefinitely after the decay heat rate in the reactor drops low enough so that cooling water from tanks 106 can be recycled and condensed in empty tanks 106 faster than it is needed to cool the reactor. In particular, deep underground tanks 106 as shown in FIG. 2 can provide substantial cooling for water and-or steam recycled from the reactor as described above. This CGES embodiment eventually can cool a reactor below the desired “cold shutdown” condition whereby the temperature within the reactor is less than 100° C. The CGES system is a substantial improvement over the AP1000 system. The AP1000 system requires continuous primary cooling water circulation through a reactor. The CGES embodiment can cool a reactor even after its primary cooling water has been lost. Delivery of Gas to Another Location A pipe 265 in FIG. 2 branches from pipe 124 and is arranged to deliver compressed gas to other locations. For instance, if the compressed gas is nitrogen, the nitrogen can be delivered to storage tanks within a nuclear plant 100 for nitrogen actuated control valves. When it is desired to deliver gas through pipe 265, valves 126 and 270 are opened and gas from tank 112 exits the CGES via pipe 265. Closing either valve 126 or 270 stops the delivery of gas. FIG. 4 is a schematic plan view of one aspect in which a plurality of CGES units 400 are connected to a nuclear plant 100. This arrangement provides redundancy in case one or more CAES units are damaged or inaccessible. A plurality of pipes 405 connect the water storage volume in each of tanks 106 (FIG. 1) to that of its neighbors. Another plurality of pipes 140 are arranged to deliver water to plant 100 when required after a SCRAM, as explained above. An additional plurality of pipes 245 are arranged to pass water and steam back from plant 100 to CGES units 400. Manual on-off valves 410 are installed at the beginning and end of each pipeline segment so that a damaged or leaking pipeline segment can be isolated by closing the appropriate valves 410. Some or all of valves 410 are automatic if required. Additional pipes 120 and 134A permit CG and water from external sources to be injected into tanks 112 and 106, respectively. Additional tanks 415L and 415R contain liquid nitrogen that is delivered to pipes 120 via pipes 420L and 420R, respectively. The liquid nitrogen in these tanks is used to recharge tanks 112 106 (FIG. 1) as necessary. Additional tanks 420L and 420R are above-ground, standard water storage tanks of the type commonly employed by water utility districts for bulk storage of water. Their capacity is typically 3,700 m3 (1,000,000 gal) or more. Tanks 420L and 420R replenish the water in tanks 106 in CGES units 400 (FIGS. 1 and 2) for long-term reactor cooling operations after the critical first days after reactor shutdown. Valve 425L controls water flow from tank 420L to a tank 106 in a CGES unit 400 by line 134A. Valve 425R controls water flow from tank 420R to the input line 134A to a tank 106 in a CGES unit 400. Many tanks 420L or 420R can be stationed around a nuclear plant and connected as shown to any or all of tanks 106 in CGES units 400. Water is stored in tanks 420 at ambient atmospheric pressure, and flows from tanks 420 to tanks 106 by gravity. When no pressure is in the reactor and it is underground below the level of tanks 420, water from these tanks can be sent directly to the reactor by appropriate settings of pipeline valves shown in FIG. 4. All pipes are placed on or under open buffer areas or beneath parking areas around plant 100 so that there is no loss of surface space around plant 100. FIGS. 5 to 7 show aspects of an embodiment of the CGES system that uses standard railroad (RR) tank cars for transport an above-ground version of water storage tank 106 (FIG. 2). With CG tanks 112 attached to each tank car or supplied separately, this provides a simple, inexpensive, and very reliable CGES system that can be installed immediately at most any nuclear power station. This CGES system embodiment is called a RRTC (Railroad Tank Car) system. The RR tank car CGES modules comprise portable units that can be transported from one location to most any other nuclear plant in trouble within 24 hours. FIG. 5 is a schematic view showing a plurality of railroad tank cars 500A through 500D in the vicinity of a nuclear plant 100. Cars 500 are moved on a plurality of track segments 505 that are located at predetermined positions and distances in the vicinity of plant 100. Although four tank cars are shown, fewer or more can be used. Railroad tank cars can hold up to 128.7 m3 (34,000 gal) of liquid, with some older tank cars holding up to 189.27 m3 (50,000 gal). Although rail cars 500 serve the function of tank 106 (FIG. 2), they have greater capacity than the 100 m3 capacity of tank 106 used in the design calculation example above. A plurality of tanks 510 are mounted on each rail car. Tanks 510 serve the function of tank 112 (FIG. 2). In one aspect, each of tanks 510 is a 15.2 m long, 0.36 m in diameter high-pressure steel pipe, and there are between 6 and 8 tanks 510 on each rail car. The combined volume of the 6 to 8 tanks 510 is about equal to that of tank 112 used in the design example above to provide at least 10 m3 of CG at 200 bars (2900 psi) of pressure. Railroad tank cars can withstand internal gas pressures up to 6.9 bars (100 psi). That is more than enough pressure to force water at high flow rates into plant 100 during a nuclear emergency. In FIGS. 5 to 7, the same pipelines and valves that interconnect CG tank 112 to water tank 106 in FIG. 2 (with the same reference numbers) are used to connect tanks 500 and 510. Although present as indicated in FIG. 2, valves, other than valve 126, are not shown in FIGS. 5 to 7. In FIG. 5, tank car 500A is expanded to better show the components used in this aspect of the present embodiment. The rail car tank on cars 500 is equivalent to tank 106 in the first embodiment. Tanks 510 are connected to a sealed manifold 515 by a plurality of pipes 520. Tanks 510, pipes 520, and manifold 515 are collectively equivalent to tank 112 in the first embodiment. When valve 126 is opened, CG flows from manifold 515, through pipe 120 and valve 126 into expansion and pressure regulator valve 128A (FIG. 2), indicated by pressure adjusting handle 260, through heat exchanger 116, and out via pipe 136 to pressurized tank 500A. Pressurized water flows out line 138 from tank 500A into heat exchanger 116 and then out pipe 140 to plant 100. Thus pressure is applied to the interior of tank 500A via pipe 136 and water is urged to leave tank 500A via pipe 138, passing through volume 132 of heat exchanger 116, and finally on to plant 100 via pipe 140. The RRTC system sacrifices some protection against assault by acts of nature or terrorism that is inherent in the underground CGES system of the first embodiment. However the immediate availability of many railroad tank cars and flat cars around the industrialized world makes the RRTC system a very practical and inexpensive CGES that can be protecting a nuclear power plant within a few months from the start of installation, at very little cost. The RRTC has all the operational characteristics of the other CGES embodiments described above. Second Version—FIG. 6 FIG. 6 is a schematic diagram showing an alternative form of the tank car system of FIG. 5. In this aspect a plurality of tanks 510A serves a plurality of rail cars 500E and 500F. Although two cars are shown, more can be added. Tanks 510A are a stack of pipe sections that terminate in sealed manifolds 515A and 515B at each end of the stack. Tanks 510A are mounted on a flat rail car 600. CG is released from tanks via pipe sections 124 when valves 126 and 128 (not shown in this drawing) are opened. When the CG is released, CG pressure is applied to tanks 500 via pipes 136, water is forced out of tanks 500 through pipes 138, through heat exchangers 116, and out through pipes 140 to nuclear plant 100 (not shown in this figure). To provide enough CG to pressurize at least six water storage tanks 500, 40 to 60 pipe sections 510A with dimensions described above should be provided. In the present arrangement, many railroad tanks 500 can be brought near flatcar 600 with tanks 510A and connected to lines 136, 138, and 140. When a water tank car is empty, it can be quickly disconnected from the CG source on flatcar 600 and replaced by another, thereby providing a continuing source of water to plant 100. Third Version—FIG. 7 FIG. 7 shows a third version, specifically a cross-sectional end view of a housing 700 that protects and conceals an RRTC. A safe parking and connection area for the RRTC and flat cars placed around a nuclear plant can be as simple as a shallow trench that has concrete walls on either side to protect the railroad cars from the high winds of a tornado and terrorist attacks with vehicles. Empty cars can be rolled away during a nuclear emergency and either refilled or replaced by previously filled cars. Tank cars 500 and flat cars 600 with CG storage tanks 510 are parked in a trench 705 dug into the ground to partially protect the RRTC system. A strong concrete fence 710 runs along both sides of trench 705. A roof structure 715 sits on fence 710 and spans across trench 705 above the rail cars. A plurality of railroad tracks 720 are laid on the ground at the bottom of trench 705. This embodiment of the RRTC system provides substantial protection from tornadoes and vehicular attacks by terrorists. The below ground level placement of tracks 720 also allows easy movement of replacement tank cars 500 and CG storage cars 600. This RRTC system is not expensive to construct. It requires mainly dirt excavation and placement of prefabricated fence sections 710 along the edges of trench 705. Fence sections 710 can be standard K-rail concrete sections used as traffic barriers on roadways, for example. Full CGES System for Large Nuclear Plant A full CGES Emergency Core Cooling System using the CGES embodiments shown in FIGS. 2-4 and FIGS. 5-7 will normally comprise many interconnected water tanks 106 and CG tanks 112 as shown in FIG. 4. For instance, the total decay heat from a 1300 MWe (3900 MWt) reactor for the first three days after SCRAM requires at least 22 tanks 106 each holding 100 m3 of cooling water plus as many companion tanks 112 holding 10 m3 of CG at 200 bar (or equivalent CG volume stored by other means). A smaller nuclear plant, say, 1000 MWe, would require only 17 tanks 106 of 100 m3 capacity each. The tanks 112 sources of CG shown in FIGS. 1 and 2 can be realized in many forms other than those shown in the embodiments described herein. Tanks 112 can be separate bore holes lined with steel nearby the tanks 106 that they pressurize. Above ground tanks 112 can be segments of high pressure pipe, such as used for high pressure natural gas pipelines. One of the least expensive ways to store and deliver large quantities of CG is to use many segments of small diameter, thick-walled pipe stacked together and connected as shown for item 510A on flatcar 600 in FIG. 6. CG tanks can be realized with large capacity plastic bags or bladders that are filled with CG and held deep underwater to balance the internal pressure of the CG. An advantage of this approach is that the CG is delivered at constant pressure equal to the water pressure on the bags. Inherent Safety of Compressed Gas Supply Systems. The CG energy sources for all embodiments of the CGES system can be replenished by external or emergency air compressors and compressed air sources such as the liquid nitrogen tanks 415L and 415R shown in FIG. 4. Attaching additional sources of compressed gas to an existing CG delivery system is both far easier and safer than attaching backup electrical generators to an existing power grid. Backup generators attached to a nuclear plant must be matched carefully in voltage and phase with the internal electrical system. However, using check-valve connectors, additional sources of compressed air, such as tanks 112 in the above embodiments, can be attached easily to an existing CG supply system with reduced concerns for safety or compatibility. CG will flow from the new system to the existing supply system only when the new system pressure is greater than the pressure in the existing system. The importance of this feature of the CGES system was demonstrated by the frustration of operators during the Fukushima accident in 2011 when they tried to connect outside backup electrical generators to the nuclear plant electrical switching equipment that had been disabled by the tsunami. NCAP—New Compressed Air Energy Storage (CAES) System—Description and Operation—FIGS. 8-12 The system described below is called NCAP for “New Compressed Air Power.” The NCAP system is a Compressed Air Energy Storage (CAES) system combined with a conventional thermal power plant that functions in several configurations. In one configuration, it is an efficient energy storage system that can be used to store energy, in particular, renewable energy, at any time so that the stored energy can be returned to an electrical grid at a later time. In another NCAP configuration, the CAES is an Emergency Core Cooling System (ECCS) for a nuclear power plant when its companion thermal power plant is a nuclear power plant. The system of FIG. 8 et seq. shows the NCAP system. The NCAP system combines an ordinary thermal power plant and a compressed air energy storage (CAES) facility to alleviate the problems and deficiencies of existing, old-style CAES systems. The NCAP system uses the waste heat generated by a thermal power plant (TPP) to replace the large amount of natural gas energy used by old-style CAES facilities to provide the HOE for the expanding compressed air. The connection of a standard thermal power plant and a nearby CAES provides a “new style” CAES that is much more cost efficient and non-polluting. In addition, the NCAP system allows conversion of some of the TPP's waste heat into useful power. Also, when the TPP plant is a nuclear plant, the NCAP system provides another measure of safety during a nuclear emergency. The large amount of stored compressed air in the companion CAES of the NCAP system can be directed into the reactor vessel during an emergency to cool the fuel rods, thereby preventing a meltdown disaster like Fukushima. The NCAP system uses existing hardware and infrastructure that can be combined and installed quickly on many electrical grids all over the world. FIG. 8 shows, for example, a TPP 100 with RPV 102 within a building 101. RPV 102 can also be a boiler that runs on fossil fuel, geothermal energy, gas, or molten salt or metal heated by solar power. It produces high pressure steam in line 802 that carries the steam into steam turbine 840. Below line A-A′ is an old-style CAES 800 which uses natural gas in a mixer 828 to supply the heat of expansion to the compressed air entering turbine 815. Heat exchanger 810 that of the NCAP system is also shown in FIG. 8. Prior to discussing the NCAP system, the operation of TPP 100 and CAES 800 without the NCAP apparatus will be described. TPP 100 can operate on fossil fuel, biomass, geothermal, or solar energy, as well as nuclear energy. In this example, TPP 100 is a steam-generating power plant which generates high-pressure steam that is supplied from RPV or other type of boiler to turbine 840 that drives a generator 845. Exhaust steam 860 (the waste heat fluid) comes out of the steam turbine 840 on line 860. The fluid in exhaust line 860 has a pressure of at least 2 bar and a temperature of at least 120° C. A “condenser” (heat exchanger) 808 cools the exhaust steam in line 860 and then returns the resulting condensate water 809 to the TPP's boiler. Virtually all of the heat in this steam is exhausted into the environment. (The external cooling medium for the secondary side of the condenser, heat exchanger, 808, usually cooling water, is not shown). This is the usual configuration for a standalone TPP plant. The waste heat generated and released to the environment by a TPP is conveyed in a waste heat fluid (WHF), such as steam or hot water, hot exhaust gases from a gas turbine, or molten metal used to store energy. Forty to sixty percent of the thermal energy generated in a TPP plant is waste heat. Ninety percent of the power plants in the world are steam generating TPPs. In all such plants thus far, more thermal energy is exhausted into the environment than is converted into useful electrical power. The apparatus below line A-A′ is an old-style CAES plant 800 with an added heat exchanger 810 and appropriate conduits and valves to make it part of the NCAP system. The addition of the heat exchanger 810 to make the NCAP system can be ignored temporarily. When valves 860X, 865X, 832, and 814X are closed, FIG. 8 is a conventional TPP plant above line A-A′ and an old style CAES below line A-A′. When valve 829 is open, the old-style CAES 800 uses a mixer 828 to mix natural gas with the incoming compressed air 831 before it ignites and expands in air turbine 815 to drive generator 820. When valve 829 is closed and valves 860X, 865X, 832, and 814X in FIG. 8 are open, CAES 800 is connected to TPP 100 through a heat exchanger 810 that supplies the waste heat from TPP 100 to replace the natural gas energy used in the old-style CAES configuration described above. The heated compressed air 814 enters turbine 815. The compelling fact is that a 300 MW CAES system needs external heat energy (HOE) added to its compressed air released so it can produce electrical power with an air turbine. An ordinary fossil fuel 500 MW electric steam-generating power plant (1250 MW thermal operating at 40% efficiency) exhausts at least 750 MW of waste heat (60%) into the environment. (Almost 60% the thermal power generated in the world is exhausted into the atmosphere). The full details of how TPP 100 and CAES 800 are connected to make the NCAP system in FIG. 8 are described in detail below, but the main modifications are as follows: A relatively inexpensive heat exchanger 810 is provided and is connected to receive waste heat from turbine 840 via valve 860X and return the condensate on line 865 and valve 865X. Thus, some of the formerly wasted heat generated by the TPP is transferred to heat exchanger 810 where it is used to heat the compressed air 830 coming out of tank 805 of the CAES 800. Heat exchanger 810 is a multi-stage exchanger similar in design to the heat exchangers used at all thermal power plants to carry away the waste heat generated by the TPPs. More details of exchanger 810 are provided in FIG. 9. Heat exchanger 810 simply transfers some of the TPP's waste heat to the compressed air instead of allowing it to exhaust into the atmosphere. I.e., the NCAP system simply utilizes this free waste heat to satisfy the heat energy requirement of a companion CAES, thereby eliminating the need for costly and polluting fossil fuel to heat the compressed air and converting some of that TPP waste heat into useful power. The CAES stores energy in the form of compressed air in a reservoir or tank 805. This tank does not have to be located very close to the TPP in the NCAP system. Instead, the tank can be located remotely from the TPP, even at long distances therefrom. The compressed air can be transported (piped) long distances from its reservoir 805 to the companion TPP plant just as easily as piping natural gas to the CAES facility. There is no heat loss because the stored isothermal compressed air is at ambient temperature. Likewise, the renewable energy generation can be far away from either the CAES or the TPP. The renewable energy is normally transmitted as electrical power 826 over the gird to drive the compressor 825 that charges the CAES air reservoir at some other location on the grid. Potential underground air reservoir sites exist near many TPP power plants at all latitudes around the world. It has been estimated that underground CAES air reservoirs could be located under 80% of the land area of the U.S. That means there can be CAES facilities near most of the thousands of thermal power plants. And a CAES storage facility can be many miles distant from its companion power plant. Detailed Explanation of NCAP System—FIGS. 8-13 For easy comparison and explanation, FIG. 8 shows all the apparatus in both an old-style CAES with the additions of the new NACP system. The valves marked “X” control the flow of fluids and transform the apparatus into one of the two different systems. The apparatus below line A-A′ is a CAES plant 800 with underground compressed air storage tank or reservoir 805. CAES 800 has its own compressed air turbine 815 and electrical generator 820 for returning its stored energy to the grid. Plant 800 also has a compressor 835 powered by electrical grid lines 826 for filling or charging tank 805. Tank 805 will normally be charged by off-peak power or renewable energy transmitted over gird lines 826 to power compressor 815. An old-style CAES system is formed when valve 829 is open and valves 832, 870, 814X, and 860X are closed. Under these settings, TPP 100 and CAES 800 are not connected in any way. High-pressure compressed air is stored in reservoir 805 at about 70 bar (1000 psi), but can be between about 35 bar (500 psi) and 200 bar (2000 psi). The high-pressure air enters mixer 828 where it is combined with natural gas at about the same pressure. The natural gas provides the HOE for the compressed air entering turbine 815. The combination is ignited and expands in turbine 815 to drive an electrical generator 820. This is the configuration used in the existing McIntosh and Huntorf CAES systems. When valve 829 is closed and valves 832, 860X, and 814X are open, the CAES apparatus below line A-A′ is transformed into part of the NCAP system. Some of the waste heat fluid (WHF) in line 860 from TPP 100 (the output of turbine 840) is directed into heat exchanger 810 to heat the high-pressure compressed air entering through valve 832 and circulating in conduit 812 within heat exchanger 810. The heated compressed air leaves heat exchanger 810 via conduit 814 and is then directed to air turbine 815, which drives its generator 820 to produce electrical power. Since it has been heated by the WHF, it will have the necessary HOE to expand and will not freeze the equipment or surrounding area. The power from generator 820 is returned to the gird over lines 822. Natural gas is no longer needed to heat the expanding compressed air from tank 805. In the NCAP configuration, mixer 828 in FIG. 8 is not used or does not exist. Thus, the NCAP system uses the free waste heat from TPP 100 to eliminate the great expense and pollution generated by using fossil fuel energy to provide the HOE to the compressed air in old style CAES facilities. In addition there are other major benefits, described below. At peak power demand times on most electrical grids, TPP 100 in the NCAP system of FIG. 8 will normally be running at full power output (and maximum waste heat production). This is also when the stored energy in CAES 800 is also desired. This means that maximum TPP waste heat will be available when the CAES facility needs it to return the CAES stored energy to the grid. The NCAP Heat Exchanger Gerard, supra, shows a multi-stage heat exchanger and air turbine. Gerard uses water at ambient temperature to heat the compressed air going into each turbine stage. In contrast, heat exchanger 810 in the NCAP heats the compressed air entering each stage of a multi-stage air turbine with higher temperature waste heat fluid (WHF) from a TPP. FIG. 9 shows a three-stage heat exchanger 810X feeding a three-stage (TA, TB, and TC) air turbine 815 in the NCAP system, although the turbine may have more or fewer stages. The temperature of the compressed air 814A, 814B, and 814C entering each stage of the turbine is raised by waste heat transferred to the compressed air by WHF 860 from TPP 100 flowing through heat exchangers 810A, 810B, and 810C. (WHF 860 from TPP 100 will typically be steam, hot water, hot exhaust gases, molten salt, molten metal, etc.). WHF 860 enters each heat exchanger stage 810A, 810B, and 810C and exits as cooler WHF 865 from each stage. Cooler WHF 865 returns to the boiler of plant 100. The compressed air flows through each heat exchanger stage in a sealed conduit 812 that is surrounded by hot WHF 860, thereby filling the heat exchanger's enclosure 811 inside the heat exchanger. The heated compressed air 814A, 814B, and 814C leaving each heat exchanger gains energy at each stage. The output of the final turbine stage 810D is cold air 830D because it expanded in turbine stage TC. The high pressure compressed air 830 in FIG. 8 enters the multi-stage heat exchanger as 830A. The heated compressed air 814A, 814B, and 814C from each heat exchanger stage enters its corresponding turbine stage TA, TB, and TC where it expands to produce power. Cold exhaust air 830D is either sent back to condenser 808 of TPP 100 to further cool waste heat fluid 860 before it returns to plant 100 or 830D is sent to equipment for co-generation. The flow rates of waste heat fluid 860 and compressed air 830A entering heat exchanger 810X are adjusted to provide the desired heat transfer rate for the expanding compressed air. Conduit 812 inside each heat exchanger can be replaced by a number of tubes contained within a sealed manifold inside each heat exchanger. This will cause each tube containing compressed air to be surrounded by hot WHF 860 flowing through enclosures 810A, 810B, and 810C. This configuration of multiple tubes is common in exhaust steam condensing heat exchanger 808 in FIG. 8 of a typical steam TPP 100. Combined Turbine and Generator for Power Plant and CAES FIG. 10 shows an NCAP system that combines turbine 815 and generator 820 (FIG. 8) normally used by CAES 800 with turbine 840 and generator 845 in steam plant 100 (also FIG. 8). While the stand-alone CAES facility in FIG. 8 uses its own turbine 815 and generator 820 to convert the stored energy in the compressed air to electrical power, the system of FIG. 10 eliminates separate turbine 815 and generator 820 by combining them with the turbine and generator of the TPP 100. Compressed air is released from CAES reservoir 805 (FIG. 10) to generate power for the grid to which the NCAP is connected. This air enters heat exchanger 810 through valve 832. Hot steam in turbine output line 860 from TPP 100 is sent to exchanger 810 through valve 860X. The compressed air is thus heated and leaves exchanger 810 through line 834A. When valve 834X is open, the heated compressed air flows into line 834 that enters steam turbine 840 of TPP 100 at a stage position appropriate to the pressure of the compressed air in line 834. Heated compressed air 834 entering turbine 840 will normally be in the range of 600 to 1000 psi. Compressed air 834 will normally be combined with TPP steam 802 at a similar pressure at some stage of the turbine 840. Using the standard re-heat cycle in modern fossil fuel TPP plants, the lower pressure steam coming out of the first stage in multi-stage turbine 840, plus compressed air 834 added to it, may be re-heated by exhaust gases from the boiler before reentry into a lower pressure stage of the turbine. This brings air 834 to a higher temperature that is compatible with the steam entering turbine 840. The system of FIG. 10 is most appropriate for new steam plant installations where turbine 840 can be designed to accommodate both steam from TPP 100 and compressed air from CAES 800. In an optimized turbine design the compressed air in line 834 will enter turbine 840 in several sequential “pressure stages” so that the exhaust from one stage is reheated before entering the next stage at a lower pressure (FIG. 9). This is a common design used in other applications of compressed air energy. The generic multi-stage heat exchanger 810 in FIG. 9 can be placed close to or around turbine 840 to allow reheating of multiple stages of compressed air at sequentially lower pressures. Other Important Features and Benefits of the NCAP System: Power Plant Waste Heat is Converted into Useful Power with NCAP System In addition to using free waste heat from a TPP to heat compressed air (described in FIGS. 8 to 10), the NCAP system provides another major benefit that has not been realized in any previous systems or designs for power generation by fossil fuel or nuclear power plants. In FIG. 8, most of the TPP's waste heat 860 transferred to the CAES compressed air 814 is converted into useful power when the heated compressed air 814 expands in a turbine 815 to produce electricity. With a very large CAES absorbing most of the waste heat from a TPP, the energy efficiency of the companion TPP plant can be improved substantially when the CAES is delivering its power to the grid. The only other ways that TPP low temperature waste heat has been used is with co-generation plants that use the low-temperature waste heat fluid 860 as process steam in chemical plants and/or to heat or cool buildings. The exhaust steam 860 from a fossil fuel or nuclear power plant TPP is too low in temperature to drive a turbine and/or otherwise generate mechanical work. No other prior art designs actually convert the low-temperature waste heat 860 from a fossil fuel or nuclear TPP into useful electrical power as is accomplished by the NCAP system. Cooling Water Saved When TPP 100 (FIG. 8) is a steam generating plant that drives a steam turbine, the NCAP system also reduces the amount of external “cooling water” that is needed to cool steam condensate 860 returning to plant 100 when the CAES is returning its energy to the grid. The heat exchanger 810 extracts heat from the exhaust steam 860 (this is the waste heat fluid WHF). In many TPP, the cooling of exhaust steam 860 is done with fresh water circulating in the standard heat exchanger 808. The cooling water is then cooled by evaporation in cooling water towers that dissipate large amounts of the water into the atmosphere (8). Ninety percent of the power plants in the world are steam generating TPPs. If even 20% of the waste heat from a 1000 MW TPP is taken by a companion CAES when it is operating for ten hours per day, 10,000 gallons of fresh (cooling) water can be saved each day. This amounts to 3,650,000 gallons per year. Conserving the cooling water used by power plants is an important feature. There are predictions that many power plants around the world will be limited in production in future decades because of a shortage of local cooling water. NCAP Provides Backup Emergency Power for Nuclear Plants There is another feature inherent in the NCAP system for a nuclear plant. If the Emergency Core Cooling Systems (ECCS) safety systems in a nuclear plant are functional but the plant has lost power from the grid and its emergency backup power is disabled (as happened at Fukushima), a companion CAES facility with its own turbine and generator can immediately supply the emergency electrical power needed by the ECCS systems in the companion nuclear plant. Unlike a nuclear plant turbine and generator, the power output from a CAES facility can be controlled quickly to match the power demand of the electrical pumps and equipment in a standard ECCS system. A standard ECCS system in a nuclear plant continues to circulate the hot water and steam generated by the decay heat in the reactor and cool this hot water and steam in a heat exchanger that transfers the reactor heat to the environment. Hence, this waste heat is still available for the compressed air in a CAES (in FIG. 8) that is supplying emergency power to the nuclear plant ECCS. Compressed Air from a CAES can be Piped to Remote NCAP Companion TPP Plant The CGES compressed air reservoir 805 in the NCAP system of FIG. 8 does not have to be very close to its companion TPP 100. The isothermal compressed air from the CGES reservoir 805 can be piped to a TPP 100 located remotely (about one to about 20 km) from the TPP since the compressed air can be transported to the TPP as easily as transporting natural gas the same distance. E.g., oil well operators in the Permian Basin in Texas pipe large amounts of pressurized CO2 at ambient temperature from New Mexico (over 200 miles) to be used for enhanced oil recovery (EOR). Air turbine 815 (FIG. 8) that transforms the compressed air into electricity must be located at TPP 100 so that waste heat 860 from TPP 100 can be transferred to the compressed air by heat exchanger 810. While compressor 825 is best located at compressed air reservoir 805 if power lines are available to power the compressor, it can also be located at TPP 100. In the latter case, the compressed air must be transported in both directions, to and from reservoir 805. This flexibility allows a CAES reservoir to be anywhere within hundreds of square miles around a TPP. Renewable Energy Generation can be Far from CAES Storage on Same Grid In similar fashion to the location of CAES air reservoir 805 (FIG. 8) far away from its companion TPP in the NCAP system, renewable wind and solar energy sources do not have to be close to the CAES reservoir 805 that stores their energy if they are all connected to the same electrical grid. The renewable energy can be transported over the gird to the compressor 825 that charges the air reservoir in FIG. 8. Hence, in the NCAP system the only things that must be co-located at the companion TPP plant are the heat exchanger 810 and the air turbine that converts the heated compressed air to electricity. Electrical grid transmission lines are available at all TPP plants to allow remote location of the other facilities in the NCAP system. Small TPP Power Plant is Sufficient for Large NCAP Renewable Energy Storage System The 40 to 60% waste heat from even a small steam generating or gas turbine power plant (300 to 400 MW) is more than sufficient to supply the heat of expansion needed by a large CAES facility in the NCAP system of FIG. 8. For example, a small 300 MW combined cycle gas turbine TPP operating at 60% efficiency will produce (continuously) at least 200 MW of waste heat. The common 1000 MW nuclear TPP can supply a CAES that produces at least 3000 MW, which would be an enormous CAES (so far, the largest CAES proposed is 500 MW). Comparison of NCAP with Molten Metal and Battery Energy Storage The best comparison of different energy storage systems is to compare their megawatt-hours (MWH) of energy returned to a grid. The Crescent Dunes facility in Nevada (6) is a large solar thermal generating plant that uses molten salt as a storage medium. It is advertised to store 1100 MWH of energy (3960 GJ). The existing McIntosh CAES facility in Alabama that has been operating since 1999 stores 2860 MWH (10,290 GJ). Thus the McIntosh stores two and one half times as much as the Crescent Dunes molten salt facility. Newer CAES systems on the drawing board will store four to five times as much as the McIntosh CAES. These CAES systems incorporated into the NCAP system will be able to cool a 1300 MW nuclear reactor well beyond the critical first three days after shutdown. Existing CAES facilities store hundreds of times more energy than the largest battery energy storage facility, the Battery Energy Storage System in Fairbanks, Ak. New CAES systems being built are three to five times bigger than the existing McIntosh CAES facility. It is likely that newer CAES facilities will continue to be many times bigger than the largest battery storage facilities at similar cost. NCAP System Provides ECCS when TPP is Nuclear Plant There is an added benefit of the NCAP system when the steam plant (TPP) is a nuclear plant. The compressed air in the companion CAES facility can be used to cool the hot fuel rods in a nuclear reactor during a nuclear emergency when the internal ECCS is disabled (as happened at Fukishima). This feature comes at almost no additional expense once the NCAP system is built to provide a very profitable Renewable Energy Storage facility. Opening valve 870 in FIG. 8 allows the compressed air in line 830 from CAES reservoir 805 to flow directly into RPV (or other boiler) 102 and expand around the hot fuel rods. This compressed air ECCS system is a passive ECCS. I.e., it requires no other emergency power supply or energy source other than the compressed gas stored in the CAES facility. Thus, the NCAP operating as an energy storage system also provides at no extra cost a valuable ECCS safety system for a companion nuclear power plant. The compressed air ECCS by itself might not be built because of the cost if it were not used for another very profitable purpose such as a renewable energy storage facility. The details of this design feature are shown in FIGS. 11 and 12. In a large 1300 MW electric nuclear power station with a nuclear reactor that produces 3900 MW of thermal heat energy, the decay heat from the reactor fuel rods immediately after the reactor shutdown is normally 7% of the rated 3900 MW thermal. This is equal to 273 MW immediately after shutdown; this reduces to 1.5% after one hour. This decay heat must be removed from the fuel rods. Otherwise, they will melt within a few hours. In a NCAP system using the existing McIntosh CAES, the compressed air in the CAES alone could cool a companion 1300 MW nuclear reactor for the critical first 24 hours or more after shutdown. Moreover it does this without using any cooling water. This is not a trivial consideration in light of what happened at Fukushima and could happen at any of the 420+ aging nuclear plants in the world in the future. The rescue crews at Fukushima were within a few hours of being able to restore internal cooling water circulation when the irreversible meltdown process began. A single CAES facility connected to the Fukushima reactors might have saved the day by giving them a few hours of reactor cooling. New CAES facilities now being built are three to five times bigger than the McIntosh facility. They will be able to cool a 1300 MW reactor for more than three days—until it has reached cold shutdown. For example, the energy to compress the air from 1 bar to 70 bar (1029 psi) stored in the McIntosh facility is approx. 15,600 GJ. (This is consistent with a total energy return to the grid of 110 MW for 26 hours (10,296 GJ) as advertised for the McIntosh facility.) Assume that the compressed air in the McIntosh facility will absorb 30% of its stored energy as heat if it is released to expand around the hot fuel rods in a vented nuclear reactor pressure vessel (RPV) during an emergency. (If the primary water circulation equipment has ruptured, the pressure in the RPV will be low). The decay heat absorption available is 30% of 15,600=4,680 GJ. The total decay heat from the fuel rods is 3220 GJ for the first 24 hours after reactor shutdown of a 1300 MWE nuclear reactor. Hence, in a NCAP system using the existing McIntosh CAES, the compressed air in the CAES alone can cool a companion 1300 MW nuclear reactor for the critical first 24 hours or more after shutdown. Again this is done without cooling water. These values are used for explanatory purposes and are not intended to be limiting in any way. Those skilled in the art of nuclear power plant design will understand how to scale the emergency cooling requirements described herein to fit smaller and larger nuclear power plants. Detailed Description of Baffle 905 and Manifold 910—FIGS. 11 and 12—Gas Path Inside a Reactor—FIG. 11. FIG. 11 is a schematic diagram that shows one embodiment of the compressed gas emergency core cooling system (CGECCS) inherent in the NCAP system of FIG. 8. Compressed air (CG) from storage cavern 805 is injected directly into RPV 102 during a nuclear emergency. When valve 870 is opened and valves 829 and 832 are closed, CG flows directly from reservoir 805 into the RPV. In FIG. 11, a plurality of fuel rods 900 inside the RPV are positioned within a baffle 905, a tall, cylindrical structure that surrounds rods 900. Such baffles are commonly used to shield the inner wall of the RPV from neutron radiation coming from the fuel rods and for channeling the primary heat transfer water around rods 900 during normal operation. A closed, cylindrical manifold 910 is tightly mounted outside and around a lower portion of baffle 905. Pipes 140 and 315 deliver CG to manifold 910 when valve 310 is opened during a nuclear emergency. CG leaves manifold 910 via a plurality of openings 915 into baffle 905. The CG expands in the volume surrounding rods 900. Manifold 910 is described in greater detail in FIG. 12. The CG expands as it flows out of manifold 910 and over rods 900, thereby extracting heat from the fuel rods. The expanded CG exits via pipe 925, either into the atmosphere outside PCV 104 or into PCV 104, or both. Gases within PCV 104 are vented to the atmosphere via pipe 945 when valve 950 is opened. PCV 104 can contain a pool of suppression water 920 surrounding the RPV up to a predetermined level. The RPV can also be a spent fuel rod containment vessel (not shown) in which the spent fuel rods are surrounded by water (a water bath storage) until they cool. In this case, the CG could be used in the same fashion as described herein for an RPV in case the water is lost from the fuel rod containment and there is a danger of the fuel rods catching on fire and releasing radioactive material (as nearly happened at Fukushima). In FIG. 11, valve 108 is the safety relief valve on the RPV that opens when pressure in the RPV exceeds the maximum allowable level. Relief valve 108 can be opened to vent CG and/or steam out of the RPV if necessary. When the standard ECCS systems in a SCRAMMED plant 100 are disabled, valve 108 normally will be opened to vent fluids in the RPV and allow injection of cooling fluids into it. (However, this was not done during the Fukushima accident until it was too late.) A pipe 925 conducts fluids from the RPV. Valve 930 releases fluids into the atmosphere when it is opened. Another pipe 935 branches from pipe 925. A valve 940 in pipe 935 vents fluids into PCV 104 when it is opened. When desired, valves 925 and 940 can be opened at the same time, venting fluids into PCV 104 and the atmosphere outside the RPV. An additional pipe 945 and valve 950 are provided to vent fluids from PCV 104 when valve 950 is opened. FIG. 12 is a top view of manifold 910 and baffle 905. CG enters manifold 910 via pipe 315 and flows inward via a plurality of openings 915 in baffle 905. The fuel rods are contained inside manifold 910 as shown in FIG. 11. The feasibility and utility of the NCAP embodiment is demonstrated by the fact that the stored CG in a large CAES facility, such as the one in Huntorf, Germany, can cool the fuel rods in a 3900 MW thermal reactor for at least three days with no regeneration of the CG in its storage unit 805 (FIG. 8). In addition, the CAES can begin delivering enormous quantities of CG very quickly—within minutes—to handle the immediate maximum decay heat (7% of rated thermal power) that must be removed to avoid damage to the fuel rods in a companion nuclear plant in the NCAP embodiment. It can be expected that either standby electrical generators or air compressors will be available within three days to regenerate the CG in cavern 805 (FIG. 8) so that reactor cooling can continue indefinitely. The expanding CG can bring a nuclear reactor to cold shutdown condition (fuel rods under 100° C.). Note that only the passive energy in the large CG cavern 805 is necessary to provide a security blanket of emergency cooling for nuclear plant 100 in FIG. 8. The unique nature of the NCAP is demonstrated by the fact that the nuclear power industry heretofore has not utilized this tremendously valuable safety enhancement plus the great operational cost savings available by connecting a relatively inexpensive CAES system to a nuclear plant to implement the NCAP system as shown in FIG. 8. The standard ECCS systems in existing nuclear plants that utilize internal cooling water cost far more than a companion underground CAES facility at most locations. The Fukushima nuclear meltdowns in 2011 would not have happened if the reactors had been connected to even one inexpensive CAES power plant as described for the NCAP embodiment. Fifty billion dollars or more of loss could have been prevented with the addition of the 100 million dollar CAES plant shown in FIG. 8. The “safety blanket” for nuclear power plants provided by the NCAP embodiment of FIG. 8, as described in FIGS. 11 and 12, is extremely valuable. This degree of safety has not been available for land-based nuclear reactors by any other means. The cost of a large CAES compressed air underground storage system is less than 3% of the cost of a typical nuclear power plant. The availability of a 100 million dollar or less CAES during a nuclear emergency can avoid the loss of a 10-billion-dollar nuclear plant and the tens of billions of additional cost for environmental damage and clean up, not to mention the cost of injuries and loss of public confidence in nuclear power. Further, a CAES plant connected to a nuclear plant as shown in the NCAP system of FIG. 8 pays for itself many times over as a utility-scale renewable energy storage facility that can be used every day. Cooling Fuel Rods in a Ruptured RPV The NCAP combined nuclear plant and compressed air energy storage embodiment is the only ECCS that can cool a reactor after the RPV has been damaged to the point that it cannot hold any cooling water around its fuel rods. In fact, the NCAP works best under this severe condition. A ruptured RPV allows full expansion of the CG injected around the fuel rods (and maximum decay heat extraction) without any pressure build up in the RPV. All ECCS systems that rely on cooling water, including the AP1000, cannot stop a meltdown in a nuclear power plant with a ruptured RPV or primary circuit that is damaged such that cooling water cannot be circulated. Serious corrosion was discovered in the thick steel lid of at least one older U.S. nuclear reactor RPV. That lid was very close to rupturing. Had that occurred under operating pressure, all internal ECCS systems (and operators) in that nuclear plant would have been disabled. Even a plant using the AP1000 design would have been helpless to avoid a meltdown. With hundreds of aging reactors around the world, it can be expected that more such events will happen in time. CAES systems are now being planned will use long segments of high pressure pipe on the ground to replace underground caverns for storing large quantities of CG. This CG storage can be installed easily in the open areas around most nuclear power stations. The use of a high-pressure pipe for CG storage in FIG. 8 would provide the NCAP emergency cooling for a reactor even if the CG stored is not used for peak power production, as in a full CAES peak power plant. In order to moderate climate warming, the world needs thousands of inexpensive, large capacity, non-polluting Renewable Energy Storage (RES) facilities in order to capture and utilize the enormous amount of untapped wind and solar renewable energy available. It is imperative that the aging nuclear power stations of the world be provided enhanced safety systems as soon as possible. There are only a few large CAES energy storage facilities operating today. The NCAP system described herein provides both of these necessities in the form of cost-effective, utility-scale compressed air energy storage systems. The NCAP system has the following specific advantages: The NCAP system (ironically) uses and converts the (free) waste heat from a companion thermal power plant to provide a compressed air energy storage (CAES) facility that is cost effective and emits no additional CO2. The NCAP system actually converts some of the TPP plant waste heat into useful power. No prior CAES designs can do this. The NCAP system can be built all over the world using only existing, proven infrastructure—and it can be done in a few years' time, not decades from now. The NCAP Emergency Core Cooling System (ECCS) feature using stored compressed air from a large CAES can carry out the reactor cooling operation under worst case conditions that cannot be handled by any of the existing ECCS systems approved or in nuclear plants today. The NCAP system can cool a water-cooled reactor in the most extreme case, i.e., where all primary coolant has been lost and the internal plumbing connected to the reactor has been damaged such that even new cooling water cannot be circulated through the reactor's “primary boundary” by the in-plant equipment. Also, a reactor pressure vessel (RPV) can be cracked such that it cannot hold high-pressure water or steam. All embodiments of the CGES external, add-on safety systems for nuclear power plants use stored compressed air to inject external cooling fluid directly into a RPV under relatively low-pressure conditions. The cooling fluid injected does not have to be retained long-term in the RPV under high pressure. The injected coolant absorbs decay heat as it passes through the RPV. Then it is vented directly out of the reactor to the primary containment vessel (PCV), and-or empty external water tanks, and-or the environment. All the CGES embodiments utilize small, inexpensive tanks to store the compressed air and the cooling water that is not pressurized until it is needed. All embodiments can bring a nuclear reactor to cold shutdown condition and cool it indefinitely thereafter with recharging of cooling fluids. The systems can be installed immediately from robust, reliable components that are available worldwide without the development and certification of new technology. All embodiments are compatible with and can add to the capability of existing ECCS systems if the existing systems are capable of operating to some degree after a nuclear power station has been shut down. None of the embodiments described require extensive regulatory agency approval or delay. They can be installed outside a nuclear power station with physical connection only to the existing external cooling water input port(s) and output fluid venting port(s) on all nuclear RPVs. They are the equivalent of parking fire trucks loaded with water outside a nuclear plant. Installation can be done without significant interruption of nuclear power station operation. No new or unproven technology is required to construct any of these embodiments. Various embodiments of the CGES system can be installed outside the buildings and infrastructure of a nuclear plant where they are relatively safe from destruction by acts of nature and-or terrorists that can disable the internal ECCSs in use today (as happened at Fukushima in 2011). The embodiments described all contain their own power system in the form of stored high-pressure CG that is more reliable and robust than fragile electrical generators (which have failed recently at U.S. nuclear plants) or DC battery banks. The embodiments described are all passive safety systems. All embodiments shown can bring a nuclear reactor to a cold shutdown condition and cool it indefinitely thereafter with recharging of cooling fluids. All embodiments of the CGES can be tested at any time without interruption of power generation at a nuclear plant. A large number of nuclear plants have been shut down since the Fukushima accident in 2011. Owners are now planning to start up some of these plants. Installation of one or more of the embodiments of the present system can be tested thoroughly on any idle plant before it is reactivated by injecting cooling fluid into the RPV. This would demonstrate enhanced safety to the press and public in a manner that they can understand. The NCAP system provides the inexpensive, utility-scale renewable energy storage system that world has been seeking to allow use of a great deal more wind and solar energy on existing electrical grids. It creates no new expense or pollution beyond what is already being produced by an existing companion TPP plant. Prior designs do not have the capability of converting some of the TPP plant waste heat into useful power through the medium of the compressed air and thereby improve the efficiency of the TPP plant. All prior CAES systems use some heat source that adds substantial capital cost and/or operational expense and/or emits extra pollution to the operation of the CAES. Most prior designs burn natural gas in a gas turbine of some sort to provide the heat of expansion (HOE) needed by the compressed air which adds substantial CO2 to the atmosphere. However, because of the great need for renewable energy storage, new CAES facilities are being planned that are old style because they propose using natural gas to heat the stored compressed air. Large reductions in cost and pollution generated can be accomplished if these new CAES facilities are connected to thermal power plants to implement the NCAP system. When the companion TPP plant in a NCAP system is a nuclear power plant, the compressed air in the CAES reservoir provides the means to cool the nuclear reactor during an emergency at no almost no additional cost to the facility that produces power on a continuous basis. When a TPP is a steam generating plant that drives a steam turbine, the NCAP system also reduces the amount of external “cooling water” that is needed to cool the steam condensate returning to the plant when the CAES is returning its energy to the grid. The NCAP's additional heat exchanger transfers heat from the exhaust steam or waste heat fluid to the compressed air, thereby reducing the amount of cooling water needed to cool the exhaust steam and condense it to feed water that is returned to the boiler. Thus the size of the cooling water towers and the evaporative water loss can be reduced. This is a substantial advantage since 90% of the power plants in the world are steam generating TPPs. If even 20% of the waste heat from a 1000 MW TPP is taken by a companion CAES when it is operating for ten hours per day, 45.5 kilo-liters or 10,000 gallons of fresh (cooling) water can be saved each day. This amounts to 3,650,000 gallons or 16,607.5 kilo-liters per year. There are predictions that many power plants around the world will be limited in production in future decades because of a shortage of local cooling water. The NCAP system provides a roadmap for developing countries to leap beyond building more fossil fuel power plants (right now) by greatly expanding their renewable energy generation and/or greatly improving the safety of nuclear power plants. Almost every electrical grid in the world includes one or more large TPP plants, usually a fossil fuel or nuclear steam generating plant that can be the companion CAES facility for a NCAP system. The combination of a large CAES facility connected to a nearby TPP power plant makes the NCAP system an ideal Renewable Energy Storage facility that can be used around the world—and allow the capture of ten times more wind and solar renewable energy on a continuous basis. Using some existing TPP plants to implement NCAP systems will allow grids to forego building more fossil fuel TPP plants. There is no energy storage facility in use today or in prototype testing that can match the NCAP system for a compressed air storage (CAES) as measured by 1) very large capacity, 2) inexpensive per megawatt-hour (MWH) of storage, 3) very robust and reliable, 4) does not emit CO2 to the atmosphere, 5) uses no toxic materials for energy storage that can be a danger to the environment, 6) can be built immediately using existing technology that has been thoroughly tested, and 7) appropriate for use all over the world. The total cost of building the test equipment and field testing the NCAP Renewable Energy Storage system as described above should not exceed two million dollars if the owner of a steam plant will allow the connection of a basic heat exchanger 810 as shown in FIG. 8. Compare this to the many hundreds of millions of dollars being spent every year on Renewable Energy Storage prototypes using batteries, molten metal, and chemical processes that can store only a fraction of the energy stored in the existing CAES facilities at McIntosh, Ala., and Huntorf, Germany. While the above description contains many specificities, these should not be construed as limitations on the scope, but as exemplifications of some present embodiments. Many other ramifications and variations are possible within the teachings. Thus the scope should be determined by the appended claims and their legal equivalents, and not by the examples given.
	description	This application is a continuation-in-part of International Patent Application No. PCT/CN2015/079887 with an international filing date of May 27, 2015, designating the United States, now pending, and further claims foreign priority benefits to Chinese Patent Application No. 201410265965.3 filed Jun. 13, 2014, and to Chinese Patent Application No. 201420318554.1 filed Jun. 13, 2014. The contents of all of the aforementioned applications, including any intervening amendments thereto, are incorporated herein by reference. Inquiries from the public to applicants or assignees concerning this document or the related applications should be directed to: Matthias Scholl P. C., Attn.: Dr. Matthias Scholl Esq., 245 First Street, 18th Floor, Cambridge, Mass. 02142. Field of the Invention The invention relates to a construction layout for caverns of an underground nuclear power plant. Description of the Related Art Conventional construction layouts for caverns of an underground nuclear power plant leave much to be desired. In general, the distribution of the caverns is irregular, which means parts of the terrain are heavily excavated posing hidden risks. In addition, the random distribution of the primary caverns adversely affects the construction of adits. This delays the construction time and increases the investment cost. In view of the above-described problems, it is one objective of the invention to provide an improved construction layout for caverns of an underground nuclear power plant that is highly modularized and allows for expedient construction of the power plant. To achieve the above objective, in accordance with one embodiment of the invention, there is provided a construction layout for caverns of an underground nuclear power plant. The construction layout comprising: two primary caverns accommodating nuclear reactor powerhouses, electric powerhouse caverns, safe powerhouse caverns, auxiliary powerhouse caverns, nuclear fuel powerhouse caverns, connecting powerhouse caverns, a first primary traffic tunnel, a third primary traffic tunnel, a second primary traffic tunnel, a fourth primary traffic tunnel, and a primary steam channel. Each electric powerhouse cavern, each safe powerhouse cavern, each nuclear auxiliary powerhouse cavern, each nuclear fuel powerhouse cavern, and each connecting powerhouse cavern form a circle surrounding each primary cavern. The connecting line of medial axes of the two primary caverns is perpendicular to the longitudinal direction of a mountain in which the underground nuclear power plant is constructed. The electric powerhouse caverns, the safe powerhouse caverns, and the nuclear fuel powerhouse caverns are arranged along the longitudinal direction of the mountain. The auxiliary powerhouse caverns and the connecting powerhouse caverns are arranged perpendicular to the longitudinal direction of the mountain. Each of the safe powerhouse caverns and each of the nuclear fuel powerhouse caverns are disposed on two sides of each of the two primary caverns in the longitudinal direction of the mountain, respectively. Each of the electric powerhouse caverns and each of the safe powerhouse caverns are located on a same side of each the two primary caverns. Each of the auxiliary powerhouse caverns and each of the connecting powerhouse cavern are disposed on two sides of each of the two primary caverns perpendicular to the longitudinal direction of the mountain. The first primary traffic tunnel and the third primary traffic tunnel are disposed on two sides of two primary caverns in the longitudinal direction of the mountain. The second primary traffic tunnel is disposed between the two primary caverns. The first primary traffic tunnel communicates with the third primary traffic tunnel via the fourth primary traffic tunnel. One end of the second primary traffic tunnel communicates with the fourth primary traffic tunnel. The other ends of the first primary traffic tunnel, the second primary traffic tunnel, and the third primary traffic tunnel communicates with a ground surface. Each of the two primary caverns communicates with a corresponding connecting powerhouse cavern via the primary steam channel communicating with the ground surface. A skewback or an endwall of a top arch of each cavern of the nuclear island powerhouse cavern group communicates with the ground surface via a top adit system functioned in construction. A bottom of a sidewall of each cavern of the nuclear island powerhouse cavern group communicates with the first primary traffic tunnel, the second primary traffic tunnel, the third primary traffic tunnel, the fourth primary traffic tunnel, and the primary steam channel via a bottom adit system functioned as a slag discharging channel. In a class of the embodiment, the top adit system comprises: a first primary adit a first top adit of a first primary cavern, a second top adit of a second primary cavern, a second primary adit a third primary adit a fourth primary adit third top adits of the nuclear fuel powerhouse caverns, fourth top adits of the connecting powerhouse caverns, fifth top adits of the electric powerhouse caverns, seventh top adits of the safe powerhouse caverns, an eighth top adit of the auxiliary powerhouse caverns, and ninth top adits of the safe powerhouse caverns. An elevation of the first primary adit is higher than an elevation of the second primary adit an elevation of the third primary adit and an elevation of the fourth primary adit. The first primary adit is connected to a skewback of a top arche of the first primary cavern and a skewback of a top arche of the second primary cavern via the first top adit and the second top adit respectively. The second primary adit and the fourth primary adit are respectively connected to endwalls of top arches of outer end faces of the two auxiliary powerhouse caverns. Two ends of the eighth top adit are respectively connected to endwalls of top arches of inner end faces of the two nuclear auxiliary powerhouse cavern. One end of the third primary adit is connected to a middle section of the eighth top adit. The second primary adit is connected to an endwall of a top arch of a first nuclear fuel powerhouse cavern via one of the third top adits. The third primary adit is connected to an endwall of a top arch of a second nuclear fuel powerhouse cavern via the other of the third top adits. The third primary adit is connected to an endwall of a top arch of a first connecting powerhouse cavern via one of the fourth top adits. The fourth primary adit is connected to an endwall of a top arch of a second connecting powerhouse cavern via the other of the fourth top adits. The fourth top adits are connected to endwalls of top arches of the electric powerhouse caverns via the fifth top adits respectively. The third primary adit is connected to an endwall of a top arch of one end of a first safe powerhouse cavern via one of the seventh top adits. The eighth top adit is branched to form one of the ninth top adits connected to an endwall of a top arche of the other end of the first safe powerhouse cavern. One of the ninth top adits is disposed between an inner end face of a first nuclear auxiliary powerhouse cavern and the third primary adit. The fourth primary adit is branched to form the other of the seventh top adits. The other of the seventh top adits and the other of the ninth top adits are connected to endwalls of top arches of two ends of a second safe powerhouse cavern. In a class of the embodiment, each top adit of the top adit system has a longitudinal slope smaller than 12%. In a class of the embodiment, the bottom adit system comprises: first bottom adits of the two primary caverns, second bottom adits of the electric powerhouse caverns, third bottom adits of the safe powerhouse caverns, fourth bottom adits of the auxiliary powerhouse caverns, fifth bottom adits of the nuclear fuel powerhouse caverns, and sixth bottom adits of the auxiliary powerhouse caverns. The first primary traffic tunnel is connected to a bottom of a sidewall of a first primary cavern via one of the first bottom adits. The second primary traffic tunnel is connected to a bottom of a sidewall of a second primary cavern via the other of the first bottom adits. Each primary steam channel is connected to a bottom of a sidewall of corresponding electric powerhouse cavern via each second bottom adit. The two third bottom adits are disposed on the fourth primary traffic tunnel and are connected to bottoms of endwalls of corresponding safe powerhouse caverns. The two fifth bottom adit are disposed on the fourth primary traffic tunnel and are connected to bottoms of endwalls of corresponding nuclear fuel powerhouse caverns. The two fifth bottom adit are connected to bottoms of endwalls of one ends of corresponding auxiliary powerhouse caverns via sixth bottom adits of the auxiliary powerhouse caverns; The two third bottom adit are connected to bottoms of endwalls of the other ends of corresponding auxiliary powerhouse caverns via fourth bottom adits of the auxiliary powerhouse caverns. In a class of the embodiment, each bottom adit of the bottom adit system has a longitudinal slope smaller than 15%. In a class of the embodiment, the two primary caverns are provided with first apparatus conveying channels for communicating with a corresponding connecting powerhouse caverns. A bottom of a sidewall of each of the connecting powerhouse caverns communicates with the ground surface via a second apparatus conveying channel. Advantages of the construction layout of the circular-shaped nuclear island cavern group of the underground nuclear power plant according to embodiments of the invention are summarized as follows: The primary traffic tunnels having the elevation of A m are fully utilized to construct bottom adits of the powerhouse caverns as the slag discharging channels and to construct top adits of the powerhouse caverns as the construction channels communicating with the ground surface, so that the construction of the powerhouse caverns can be organized using the fewest construction channels according to the ore pass method (the ore pass method refers to arrange a slage discharging pit in a vertical direction between an upper channel and a lower channel in an excavation region, use the upper channel as the construction channel, and introduce slags produced in the construction to the lower channel via the slag discharging pit so as to carry the slag out of the construction site). Each powerhouse cavern is only provided with the bottom adits and the top adits. The top adit system communicates with the top arches of the powerhouses and the bottom adit system communicates with the bottom of the walls of the powerhouses. A bottom layer of each connecting powerhouse cavern adopts the primary steam channel as the construction channel, and the top layer of the connecting powerhouse cavern is configured with the fourth top adit. Each nuclear auxiliary powerhouse cavern is configured with the fourth bottom adit and sixth bottom adit at the bottom layer and is configured with the second primary adit, the fourth primary adit, and the eighth top adit at the top layer. The safe powerhouse cavern is configured with the third bottom adit at the bottom layer and is configured with the seventh top adit and the ninth top adit at the top layer. The nuclear fuel powerhouse cavern is configured with the fifth bottom adit at the bottom layer and is configured with third top adit at the top layer. The electric powerhouse cavern is configured with a second bottom adit at the bottom layer and is configured with fifth top adit at the top layer. Bottom adits of the primary cavern, the electric powerhouse cavern, the safe powerhouse cavern, the connecting powerhouse cavern, the nuclear fuel powerhouse cavern, and the nuclear auxiliary powerhouse cavern communicate with the first primary traffic tunnel, the second primary traffic tunnel, the third primary traffic tunnel, the fourth primary traffic tunnel, and the primary steam channel. The safe powerhouse cavern, the nuclear fuel powerhouse cavern, the connecting powerhouse cavern, the nuclear auxiliary powerhouse cavern, and the electric powerhouse cavern are configured with top adits at the top layers for communicating with the second primary adit, the third primary adit, and the fourth primary adit. As possessing the highest top arches, the two primary caverns are unable to share common construction channels with other caverns either on the vertical face or the horizontal face, thus, a separate first primary adit is arranged. The first primary adit is bifurcated to form two top adits of the two primary caverns respectively connected to the skewbacks of the two primary caverns. The construction layout of the invention has fewest construction channels arranged according to the ore pass method. The construction layout of the invention is short, safe, economic, highly modularized, and convenient in construction, which satisfies the general requirement of the construction. For further illustrating the invention, experiments detailing a construction layout for caverns of an underground nuclear power plant are described below. It should be noted that the following examples are intended to describe and not to limit the invention. As shown in FIGS. 1-3, a construction layout for caverns of an underground nuclear power plant comprises: two primary caverns accommodating nuclear reactor powerhouses 1, electric powerhouse caverns 2, safe powerhouse caverns 3, nuclear fuel powerhouse caverns 4, connecting powerhouse caverns 5, auxiliary powerhouse caverns 6, fourth primary traffic tunnel 7, first primary traffic tunnel 8, second primary traffic tunnel 9, third primary traffic tunnel 10, a top adit system, a bottom adit system, a primary steam channel 11, a second device conveying channel 12, and a first device conveying channel 13. Two primary caverns accommodating nuclear reactor powerhouses 1 are provided. A connecting line of medial axes of the two primary caverns is perpendicular to a longitudinal direction of a mountain. The electric powerhouse cavern 2, the safe powerhouse cavern 3, the nuclear auxiliary powerhouse cavern 6, the nuclear fuel powerhouse cavern 4, and the connecting powerhouse cavern 5 are arranged in a circle surrounding each primary cavern 1. Electric powerhouse caverns 2, safe powerhouse caverns 3, and nuclear fuel powerhouse caverns 4 are arranged along the longitudinal direction of the mountain. auxiliary powerhouse caverns 6 and connecting powerhouse caverns 5 are arranged perpendicular to the longitudinal direction of the mountain. Each of the safe powerhouse caverns 3 and each of the nuclear fuel powerhouse caverns 4 are disposed on two sides of each of the two primary caverns 1 in the longitudinal direction of the mountain, respectively. Each of the electric powerhouse caverns 2 and each of the safe powerhouse caverns 3 are located on a same side of each the two primary caverns 1. Each of the auxiliary powerhouse caverns 6 and each of the connecting powerhouse cavern 5 are disposed on two sides of each of the two primary caverns 1 perpendicular to the longitudinal direction of the mountain. A nuclear island powerhouse cavern group is formed by each primary cavern 1 and corresponding electric powerhouse cavern 2, safe powerhouse cavern 3, nuclear auxiliary powerhouse cavern 6, nuclear fuel powerhouse cavern 4, and connecting powerhouse cavern 5. A first primary traffic tunnel 8 and a third primary traffic tunnel 10 are disposed on two sides of two primary caverns in the longitudinal direction of the mountain. A second primary traffic tunnel 9 is disposed between the two primary caverns. The first primary traffic tunnel 8 communicates with the third primary traffic tunnel 10 via a fourth primary traffic tunnel 7. One end of the second primary traffic tunnel 9 communicates with the fourth primary traffic tunnel 7. The other ends of the first primary traffic tunnel 8, the second primary traffic tunnel 9, and the third primary traffic tunnel 10 communicates with a ground surface. Each of the two primary caverns 1 communicates with a corresponding connecting powerhouse cavern 5 via a primary steam channel 11 communicating with the ground surface. A skewback or an endwall of a top arch of each cavern of the nuclear island powerhouse cavern group communicates with the ground surface via a top adit system functioned in construction. A bottom of a sidewall of each cavern of the nuclear island powerhouse cavern group communicates with the first primary traffic tunnel 8, the second primary traffic tunnel 9, the third primary traffic tunnel 10, the fourth primary traffic tunnel 7, and the primary steam channel 11 via a bottom adit system functioned as a slag discharging channel. The top adit system comprises: a first primary adit 41, a first top adit 42 of a first primary cavern, a second top adit 43 of a second primary cavern, a second primary adit 31, a third primary adit 32, a fourth primary adit 33, third top adits 34 of the nuclear fuel powerhouse caverns, fourth top adits 35 of the connecting powerhouse caverns, fifth top adits 36 of the electric powerhouse caverns, seventh top adits 37 of the safe powerhouse caverns, an eighth top adit 38 of the auxiliary powerhouse caverns, and ninth top adits 39 of the safe powerhouse caverns. An elevation of the first primary adit 41 is higher than an elevation of the second primary adit 31, an elevation of the third primary adit 32, and an elevation of the fourth primary adit 33. The first primary adit 41 is connected to a skewback of a top arche of the first primary cavern 1 and a skewback of a top arche of the second primary cavern 1 via the first top adit 42 and the second top adit 43, respectively. The second primary adit 31 and the fourth primary adit 33 are respectively connected to endwalls of top arches of outer end faces of the two auxiliary powerhouse caverns 6. Two ends of the eighth top adit 38 are respectively connected to endwalls of top arches of inner end faces of the two nuclear auxiliary powerhouse cavern 6. One end of the third primary adit 32 is connected to a middle section of the eighth top adit 38. The second primary adit 31 is connected to an endwall of a top arch of a first nuclear fuel powerhouse cavern 4 via one of the third top adits 34. The third primary adit 32 is connected to an endwall of a top arch of a second nuclear fuel powerhouse cavern 4 via the other of the third top adits 34. The third primary adit 32 is connected to an endwall of a top arch of a first connecting powerhouse cavern 5 via one of the fourth top adits 35. The fourth primary adit 33 is connected to an endwall of a top arch of a second connecting powerhouse cavern 5 via the other of the fourth top adits 35. The fourth top adits 35 of the connecting powerhouse caverns are connected to endwalls of top arches of the electric powerhouse cavern 2 via the fifth top adits 36, respectively. The third primary adit 32 is connected to an endwall of a top arch of one end of a first safe powerhouse cavern 3 via one of the seventh top adits 37. The eighth top adit 38 is branched to form one of the ninth top adits 39 connected to an endwall of a top arche of the other end of the first safe powerhouse cavern 3. One of the ninth top adits 39 is disposed between an inner end face of a first nuclear auxiliary powerhouse cavern 6 and the third primary adit 32. The fourth primary adit 33 is branched to form the other of the seventh top adits 37. The other of the seventh top adits 37 and the other of the ninth top adits 39 are connected to endwalls of top arches of two ends of a second safe powerhouse cavern 3. Each top adit of the top adit system has a longitudinal slope smaller than 12% to satisfy the requirement of trackless transportation. The bottom adit system comprises: first bottom adits 21 of the two primary caverns, second bottom adits 22 of the electric powerhouse caverns, third bottom adits 23 of the safe powerhouse caverns, fourth bottom adits 24 of the auxiliary powerhouse caverns, fifth bottom adits 25 of the nuclear fuel powerhouse caverns, and sixth bottom adits 26 of the auxiliary powerhouse caverns. The first primary traffic tunnel 8 is connected to a bottom of a sidewall of a first primary cavern 1 via one of the first bottom adits 21. The second primary traffic tunnel 9 is connected to a bottom of a sidewall of a second primary cavern 1 via the other of the first bottom adits 21. Each primary steam channel 11 is connected to a bottom of a sidewall of corresponding electric powerhouse cavern 2 via each second bottom adit 22. The two third bottom adits 23 are disposed on the fourth primary traffic tunnel 7 and are connected to bottoms of endwalls of corresponding safe powerhouse caverns 3. The two fifth bottom adit 25 are disposed on the fourth primary traffic tunnel 7 and are connected to bottoms of endwalls of corresponding nuclear fuel powerhouse caverns 4. The two fifth bottom adit 25 are connected to bottoms of endwalls of one ends of corresponding auxiliary powerhouse caverns 6 via sixth bottom adits 26 of the auxiliary powerhouse caverns. The two third bottom adit 23 are connected to bottoms of endwalls of the other ends of corresponding auxiliary powerhouse caverns 6 via fourth bottom adits 24 of the auxiliary powerhouse caverns. Each bottom adit of the bottom adit system has a longitudinal slope smaller than 15% to satisfy the requirement of trackless transportation. The two primary caverns 1 are provided with first apparatus conveying channels 13 for communicating with a corresponding connecting powerhouse caverns 5. A bottom of a sidewall of each of the connecting powerhouse caverns 5 communicates with the ground surface via a second apparatus conveying channel 12. The underground nuclear power plant of this embodiment adopts the CUP 600 arrangement of the double reactors. The ore pass method is adopted to organize the excavation of the underground powerhouse caverns, that is, the excavation is organized on an upper working face, and slags are discharged from a bottom working face via a slag discharging pit. The ore pass method is able to satisfy the construction requirement by fewest construction channels. Thus, it only requires designing the bottom adit system and the top adit system to communicate with the powerhouse caverns. The bottom adits are in bifurcate arrangement based on the first primary traffic tunnel 8, the second primary traffic tunnel 9, and the third primary traffic tunnel 10, and the top adits are in bifurcate arrangement based on the first primary adit 41, the second primary adit 31, the third primary adit 32, and the fourth primary adit 33, which is beneficial for reducing the project quantity of the project construction. The bottom adit system and the top adit system keep a certain distance away from the adjacent caverns for ensuring the stability and safety of the cavern group. The excavated adits are few, the excavated distance is short, and the spaces between the adits and the caverns are proper, so that the cost is saved and the whole project is safe. Because the powerhouse caverns in the CUP600 arrangement of the double reactors adopt modularized design, the modularization design is realized in the structure of the invention, which is specifically as follows. Because the first primary traffic tunnel 8, the second primary traffic tunnel 9, and the third primary traffic tunnel 10 used as the permanent channels and the primary steam channel 11 optionally used as the channel are all arranged at the elevation of ±0 m, which is greatly different from bottom elevations of the powerhouse caverns except the connecting powerhouse cavern: for example, the bottom elevation of the primary cavern 1 is −9.0 m, the bottom elevation of the electric powerhouse cavern 2 is −7.55 m, and the bottom elevation of the safe powerhouse cavern 3, the nuclear auxiliary powerhouse cavern 6, and the nuclear fuel powerhouse cavern 4 are −12.5 m. Thus, in addition to that the connecting powerhouse cavern 5 is able to directly use the primary steam channel 11 as the construction channel, the first primary traffic tunnel 8, the second primary traffic tunnel 9, the third primary traffic tunnel 10, the fourth primary traffic tunnel 7, and the primary steam channel 11 used as the permanent channels cannot be directly utilized in either a horizontal face or a vertical face. In order to satisfy the requirement for discharging the excavated slag of the cavern bottoms, it is necessary to arrange the bottom adits to the cavern bottoms. In the meanwhile, in order to reduce the excavation rate of the mountain, to reduce the negative factors affecting the stability of the caverns, and to decrease the cost on the bottom adits, no primary bottom adit is individually designed, while the first primary traffic tunnel 8, the second primary traffic tunnel 9, the third primary traffic tunnel 10, the fourth primary traffic tunnel 7, and the primary steam channel 11 used as the permanent channels are fully utilized and bifurcated to form the bottom adits 21, 22, 23, 24, 25, 26, and 27 to reach the bottom elevation of the caverns. The slops of the bottom pits are all smaller than 15% to satisfy the requirement of trackless transportation. Because the top elevations of the powerhouse caverns are relatively high and vary from one another. For example, the top elevation of the primary cavern 1 is +78.0 m, the top elevation of the electric powerhouse cavern 2 is +46.45 m, the top elevation of the safe powerhouse cavern 3 is +41.5 m, the top elevation of the nuclear auxiliary powerhouse cavern 6 is +47.5 m, and the top elevation of the nuclear fuel powerhouse cavern 4 is +54.5 m, and the top elevation of the connecting powerhouse cavern 5 is +46.0 m, thus the first primary traffic tunnel 8, the second primary traffic tunnel 9, the third primary traffic tunnel 10, the fourth primary traffic tunnel 7, and the primary steam channel 11 functioned as the permanent channels cannot be directly utilized in either a horizontal face or a vertical face. In addition, the elevation of the apparatus conveying channel 12 is +20.0 m, which is greatly different from the elevation of other caverns, thus, it is necessary to design top adits to connect to the skewback or the endwall of the top arch of each cavern. As the top arch of the primary cavern 1 is highest and is unable to share a common construction channel with other caverns in either the vertical face or the horizontal face, the first primary adit 41 is separately designed and bifurcated to form the first top adit 42 and the second top adit 43 to connect to the skewbacks of the top arches of corresponding two primary caverns accommodating nuclear reactor powerhouses 1 having an elevation of +65.0 m. As the electric powerhouse cavern 2, the safe powerhouse cavern 3, the nuclear auxiliary powerhouse cavern 6, the nuclear fuel powerhouse cavern 4, and the connecting powerhouse cavern 5 have small height differences in the vertical face, and these caverns share the second primary adit 31, the third primary adit 32, and the fourth primary adit 33. The second primary adit 31, the third primary adit 32, and the fourth primary adit 33 are bifurcated to form the third top adits 34 of the nuclear fuel powerhouse caverns, the fourth top adits 35 of the connecting powerhouse caverns, the fifth top adits 36 of the electric powerhouse caverns, the seventh top adits 37 of the safe powerhouse caverns, the eighth top adit 38 of the auxiliary powerhouse caverns, and the ninth top adits 39 of the safe powerhouse caverns to communicate with the caverns. The second primary adit 31, the third primary adit 32, and the fourth primary adit 33 are bifurcated to form the top adits 34, 35, 36, 37, 38, and 39, which are further connected to the elevation of the skewbacks of the caverns. Elevations of the top arches of the third top adits 34, the fourth top adits 35, the fifth top adits 36, the seventh top adits 37, the eighth top adit 38, and the ninth top adits 39 are +45.0 m, +37.0 m, +33.0 m, +38.0 m, and +33.0 m, respectively. The fourth top adits 35 is bifurcated to form the fifth top adit 36 to connect to the endwall of the top arch of the electric powerhouse cavern 2 having the elevation of +37.0 m. As the size of the nuclear auxiliary powerhouse cavern 6 is relatively large, the second primary audit 31, the fourth primary adit 33, and the eighth top adit 38 are particularly arranged as the top adits. And the fourth bottom adit 24 and the sixth bottom adit 26 are adopted as the bottom adits. Because the axis of the safe powerhouse cavern 3 is relatively long, the seventh top adit 37 and the ninth top adit 39 are particularly arranged as the top adits so as to enhance the working face, balance the construction period, and facilitate the supporting of the skewbacks therefore ensure the stability and safety during the construction period of the skewbacks. The slops of the top adits are all smaller than 12% to satisfy the requirement of trackless transportation. While particular embodiments of the invention have been shown and described, it will be obvious to those skilled in the art that changes and modifications may be made without departing from the invention in its broader aspects, and therefore, the aim in the appended claims is to cover all such changes and modifications as fall within the true spirit and scope of the invention.
045432333	description	DETAILED DESCRIPTION The overall function of load pads is illustrated by FIGS. 1 and 2. In FIG. 1, a fuel assembly 1 is shown having a duct 2, this example being a six sided fuel assembly 1 and duct 2. In a section of the fuel duct 2, generally above the active fuel region of fuel assembly 1, one or more load pads 3 are installed. Each face of duct 2 has at least one pad 3, unless the fuel assembly happens to be located in a core position, especially the core perimeter, at which position external faces may not bear on adjacent assemblies and load pads 3 may not be required. FIG. 2 illustrates how load pads 3 bear on one another. A beveled surface 4 (see FIG. 3) may be provided to facilitate movement of fuel assembly 1 up and down during installation and removal from the reactor. Beveled surface 4 eliminates or minimizes edge 5 upon which obstructing interference can occur. Load pad 3 has a small groove 6 machined around the circumference of a stem 7 (refer to FIGS. 3 and 4). A spirally wound retaining ring 8 is assembled into machined groove 6. A plurality, preferably four oval slots 9 are machined through the flat pad portion; these slots 9 just nick the circumference of stem 7. Duct wall 10 has a hole 12 machined nearly through it; the hole 12 diameter is slightly larger than the diameter of load pad stem 7. A groove 11 is machined into the wall of hole 12 which will be aligned with groove 6 in stem 7. Also milled into duct wall 10 are oval slots 13 to match those slots 9 milled into the pad. Load pad 3 is installed into duct wall 10 by means of a special tool (see FIGS. 5, 6 and 13). This tool has fingers 14 which drop through slots 9 in the load pad 3 and around ring 8 in its free state. By utilizing a cam action, fingers 14 are then moved toward the center of pad 3, compressing ring 8 to a diameter slightly less than that of stem 7. Load pad 3 is then positioned into hole 12 of duct wall 10 with the tool holding ring 8 compressed. The cam action of the tool is then reversed allowing ring 8 to expand into groove 11 of duct wall 10. The tool is then removed. The spirally wound retaining ring 8 now engages both groove 11 in duct wall 10 and groove 6 in load pad stem 7, locking the two together. When installed, the forces which are applied between ducts 2 in the reactor are distributed over duct wall 10 by the broad surface of pad 3. Retaining ring 8 only has to prevent load pad 3 from falling away from duct 2. There are no appreciable forces in the direction and ring 8 has the capability of withstanding over 1000 pounds force. There are no conceivable forces that may break ring 8 into pieces, however, even if this were possible, a minimum of three breaks would be necessary before load pad 3 could be disengaged from duct wall 10, thus an inherent redundancy is provided. A feature of this design is that it allows inspection of ring 8 engagement through holes 9. If a problem exists, load pad 3 can be removed through the use of the tool and the problem corrected. The presence of hole 12 in duct wall 10 tends to weaken the duct, especially if load pads 3 on each face of the duct are located at a common elevation. To minimize this weakening effect, load pads on different faces of the duct may be located at different elevations. This requires a potentially complicated core loading plan because the load pads on adjacent faces must abut, and must therefore be at the same elevation. Slots 13 in duct wall 10 are shown in the drawings to extend only partially through duct wall 10 such that no coolant flow through these holes occurs. If load pad 3 is located above the active fuel region of the fuel assembly, it may be desired to machine slots 13 completely through duct wall 10 thus allowing coolant flow to pass outside the duct via slots 13 and 9. See FIG. 14. Once installed, load pad 3 as shown in the drawings is free to rotate, which will cause a loss of alignment between slots 9 and 13. Such rotation can be prevented by a variety of methods, including a pin and hole arrangement between duct wall 10 and load pad 3. FIGS. 7 through 10 are design drawings bearing dimensions relative to a preferred embodiment. The geometric shape of load pad 3 need not be circular but could obviously be of many shapes. The geometry of pad stem 7 and hole 12 also need not be circular, although a circular geometry here is appropriate to the use of ring 8. Substitution for ring 8 by other mechanical attachment means makes other geometries feasible.
	description	This application claims priority from provisional patent application U.S. 62/787,636, filed Jan. 2, 2019, the entire content of which is incorporated herein by reference. The present invention relates to rigid structures and composite materials thereof providing radiation attenuation/shielding. Some embodiments of the invention pertain to a radiation shielding apparatus comprising: a plurality of positionable radiation-shielding stacks of tiles, wherein the stacks are subsequently and adjacently arranged in a contiguous configuration of stacks; and a tile positioning mechanism configured to allow movement of tiles within a stack between a stacked or retracted position and an extended position, wherein in the extended position, the tiles of each of the plurality of radiation shielding stacks partially overlap tiles of subsequent and adjacent tile stack at corresponding opposing side-margins thereof. X-ray equipment is routinely used in various applications and systems, including as a diagnostic tool in medical settings. As a result, health care providers and technical personnel who operate X-ray systems may be exposed to cumulative dosage of radiation and may be harmed by such X-ray exposure. Thus, in the field and art of medical imaging, there is an on-going need for improved equipment design, materials and methodologies for preventing or at least minimizing such cumulative radiation exposure, to reduce health risks. X-ray shielding equipment is part of this effort to reduce exposure from stray radiation to below specified levels. Exemplary teachings in the field and art of the invention are provided by the applicant of the present invention in the following disclosures: U.S. Pat. Nos. 8,439,564 and 8,113,713, and WO 2017/083437, which are incorporated by reference as if fully set forth herein. Additional disclosures in this field include: U.S. Pat. Nos. 6,325,538; 8,460,777; 7,897,949; 5,525,408; 5,099,134; US 2003/174802; US 2017/278585; JP 6391149; CN 2059596270; and CN 103045983, which are incorporated by reference as if fully set forth herein. Exemplary radiation shielding apparatuses are described in the following disclosures: US patent application Nos. 2018/0168525, and 2018/0249972, and are incorporated by reference as if fully set forth herein. Among the challenges associated with radiation shielding equipment is the requirement to maintain as complete a shielding as possible, preferably using materials that are low weight yet rigid and have sufficient radiation shielding/blocking properties. The present invention relates to a radiation shielding apparatus including adjacent stacks of radiation shielding tiles that can be extended for radiation shielding of an area outside the apparatus or at least mitigate the exposure to scattered radiation. Such shielding is intended to limit/reduce radiation exposure to personnel and technicians who work with and near X-ray radiation systems (e.g., a C-arm of a fluoroscopy system). The present invention provides an apparatus having stacks of tiles with a tile positioning mechanism that allows moving the tiles between a retracted stacked position and an extended position. The tiles have a unique structure which provides compact contracted arrangement when the tiles are in the retracted position and minimizes leakage of radiation when the tiles are in the extended position. The tiles include side margins with a unique structure for mitigating radiation leakage. As such, the apparatus may include at least one radiation shield assembly including a support base operatively connectable to a radiation source or a radiation detector of an X-ray system; a plurality of positionable radiation-shielding stacks of tiles, wherein the stacks are subsequently and adjacently arranged in a contiguous configuration of stacks; and a tile positioning mechanism configured to allow movement of tiles within a stack between a stacked or retracted position and an extended position, wherein in the extended position, the tiles of each of the plurality of radiation shielding stacks at least partially overlap tiles of subsequent and adjacent tile stack at corresponding opposing side-margins thereof. The tiles may include a composite radiation shielding material. This composite material allows the apparatus to be low weight, yet rigid, and still provide for radiation shielding. Structures including the herein disclosed composite materials may be configured in various combinations of material components; various layers, and/or combination of layers, and/or permutations of layers; and as flat, or non-flat, depending upon implementations thereof. A particular example of a non-flat configuration is at one or both side-margins of the tiles, which can be particularly useful for mitigating radiation leakage. Configurations of such side-margins may include a V-shaped portion, a wavy configuration and/or a zig-zag pattern, or a combination thereof. Such configurations (e.g. wavy and zig-zag) allow stable and overlap of increased surface area of the side-margins of adjacent tiles, without using/requiring additional linear space; and provide a more tortuous path for radiation to potentially leak thereby reducing the chances and/or amount of radiation leakage. The edges (peak of ridges) of the V-shape or zig-zags (and crests/valleys of the waves) define an axis A1 parallel to the extension direction of the tile stacks. Namely, edges (peak of ridges) are parallel to the direction of movement of the tile stacks. Advantages of the present invention may include (a) a reduction in radiation exposure (i.e. providing a more comprehensive a radiation shield), which may in particular include reducing radiation leakage at the corners of the radiation shield; (b) improvement in the overlap of adjacent radiation shielding tiles or tile-stack segments (i.e. stacks of tiles) to thereby mitigate radiation leakage; and (c) provide for improved strength and/or stability of the radiation shield. Examples of improved configurations or patterns of such overlap are noted above, namely wavy; V-shape, and zig-zag. Again, regardless of the particular shape of the tile's edge-margins, or whether the tiles form a face of the shielding structure or include corners thereof, the edges (peak of ridges) of the V-shapes or zig-zags and/or crests/valleys of the waves, at the overlapping side-margins, define an axis A1 parallel to the extension direction of the tile stacks. Thus, an aspect of the invention pertains to a radiation shielding apparatus comprising: a plurality of positionable radiation-shielding stacks of tiles, wherein the stacks are subsequently and adjacently arranged in a contiguous configuration of stacks; and a tile positioning mechanism configured to allow movement of tiles within a stack between a stacked or retracted position and an extended position, wherein in the extended position the tiles of each of the plurality of radiation shielding stacks at least partially overlap tiles of subsequent and adjacent tile stack at corresponding opposing side-margins thereof. In one or more embodiments, the tiles and corresponding opposing side-margins are non-flat. In one or more embodiments, the non-flat corresponding opposing side-margins have a zig-zag or V-shaped profile. In one or more embodiments, the non-flat corresponding opposing side-margins have a wavy or S-shaped profile. In one or more embodiments, the stacks of tiles form a structure having at least two faces, each face including at least one tile stack; and corner tile stacks connecting two adjacent faces thereof. In one or more embodiments, the stacks of tiles form a structure having at least three faces, each face including at least one tile stack; and corner tile stacks connecting two adjacent faces thereof. In one or more embodiments, the stacks of tiles form a structure having four faces, each face including at least one tile stack; and four corner tile stacks connecting two adjacent faces thereof. In one or more embodiments, corner tile stacks cover an area of at least about 90° angle between two adjacent faces. In one or more embodiments, the tile positioning mechanism includes a rail and a slide element for allowing sliding of slide element of one tile along a length of the rail of an adjacent (upper or lower) tile within a stack. In one or more embodiments, the rails and slide elements within a stack arranged in a graded structural form, thereby providing a compact structure of tiles in a stack. In one or more embodiments, the rails and slide elements within a stack arranged in a nesting structural form, thereby providing a compact structure of tiles in a stack. In one or more embodiments, tiles within a stack include a recess to accommodate therein a rail of the tile and a respective slide element of a sequentially adjacent tile. In one or more embodiments, the recesses of stackedly adjacent tiles of the stack are arranged such that the recess of one tile is correspondingly disposed relative the recess of its sequentially adjacent tile, such that one recess of one tile accommodates at least partially a second recess of a second adjacent tile, thereby providing for a compact structure of tiles in a stack. In one or more embodiments, each tile comprises a first side margin with a concave or V-shaped profile and an opposite second side margin with a convex or upside down V-shaped profile, and the tiles of subsequent and adjacent tile stacks are arranged such that the concave or V-shaped profile of the tiles within one stack overlap the convex upside down V-shaped profile of the tiles within the subsequent and adjacent tile stack. The materials and structures of the tiles may include one or more layers of carbon fiber and a binding material, and one or more layers of a radiation attenuation material. In some designs, the herein disclosed tiles include one or more layers of carbon fibers incorporated within a mixture of a binding material and one or more radiation attenuation material. In some designs, the herein disclosed tiles include one or more layers of radiation-attenuating material and a polymer mixture. The structures obtained from the herein disclosed materials are rigid, low-weight, and can be flat or non-flat and possess radiation shielding properties. An aspect of the invention pertains to rigid/semi-rigid structures comprising a radiation attenuating composite material, the composite material comprising a mixture of one or more polymers and one or more radiation attenuating material(s) wherein the obtained structure is a monolayered structure. In one or more embodiments, the radiation attenuating material(s) is provided as a powder which is substantially homogenously dispersed in the one or more polymers. A further aspect of the invention pertains to a radiation attenuating composite material, the composite material comprising: one or more carbon fiber layers; a binding material; and a radiation attenuating material applied onto and/or between the one or more carbon fiber layers. Yet a further aspect of the invention pertains to a radiation attenuating composite material, comprising: one or more layers of carbon fibers and a binding material; and a radiation attenuating material applied onto and/or between the one or more of carbon fiber layers. Yet a further aspect of the invention pertains to a radiation attenuating composite material, comprising: one or more layers of carbon fibers; a binding material applied onto and/or between the one or more layers of carbon fibers and configured to at least partially adhere thereto; and a radiation attenuating material applied onto and/or between the one or more carbon fiber layers. Yet a further aspect of the invention pertains to structures obtained from the herein disclosed radiation attenuating composite materials. Yet a further aspect of the invention pertains to radiation shielding apparatuses obtained from the herein disclosed radiation attenuating structures. In one or more embodiments, the herein disclosed structures include a binding material. In one or more embodiments, the herein disclosed structures do not include a binding material. In one or more embodiments, the binding material is a polymer. In one or more embodiments, the binding material is selected from a thermoset resin, polyester, vinyl ester, nylon, and a combination thereof. In one or more embodiments, the thermoset resin is epoxy resin. In one or more embodiments, the herein disclosed structures do not include a binding material selected from a thermoset resin, polyester, vinyl ester, nylon, and a combination thereof. In one or more embodiments, the thermoset resin is epoxy resin. In one or more embodiments, the herein disclosed structures do not include a thermoset resin. In one or more embodiments, the herein disclosed structures do not include a carbon fiber. In one or more embodiments, the radiation attenuating material is a metal. In one or more embodiments, the radiation attenuating material is a metal selected from tungsten, lead, bismuth, antimony, barium, tantalum, and a combination thereof. In one or more embodiments, the composite material further includes a material selected from aramid (e.g. Poly-paraphenylene terephthalamide and the like, which may be known by the trade names Kevlar, Nomex, Technora, and Twaron), aluminum, ultra-high-molecular-weight polyethylene (UHMWPE), glass fibers, and a combination thereof. In one or more embodiments, the binding material and the radiation attenuating material are provided as a liquid or semi-solid substantially homogenous mixture comprising particulates of the radiation attenuating material and the binding material. In one or more embodiments, the radiation attenuating material has a form of a foil. In one or more embodiments, the composite material is arranged as a layered structure comprising one or more layers of the carbon fibers and the binding material and one or more layers of the radiation attenuating material. In one or more embodiments, the composite material has radiation attenuating capacity that is equivalent to or greater than the attenuating capacity of a lead foil having a thickness of 0.1 mm. In one or more embodiments, a layer of the radiation attenuating material has radiation attenuating capacity that is equivalent to or greater than the attenuating capacity of a lead foil having a thickness of 0.1 mm. In one or more embodiments, a layer of the carbon fiber has a thickness of at least about 0.05 mm. In one or more embodiments, the carbon fibers define the outer surface of the layered structure. In one or more embodiments, at least two adjacent layers of the carbon fibers are spaced apart or separated from each other by the radiation attenuating material. In one or more embodiments, the composite material comprises a first and a second layer of carbon fibers, a third layer of the radiation attenuating material, and a third and fourth layers of carbon fibers. In one or more embodiments, the radiation attenuating material layer is disposed between the carbon fiber layers. In one or more embodiments, the composite material comprises one or more layers of the carbon fibers onto which the substantially homogenous composition is applied. In one or more embodiments, the substantially homogenous composition comprises 15% to 95% by weight of the radiation attenuating material and a binding material. In one or more embodiments, the substantially homogenous composition comprises 15% to 60% by weight of the radiation attenuating material and a binding material. In one or more embodiments, the substantially homogenous composition comprises 15% to 80% by weight of the radiation attenuating material and a binding material. In one or more embodiments, the composite material comprises four layers of the carbon fibers onto which the substantially homogenous composition is applied. In one or more embodiments, following curing (e.g., by heating, by applying a high pressure, or by simple drying in the ambient environment) a rigid, low-weight, and radiation attenuating product is obtained having a thickness of at least about 0.3 mm. In one or more embodiments, the composite material comprises two or more types of radiation attenuating material. In yet a further aspect, the invention provides a rigid structure produced by the radiation attenuating composite materials as herein described. In one or more embodiments, the structure has radiation shielding properties. In one or more embodiments, the structure is a rigid tile. In one or more embodiments, the structure is a non-flat rigid structure. In one or more embodiments, the structure is curved. In one or more embodiments, the tile includes one or more curves for accommodating a sliding mechanism. In one or more embodiments, the sliding mechanism includes a rail. In one or more embodiments, the rail is linear. In one or more embodiment, the sliding mechanism includes a slide element that can slide along a sliding mechanism, or a rail. In one or more embodiments, the sliding mechanism includes a friction regulator element, or a bearing element (e.g., a ball bearing) or the alike. In yet a further aspect, the invention provides a substantially homogenous radiation attenuating composition comprising a binding material and particulates of one or more radiation attenuating material. In one or more embodiments, the binding material is selected from a thermoset resin, polyester, vinyl ester, nylon, and a combination thereof. In one or more embodiments, the thermoset resin is epoxy resin. In one or more embodiments, the radiation attenuating material is a metal selected from tungsten, lead, bismuth, antimony, barium, tantalum, and a combination thereof. Unless otherwise defined, all technical or/and scientific terms used herein have the same meaning as commonly understood by one of ordinary skill in the art to which the invention pertains. Although methods and materials similar or equivalent to those described herein can be used in the practice or testing of embodiments of the invention, exemplary methods or/and materials are described below. In case of conflict, the patent specification, including definitions, will control. In addition, the materials, methods, and examples are illustrative only and are not intended to be necessarily limiting. All publications, patents, and patent applications mentioned in this specification are herein incorporated by reference to the same extent as if each individual publication, patent, or patent application was specifically and individually indicated to be incorporated by reference. It should be appreciated that for simplicity and clarity of illustration, elements shown in the figures have not necessarily been drawn to scale. For example, the dimensions of some of the elements are exaggerated relative to each other for clarity. Further, where considered appropriate, reference numerals have been repeated among the figures to indicate corresponding elements. While preferred embodiments of the present invention have been shown and described herein, it will be obvious to those skilled in the art that such embodiments are provided by way of example only. Numerous variations, changes, and substitutions will now occur to those skilled in the art without departing from the invention. It should be understood that various alternatives to the embodiments of the invention described herein may be employed in practicing the invention. It is intended that the following claims define the scope of the invention and that methods and structures within the scope of these claims and their equivalents be covered thereby. It is understood that the invention is not limited to the particular methodology, devices, items or products etc., described herein, as these may vary as the skilled artisan will recognize. It is also to be understood that the terminology used herein is used for the purpose of describing particular embodiments only and is not intended to limit the scope of the invention. The following exemplary embodiments may be described in the context of radiation shielding apparatus and portions thereof for ease of description and understanding. However, the invention is not limited to the specifically described products and methods and may be adapted to various applications without departing from the overall scope of the invention. All ranges disclosed herein include the endpoints. The use of the term “or” shall be construed to mean “and/or” unless the specific context indicates otherwise. The present invention pertains to radiation shielding apparatuses and devices that can be incorporated into radiation emitting systems (e.g. X-ray systems) so as to surround the radiation/X-ray source and/or image intensifier, thereby protecting the surroundings from exposure to scattered radiation. The term “X-ray” and its derivatives may be used interchangeably herein with the term “radiation” and its derivatives; although for the most part the term “X-ray” will be used for ease of understanding and readability, however, without intention to limit the scope of the invention. Features of the present invention pertain to radiation radiopaque tiles (or segments holding these tiles) useful in forming a contiguous shield of a radiation shielding apparatus. The radiation shielding apparatus formed from the herein disclosed tiles are useful in providing protection of the surroundings from exposure to scattered radiation emitted by X-ray systems during fluoroscopic imaging procedures. The radiation shielding apparatus/device of the invention includes an assembly of sequentially arranged stacks of radiation-blocking tiles. Each stack includes a plurality of tiles. The stacks/segments are sequentially arranged with and movably connected to (associated with) each other to form an extendable and contactable tile stack having a plurality of tiles wherein each two adjacent stacks and their tiles partially overlap forming a contiguous radiation shield. The herein disclosed tiles include at (or included in) side-margins thereof extensions to form segments/stacks having a plurality of tiles wherein the extensions (or side margins) of tiles of one stack are arranged to geometrically match (correspond to) and at least partially overlap the extensions (side-margins) of tiles of an adjacent stack, thereby forming a contiguous radiation shield. The tiles (in particular the segments) of a tile stack are deployable. In other words, the tiles arranged parallel to each other in a compact/retracted position (see for example FIG. 3) and can be deployed, forming an extended radiopaque barrier (see for example FIG. 4). The tile stacks are also retractable from an extended position, wherein the deployed tiles slide back onto each other, optionally via a sliding mechanism, returning to the compact/retracted position. In one or more embodiments, the tiles are frame-less, or include no peripheral frame. When in a retracted position, a compact, light-weight segment stack is formed. The segments are arranged sequentially and include corner segments having corner tiles disposed at corners of the assembly of tile/segment stacks, forming a shape that spans the region of X-ray imaging. The radiation shielding assembly may include at least two faces, at least three faces, at least four faces, or at least five faces. The corner segments with corner tiles may span at least about a 90° angular area between tile-segments of adjacent faces within an assembly of the shielding apparatus, thereby covering an entire corner area. The corner segments with corner tiles may span between about a 90° and about a 120° angular area between tile-stacks of adjacent faces within an assembly of the shielding apparatus, thereby covering an entire corner area. The corner segments with corner tiles may span about a 90° angular area in an exemplary square/rectangular-like assembly, thereby covering an entire corner area between tile-stacks of adjacent faces within an assembly of the shielding apparatus. The corner segments with corner tiles may span about a 108° angular area in an exemplary pentagonal-like assembly, thereby covering an entire corner area between tiles-stacks of adjacent faces within an assembly of the shielding apparatus. The corner segments with corner tiles may span about a 120° angular area in an exemplary hexagonal-like assembly, thereby covering an entire corner area between tile-stacks of adjacent faces within an assembly of the shielding apparatus. FIG. 1 schematically illustrates a prior art radiation shielding apparatus. The apparatus is shown in conjunction with a typical C-arm 20 of an X-ray system for performing an X-ray image of a patient. The X-ray system includes a radiation source 22 and a radiation detector 24 mounted on opposing ends of C-arm 20. The apparatus includes a radiopaque or radiation attenuating/blocking shield, which includes at least one radiation shield assembly 28 (e.g. above and below the patient, as illustrated) having a support base 30 connectable to radiation source 22 and/or connected to radiation detector 24, which are mounted on opposite ends of C-arm 20. Radiation shield assembly 28 includes a plurality of radiation shield stacks 32, which include a plurality of stack-tiles (such as tiles 134 shown in FIG. 2). These radiation shield stacks 32 are sequentially positioned relative to support base 30, thereby forming radiopaque screen radiation attenuating/blocking shield in a contiguous configuration. Shield assembly 28 has free ends 38 for spanning the periphery of a body region of the patient. Radiation shield stacks 32 and tiles 134 thereof are controllable to extend or contract to a selected length to position respective free ends 38 in proximity of the patient, or an object such as an X-ray table. In use, radiation source 22 and radiation detector 24 are positioned at opposite sides of the patient, in particular a specific portion of the patient. Radiation source 22 emits an X-ray beam that passes through the specific portion of the patient toward radiation detector 24, which records the exposure to X-ray radiation and sends the image or video feed to a computer and/or a display. FIG. 2 schematically illustrates an exemplary radiation shield assembly 128 of the present invention, which constitutes part of a radiation protection apparatus which is operatively connectable to an X-ray system or the like. The radiation shield assembly 128 is operatively connected to a support base 130 which in turn is connectable to a radiation source 22; and/or connectable to a radiation detector 24. Assembly 128 includes radiation shield stacks 132 (stacks of tiles) sequentially disposed to operably extend from support base 130, thereby forming a contiguous radio-opaque barrier configured for spanning an imaging area during an X-ray procedure. The radiation shield stacks 132 may be individually and actively controllable to extend and retract to a selected length; in other words, respective tiles 134 of the stacks 132 are movable to an extended and retracted position (including partially or fully retracted/extended). Tiles 134 can be considered as constituting or being a part of respective stack segments, and as such, the terms “stack-segments”; “stack-tiles”; and “tiles”, and their derivatives, may be used interchangeably throughout the specification and claims. Assembly 128 may also include flaps 136 at free ends 138 of the stacks 132, for example pivotably attached to the free ends, to aid in surrounding the patient and help limit exposure to scattered radiation. Stacks 132 can be attached to support base 130 via the innermost tile 134 as illustrated in FIG. 2. Alternatively, stacks 132 can be attached to support base 130 via the outermost tile 134. Further, a free end 138 of outermost tile 134 is connected to flap 136 via support base 146 and bracket 147. FIGS. 3 and 4 respectively show stack-tiles 134 in a retracted and an extended position; extension and retraction can be performed using a retraction-extension or tile positioning mechanism 140. In the retracted configuration (FIG. 3), tiles 134 are disposed parallel to each other, forming a stacked compact structure. Tile positioning mechanism 140 may include one or more rails 142, which may be linear, as illustrated; and a slide element 144 configured to slide along respective rails 142. Each tile 134 includes at least one rail 142 and at least one slide element 144. Optionally, each tile 134 includes two rails 142 and two slide elements 144. As can be seen in FIG. 4, in order to allow a compact retracted form, the rails 142a of one tile 134a are disposed offset the rails 142b of the other tile 134b. Similarly, in order to allow a compact retracted form, the slide elements 144a of one tile 134a are disposed offset the slide elements (not shown) of the other tile 134b. Namely, the obtained tile positioning mechanism 140 presents a graded structure that facilitates the compact stack structure. Tile positioning mechanism 140 may include a friction regulator element, or a bearing element (e.g., a ball bearing) or the like, not shown, and may be configured for manual operation, for example simply by pulling and pushing to the desired position, or including a hand crank (which may include a rack and pinion device or a pulley mechanism), not shown. Alternatively or additionally, tile positioning mechanism 140 may further include a powered mechanism including a motor, e.g., an electric motor or a pneumatic or a hydraulic mechanism, not shown. FIGS. 3 and 4 also illustrate that tiles 134 have tile side-margins 148. Side-margins 148 are critical to providing efficient radiation protection and preferably have a non-flat configuration, for example having one or more generally V-shaped or L-shaped ridges, as illustrated. However, other such configurations, for example a wavy or S-shaped configuration (shown in FIG. 9) is also efficient. FIG. 3 also illustrates how stackedly adjacent side-margins 148 correspond one to the next within each stack 132, as well as between stacks 132 on either side thereof. FIG. 4 also illustrates how side-margins 148 of neighboring tiles 134 of neighboring adjacent stacks 132 correspond, as well as between stackedly adjacent tiles. As shown, each tile 134 includes a concave-like structure 148a (FIG. 3) at a first side-margin of the tile 134 and a convex-like structure 148b at a second opposing side-margin of the tile 134. Such concave-convex structure constitutes a substantially stable shield wherein tiles 134 within stacks 132, when deployed, hold each other, maintaining a stable and contiguous radiation attenuating structure/barrier without any detachment of any of the stacks and/or tiles. Thus, it should be understood that tiles 134 of one stack 132 are disposed and arranged such that opposing/neighboring lateral sides (side-margins 148) thereof at least partially overlap lateral sides (side-margins) of tiles of an adjacent stack. Similarly, tiles 134 are disposed and arranged such that bottom and upper ends thereof overlap upper and bottom ends of other vertically (stackedly) disposed adjacent tiles of the same stack, as illustrated in FIG. 4. Such an overlap between the bottom and upper ends can be formed by the overlap of elements of the tile positioning mechanism 140. As a result, a contiguous closed and protective shield with minimal radiation leakage is provided to protect from an X-ray radiation scattering during imaging. FIG. 5 shows rails 142 and slide elements 144 of tile positioning mechanism 140 disposed and accommodated in one or more indentations or recesses 150 of the tiles 134, in particular in spaces or voids formed by the corresponding recesses in one or more stackably adjacent tiles 134 of each stack 132. The term “stackably” refers to the situation where the tiles 134 are one above (or below) a subsequent tile of the same stack 132 when the tiles are in the retracted position. As a result of the configuration of indentions/recesses 150, adjacent stackable tiles 134 (e.g. tile 134a and tile 134b) are correspondingly configured so as to accommodate a rail 142n and a slide element 144n. Such corresponding configuration may be accomplished by subsequent tiles 134 as illustrated, namely, wherein tile 134b is subsequent to tile 134a and tile 134b has a recess 150b that is narrower than a recess 150a of tile 134a and recess 150b fits within recess 150a, like a smaller/narrower tray fits within a larger/wider tray. In the particular design, the width of recess 150b is about two-thirds of the width of recess 150a; and the width of recess 150c is about one-third of the width of recess 150a, and about half the width of recess 150b. FIG. 6 is a perspective view of FIG. 5, which further emphasizes the compact nesting nature of tiles 134, which is significant for space saving. It is noted that the tiles 134 are illustrated with two rails 142 and two respective corresponding slide elements 144; however, mutatis mutandis, tile positioning mechanism 140 could include a different number of such rails and slide elements, for example one, or three, of more. Rail 142n is connected to tile 134a and slide element 144n is connected to tile 134b. Thus, an outermost tile 134 (illustrated as tile 134a) of the stack 132 is attached to support base 146 (FIGS. 3 and 4) and the subsequent adjacent inward tile (tile 134b), by way of slide element 144n, slides on rail 142n. It should be understood that the arrangement may be vice versa, mutatis mutandis, wherein the innermost tile 134 is connected to support base 146 and the subsequent adjacent outward tile, by way of its slide element 144, slides on rail 142. FIG. 7 is a top view of two stacks 132 of four radiation-blocking tiles 134 in a retracted position, illustrating a corner stack 132p. Corner stack 132p is curved or has bends therein to produce an effective corner formation. Side-margins 148 at both sides of corner stack 132p provide for the same tile overlapping as previously described. As such, stacks 132 can form a contiguous radiation protection shield, for example having a generally square profile, although shield structures having other profiles can be produced. Corner stack 132p spans/covers about 90° area located between stacks of two faces of a structure of the assembly of stacks 132. For example, when a radiation shield assembly includes a substantial rectangular or square-like structure, the corner stacks 132p cover the entire 90° area between stacks of adjacent faces. FIG. 8 is a top view of two stacks 1032 of three radiation-blocking tiles 1034 in each stack 1032, in a retracted position, illustrating a zig-zag side margins profile 1048, such that side margins 1048 of neighboring tiles 1034 of neighboring adjacent stacks 1032 at least partially overlap. FIG. 9 is a top view of two stacks 1132 of three radiation-blocking tiles 1134 in each stack 1032 in a retracted position, illustrating an S-shaped side margins profile 1148, such that side margins 1148 of neighboring tiles 1134 of neighboring adjacent stacks 1132 at least partially overlap. As noted above, there is a need to block or minimize, as much as possible, the surroundings from scattered radiation in procedures associated with X-ray-based imaging systems, in order to protect health care providers and technical personnel. To this end the herein-described tiles have the structural features as described above with reference to FIGS. 1-9. These tiles can be manufactured from composite materials as described below with reference to FIGS. 10-20. The tiles may be manufactured from rigid yet low weight radiation attenuating materials. Suitable materials may include composite materials comprising fabrics (e.g., carbon fibers), a binding material (e.g., epoxy, resin), and one or more radiation attenuating materials (e.g., tungsten). Further suitable materials may include composite materials comprising one or more polymers, and one or more radiation attenuating materials (e.g., tungsten). Tile structures obtained from such composite materials are useful in a shielding apparatus as herein disclosed with reference to FIGS. 1-9. Nevertheless, the invention further contemplates other structures that may be useful in various additional fields, such as in aerospace where the properties of radiation attenuation, rigidity and low weight are required. The herein disclosed tiles or other articles may be constructed from a monolayered composite material comprising one or more thermoplastic materials and one or more of radiation blocking materials. Alternatively, optionally or additionally, tiles or other articles, (e.g., laminate structures) may be constructed from a layered structure including a plurality of layers of fiber (e.g., layers of carbon fiber reinforced polymer; CFRP), and one or more layers of radiation blocking material. For example, the radiation blocking material may be supplied as a powder or as a flexible film. Optionally, a resin is included to immobilize the powder and/or stiffen the structure and/or adhere the layers. The structure may include outer layers of carbon fiber and one or more layers of radiation-blocking material in the middle (a “sandwich” structure). Alternatively, tile structures/other articles may include outer layers of carbon fiber and any combination of one or more layers of radiation-blocking material and carbon fiber in the middle. Optionally, the carbon fiber is cut to a desired size and/or shape and/or hardened into a final form (e.g., by heating and/or by applying high pressure, and/or by drying at room temperature). Optionally, the herein disclosed composite materials are formed by injecting a liquid or a pliable raw material of the herein disclosed mixture of a radiation attenuating material and a polymer (e.g., a thermoplastic material) into a mold and solidifying the mixture upon cooling to thereby obtain a rigid structure. For example, the herein described materials may be used to form radiation shielding tiles of desired sizes and shapes. Most commonly used radiation-attenuating materials are heavy metals having high density and atomic number. Thus, incorporating those materials in radiation attenuating devices naturally affects the weight of the resultant article. Structures manufactured from carbon fibers incorporated with the binding polymer or from a thermoplastic material afford rigidity and tensile strength and the radiation attenuating material blocks or minimizes exposure to radiation. The obtained products/tiles may be further advantageously relatively thin having a thickness of about 0.3 mm or above, and optionally, below. Various fiber/fabric types are contemplated. For example, the fiber can be a carbon fiber. Alternatively, the fiber may be a glass fiber, an aramid fiber, a boron fiber, or any combination thereof. The fiber may be in the form of a flexible sheet or a flexible fabric. The thickness of the fiber may vary, for example, the fiber may have a thickness of 0.05 mm or above. For example, 0.1 mm or above, or 0.125 mm or above. Various thermoplastic materials are contemplated. Non limiting examples include thermoplastic elastomers. As used herein the terms “radiation protection material”, “radiation attenuating material”, and their derivatives refer to materials capable of blocking, attenuating, or at least minimizing exposure to radiation. In one or more implementations, the terms include metal or metal alloys. Non limiting examples of radiation attenuating materials include antimony; bismuth; iodine; tungsten; tin; tantalum; erbium; barium; lead; and any combination thereof. Optionally, the radiation attenuating material is provided as a powder. The powder may include particulates having an average size of 0.1 mm or below (e.g., a few microns). Optionally, the radiation attenuating material is mixed with another material such as a polymer, forming a radiation attenuating material-polymer composite (e.g., Tungsten-polymer; Lead-polymer; Bismuth-polymer; Barium-polymer; and any combination of a polymer with a radiation blocking materials). Optionally, the radiation attenuating material is provided as a thin sheet or as a layer. Optionally, the sheet or layer includes an additional material such as a polymer or a rubber. The sheet or layer may be flexible. The sheet or layer may or may not include additional materials. The term “binding material” and derivatives thereof as used herein refers to materials that can act as an adhesive and contribute to the rigidity and strength of a structure when combined with the carbon fibers. Optionally, the binding material solidifies upon heating or when pressurized or when dried in open air. Optionally, the binding material has a glue/binding-like property allowing layers to adhere to each other, at least partially. The binding material optionally adheres to the fibers and optionally is at least partially integrated therewith. The binding material may be a polymer, for example a thermoplastic material (e.g., a polyamide). The binding material may be a thermoset resin. By way of example, the thermoset resins may include polyester; epoxy; phenolic; vinyl ester; polyurethane; silicone; polyamide; and polyamide-imide. In an aspect of the invention there is provided a composition comprising a radiation attenuating material and a binding material. The composition optionally includes a liquid or semi-solid form of the binding material and a radiation attenuating material dispersed therein. The radiation attenuating material may be dispersed, entrapped, and/or distributed within the binding material. Optionally, the radiation attenuating material is dispersed within the binding material as grains having a diameter of 0.1 mm or below. In an exemplary embodiment, the herein disclosed tile structure or articles are manufactured as a non-layered structure; alternatively, as a multi layered structure. A plurality of layers of carbon cloth or carbon fabric, and/or radiation attenuating material, and/or a binding material may be used. In an exemplary embodiment, the tiles/articles are manufactured from at least two, at least three, at least four, at least five, or at least six layers. The term “multi-layered” as used herein is interchangeable with the terms “plurality of layers” and “layered” and refers to two or more layers. In an exemplary embodiment, the herein disclosed tile structure/article is manufactured as a layered, or a multi layered fiber structure. A plurality of carbon fibers may be used. In an exemplary embodiment, the articles are manufactured from at least two, at least three, at least four, at least five, or at least six carbon fiber layers. Optionally, the carbon fibers are one or both external layers. Such configuration may be advantageous as the outer carbon fiber layers provide strength, rigidity and/or structural design for the article. In one or more embodiments, the binding material is applied onto the carbon fiber layers, allowing adhesive properties and optionally increases strength of the carbon fibers. Optionally, at least two of the carbon fiber layers are spaced apart by a layer of the radiation attenuating material. The radiation attenuating material may be disposed within the herein disclosed articles as layers (e.g., sheet). Alternatively, or additionally, the radiation attenuating material may be mixed with the binding material and incorporated or applied to the carbon fibers. Accordingly, the articles or structures are multilayered and include one or more carbon fiber layers onto which a substantially homogenous composition of a binding material and one or more radiation attenuating materials are applied. Non-limiting examples of a layered or a multilayer structure includes two layers of carbon fiber with an intermediate layer of radiation attenuating material. Yet a further example of a layered or multilayer structure includes four layers of carbon fibers with a middle layer of radiation attenuating material. Another non-limiting example of a layered or a multilayer structure includes two layers of carbon fiber incorporated with a mixture of a binding material and a radiation attenuating material. Another non-limiting example of a non-layered structure includes one or more of a thermoelastic material and one or more radiation attenuating materials, optionally in the form of a powder. FIG. 10 illustrates an exemplary layered carbon fiber composite material/structure 100 having a first and a second carbon fiber layer 101 onto which a substantially homogenous composition 102 is applied. Composition 102 includes a binding material 103 (e.g., an epoxy resin), a first radiation attenuating material 104 and a second radiation attenuating material 105. First radiation attenuating material 104 and second radiation attenuating material 105 may be two different materials or may be the same material presenting different forms (e.g., a powder and a sheet), or may be same material having the same form. First radiation attenuating material 104 and second radiation attenuating material 105 may each be selected from tungsten, lead, bismuth, barium, antimony, and tantalum or other radiation attenuating materials. The composition 102 may be applied on one, two, or all sides of each carbon fiber layer. The resulting product is multi-layered and advantageously low-weight, substantially rigid, and capable of attenuating radiation. FIG. 11 schematically illustrates a further exemplary structure or composite material 200 having one carbon fiber layer 201 incorporated on both elongated sides thereof a composition 202 that includes a binding material (e.g., an epoxy resin) and one or more radiation attenuating materials. FIG. 12 schematically illustrates a further exemplary composite material/structure 300 which is similar to composite material 200 but has two carbon fiber layers 301, each surrounded on both elongated sides thereof by a composition 302 having a binding material (e.g., an epoxy resin) and one or more radiation attenuating materials. FIG. 13 schematically illustrates a further exemplary layered composite material/structure 400. Here, a first radiation attenuating layer 404 and a layer of a second radiation attenuating layer 405 have a sheet-like form and may be optionally a metal foil or a rubber sheet. The radiation attenuating layers 404 and 405 are applied such that two carbon fiber layers 401 surround each of the radiation attenuating layers. Altogether, structure 400 has five layers; three carbon fiber layers 401 and two radiation attenuating layers 404 and 405. Structure 400 is shown to include two different radiation attenuating layers 404 and 405, but a similar structure is herein also contemplated wherein the two radiation attenuating layers are the same. A binding material (such as binding material 103 of FIG. 10) may be applied between each of the layers to facilitate strength and adhesiveness between the layers. Optionally, binding material 103 (not shown) may be applied on all sides of the carbon fiber layers 201, so as to harden or adhere the fibers with the binding material. FIGS. 14A-14B show another exemplary composite material 500 (FIG. 14B) and a tile 534 (FIG. 14A) manufactured from the composite material. Composite material 500 includes two outer carbon fiber layers 501 on either side of a radiation attenuating material middle layer 502. A binding material, such as resin (not shown) may be provided between the carbon fiber layers to facilitate strength, and adhesion. Optionally, a binding material may be applied on all sides of the carbon fiber layers 501. Optionally, middle layer 502 includes a radiation attenuating material in the form of a metal foil or a flexible sheet (e.g., a radiation attenuating material plus a rubber). The resulting multilayered article can provide radiation attenuation properties at least equivalent to a minimum of 0.1 mm Pb. Tile 534 is assembled from a layered structure as depicted in FIG. 14B. It should be noted that, although tile 534 illustrates a layered structure as shown in FIG. 14B, alternative structures or composite materials as described herein and depicted in the figures are contemplated and applicable with reference to tile 534. Tile 534 can be a rigid non-flat/curved structure used as a shielding element (tile) in a radiation shielding apparatus that blocks radiation emitted from an X-ray imaging system (shown for example in FIG. 2). The one or more recesses 550 of tile 534 are configured to accommodate linear rails 542 and/or other bearing means, or the alike (not shown). The tile 534 is constructed to hold one or more of a sliding mechanism, bearing means, friction rails, sensors and/or attach additional tiles, via a glue, or by screwing, or by other fastening means. FIG. 15 shows a multilayered composite material/structure 600 including eight carbon fiber sheets or layers 601 and a radiation attenuating material middle layer 602, which may be a metal foil or a flexible rubber sheet or a mixture of radiation attenuating powder and resin. A binding material such as binding material 103 of FIG. 10 may be applied between each of the layers to facilitate strength, and adhesion. Optionally, a binding material may be applied on all sides of the carbon fiber layers 601, to thereby bind the fibers. Optionally, a composition that includes a binding material (e.g., an epoxy resin) and one or more radiation attenuating materials may be applied on one or more of carbon fiber layers 601. FIG. 16 shows composite material/structure 700 including four carbon fiber layers 701 and in the middle thereof two radiation attenuating layers, i.e., layer 704 having a first radiation attenuating material and layer 705 having a second radiation attenuating material. A binding material such as binding material 103 of FIG. 10 may be applied between each of the layers to facilitate strength, and adhesion. Optionally, a binding material may be applied on all sides of carbon fiber layers 701, to thereby bind the fibers. Optionally, a composition that includes a binding material (e.g., an epoxy resin) and one or more radiation attenuating materials may be applied on one or more of carbon fiber layers 701. FIG. 17 shows a multi-layered composite material/structure 800 having altogether seven layers. Four carbon fiber layers 801 are disposed such that two layers are spaced apart by a triple middle layer sub-structure formed by two layers of a radiation attenuating material 802 sandwiching a carbon fiber layer 801. A binding material such as binding material 103 of FIG. 10 may be applied between each of the layers to facilitate strength, and adhesion. Optionally, a binding material may be applied on all sides of carbon fiber layers 801, to thereby bind the fibers. Optionally, a composition that includes a binding material (e.g., an epoxy resin) and one or more radiation attenuating materials may be applied on one or more of carbon fiber layers 801. FIG. 18 shows a multi-layered composite material/structure 900 having altogether seven layers. Four carbon fiber layers 901 are disposed such that two layers are spaced apart by a triple middle layer that includes a layer of a first radiation attenuating material 904, a layer of a second radiation attenuating material 905, and an intermediate non-radiation attenuating spacer layer 906. Spacer layer 906 may be made of a foam (e.g., polyurethane foam) or any other non-radiation attenuating material or non-carbon fiber material. Spacer layer 906 contributes to the strength and stiffness of structure 900. A binding material such as binding material 103 of FIG. 10 may be applied between each of the layers to facilitate strength, and adhesion. Optionally, a binding material may be applied on all sides of the carbon fiber layers 901, to thereby bind the fibers. Optionally, a composition that includes a binding material (e.g., an epoxy resin) and one or more radiation attenuating materials may be applied on one or more of carbon fiber layers 901. FIGS. 19A-19B show another exemplary composite material 1000 (FIG. 19B) and a tile 1034 (FIG. 19A) manufactured from the composite material. The triple layered composite material 1000 includes two carbon fiber layers 1001 sandwiching a middle of a radiation attenuating material layer 1002. A binding material, such as resin (not shown) may be provided between the layers to facilitate strength, and adhesion. Optionally, a binding material may be applied on all sides of the carbon fiber layers 1001. Optionally, middle layer 1002 includes a radiation attenuating material in the form of a metal foil or a flexible sheet (e.g., a radiation attenuating material plus a rubber). Tile 1034 is a rigid non-flat structure having curves that can be used as a shielding element (tile) in a radiation shielding apparatus that blocks radiation emitted from an X-ray imaging system (shown for example in FIG. 2). One or more recesses 1050 of tile 1034 are configured to accommodate linear rails 1042 and/or other bearing means, or the alike (not shown). FIGS. 20A-20B illustrates an exemplary single-layered tile structure 1134 (FIG. 20A) manufactured from a composite material/structure 1100 (FIG. 20B) comprising a radiation attenuating material mixed with a polymer (e.g., a thermoplastic elastomer) 1102. The resulting product is single/mono-layer and advantageously low-weight, substantially rigid, and radiopaque. Tile structure 1134 is a non-flat, curved structure which comprises one or more recesses 1150 for accommodating a sliding mechanism, which may include a linear rail 1142. Tile structure 1134 is constructed to hold one or more of a sliding mechanism/bearing means/sensors/attach additional tiles, via a glue or by screwing, or by other mechanical means. It is to be noted that any of the herein tiles, such as tiles 134 presented in FIGS. 1-7, tiles 1034 presented in FIG. 8 and tiles 1134 presented in FIG. 9 may incorporate any of the herein disclosed materials, such as the composite materials shown in FIGS. 10-20. As illustrated, the tiles of the invention may be produced from a plurality of layers. Alternatively, single layered tiles are contemplated. The tiles may constitute a part of a radiation shielding apparatus (shown for example in FIG. 2), for example an apparatus that can be integrated with or installed onto a C-arm device. As described herein, the tiles require radiation shielding properties, yet should be rigid, light weight and relatively thin. Optionally, the tiles can be made of any combination of layers including: (a) a plurality of fiber layers (e.g., carbon fiber), incorporated with or bound by a binding material (e.g. resin, epoxy) and one or more layers of radiation attenuating material, in the form of a foil or film (e.g., a foil of a radiation attenuating material, and a flexible film polymer having a radiation attenuating material); (b) a plurality of fiber layers (e.g., carbon fiber), disposed/embedded within and/or bound by a mixture of a binding material (e.g. resin, epoxy) and particles of attenuating material (e.g., in the form of powder); (c) a polymer mixed with a radiation attenuating material. Optionally, the thickness of the obtained tile product is between about 0.1 mm and about 5 mm. For example, between about 0.5 mm and about 5 mm; between about 1 mm and about 5 mm; between about 1.5 mm and about 5 mm; between about 0.1 mm and about 4 mm; between about 0.1 mm and about 3.5 mm; between about 0.1 mm and about 3 mm; between about 0.1 mm and about 2.5 mm; between about 0.1 mm and about 2 mm; between about 0.1 mm and about 1.5 mm; between about 0.1 mm and about 1 mm, or any thickness in between. Optionally, the tile has a density of between about 2 g/cc and about 15 g/cc. For example, between about 2 g/cc and about 12 g/cc; between about 2 g/cc and about 10 g/cc; between about 2 g/cc and about 8 g/cc; between about 2 g/cc and about 6 g/cc; between about 2 g/cc and about 4 g/cc; between about 4 g/cc and about 15 g/cc; between about 6 g/cc and about 15 g/cc; between about 8 g/cc and about 15 g/cc; between about 10 g/cc and about 15 g/cc; or any density value in between. Optionally, the tile is non-flat or curved in a shape that allows relative movement between two or more tiles stacked parallel to each other. In order to achieve dynamic, moving tiles, each tile may include one or more rails/slides/bearings. The one or more rails/slides/bearings may be disposed within one or more recesses (concave portions) in the tiles. For example, each tile may include two rails, each disposed within a dedicated recess. Optionally, each tile includes one rail per 10 cm width (e.g. for a tile having a width of about 32 cm, three rails can be incorporated in respective tile recesses). Optionally, the tiles can be incorporated with a tile positioning mechanism to allow relative movement of the tiles with respect to each other, forming a longitudinal dynamic radiation attenuating barrier. Various sliding mechanisms are contemplated and applicable. Non-limiting examples of sliding mechanisms include, linear rails, friction rails, sliding mechanisms with linear bearings, sliding mechanisms with rollers, sliding mechanisms with slide-guide strips. Advantageously, the obtained tiles provide radiation attenuating properties and are rigid allowing stability and stiffness. Further advantageously, the obtained tiles are sufficiently light weight, and thus efficiently dynamic and capable of sliding with respect to each other when provided as an elongated structure, such as a sleeve that can be retracted and deployed to thereby shield a space. A further advantageous property is associated with the tile structure, which is minimalistic in thickness, while still presenting rigidity sufficient to achieve long term stability, resistance to external forces and to allow efficient sliding properties. Yet another aspect of the invention pertains to a method of producing a rigid low-weight radiation attenuating structure, the method comprising: providing one or more carbon fiber fabrics; applying onto and/or between the one or more layers a binding material; and applying or providing onto and/or between the one or more layers a radiation attenuating material. In one or more embodiments, the method comprises a step of curing the carbon fibers, thereby producing a rigid radiation attenuating structure. In one or more embodiments, the method comprises a step of mixing the binding material and the radiation attenuating material to produce a liquid or semi-solid substantially homogenous mixture comprising particulates of the radiation attenuating material and the binding material. In one or more embodiments, the method further comprises applying a layer of the mixture onto the one or more layer of carbon fiber. In one or more embodiments, the radiation attenuating material is in a foil or film-like form. In one or more embodiments, the radiation attenuating material is in a powder form. In view of the above, an aspect of the present invention pertains to a radiation attenuating composite material in accordance with the disclosure herein above. Another aspect of the invention pertains to a substantially homogenous radiation attenuating composition in accordance with the disclosure herein above. Yet another aspect of the invention pertains to a rigid tile structure having a composite material in accordance with the disclosure herein above. Yet another aspect of the invention pertains to a rigid non-flat structure having a composite material in accordance with the disclosure herein above. Yet another aspect of the invention pertains to medical radiation shielding apparatus including a rigid tile structure in accordance with the disclosure herein above. Each of the following terms: ‘includes’, ‘including’, ‘has’, ‘having’, ‘comprises’, and ‘comprising’, and their linguistic equivalents, as used herein, means ‘including, but not limited to’, and is to be taken as specifying the stated component(s), feature(s), characteristic(s), parameter(s), integer(s), or step(s), and does not preclude addition of one or more additional component(s), feature(s), characteristic(s), parameter(s), integer(s), step(s), or groups thereof. The term ‘consisting essentially of’ as used herein means that the scope of the claim is limited to the specified elements and those that do not materially affect the basic and novel characteristic(s) of the claimed device and materials. Each of the phrases ‘consisting of’ and ‘consists of’, as used herein, means ‘including and limited to’. The term ‘method’, as used herein, refers to steps, procedures, manners, means, or/and techniques, for accomplishing a given task including, but not limited to, those steps, procedures, manners, means, or/and techniques, either known to, or readily developed from known steps, procedures, manners, means, or/and techniques, by practitioners in the relevant field(s) of the disclosed invention. Throughout this disclosure, a numerical value of a parameter, feature, characteristic, object, or dimension, may be stated or described in terms of a numerical range format. Such a numerical range format, as used herein, illustrates implementation of some exemplary embodiments of the invention, and does not inflexibly limit the scope of the exemplary embodiments of the invention. Accordingly, a stated or described numerical range also refers to, and encompasses, all possible sub-ranges and individual numerical values (where a numerical value may be expressed as a whole, integral, or fractional number) within that stated or described numerical range. For example, a stated or described numerical range ‘from 1 to 6’ also refers to, and encompasses, all possible sub-ranges, such as ‘from 1 to 3’, ‘from 1 to 4’, ‘from 1 to 5’, ‘from 2 to 4’, ‘from 2 to 6’, ‘from 3 to 6’, etc., and individual numerical values, such as ‘1’, ‘1.3’, ‘2’, ‘2.8’, ‘3’, ‘3.5’, ‘4’, ‘4.6’, ‘5’, ‘5.2’, and ‘6’, within the stated or described numerical range of ‘from 1 to 6’. This applies regardless of the numerical breadth, extent, or size, of the stated or described numerical range. Moreover, for stating or describing a numerical range, the phrase ‘in a range of between about a first numerical value and about a second numerical value’, is considered equivalent to, and meaning the same as, the phrase ‘in a range of from about a first numerical value to about a second numerical value’, and thus, the two equivalently meaning phrases may be used interchangeably. The term ‘about’, is some embodiments, refers to ±30% of the stated numerical value. In further embodiments, the term refers to ±20% of the stated numerical value. In yet further embodiments, the term refers to ±10% of the stated numerical value. It is to be fully understood that certain aspects, characteristics, and features, of the invention, which are, for clarity, illustratively described and presented in the context or format of a plurality of separate embodiments, may also be illustratively described and presented in any suitable combination or sub-combination in the context or format of a single embodiment. Conversely, various aspects, characteristics, and features, of the invention which are illustratively described and presented in combination or sub-combination in the context or format of a single embodiment, may also be illustratively described and presented in the context or format of a plurality of separate embodiments. Although the invention has been described in conjunction with specific embodiments thereof, it is evident that many alternatives, modifications, and variations will be apparent to those skilled in the art. Accordingly, it is intended to embrace all such alternatives, modifications, and variations that fall within the broad scope of the appended claims. All publications, patents, and patent applications mentioned in this specification are herein incorporated in their entirety by reference into the specification, to the same extent as if each individual publication, patent or patent application was specifically and individually indicated to be incorporated herein by reference. In addition, citation or identification of any reference in this application shall not be construed as an admission that such reference is available as prior art to the present invention. To the extent that section headings are used, they should not be construed as necessarily limiting.
	abstract	In a nuclear reactor core, each of a plurality of pressure tubes contains fuel elements spaced apart to permit coolant to flow through spaces between adjacent fuel elements. Each fuel element comprises fuel pellets in cladding, e.g., sapphire, having a melting temperature of at least 1900° C. and does not form significant hydrogen if exposed to high temperature steam. Each pressure tube has an internal insulator sleeve, e.g., fused silica, that has relatively low thermal conductivity over a range of normal operating temperatures and relatively high thermal radiation transmission at temperatures higher than said normal operating temperature range. When coolant is absent from said spaces, the insulator sleeve transmits to the pressure tube at least about 10%, but preferably more than about 40% of thermal radiation from the fuel for conduction through the pressure tube to the moderator and fuel temperature remains within safe limits after the reactor is shut down.
051577020	summary	FIELD OF THE INVENTION The present invention relates to a double crystal monochromator for use in synchrotron radiation studies and more particularly to a mechanically actuated double crystal monochromator which is operative in an ultrahigh vacuum in the order of 10.sup.-10 Torr. BACKGROUND OF THE INVENTION Monochromators are generally well known for use in the energy range from a few kilovolts and up. These monochromators generally use single crystals of silicon or germanium as energy dispersive elements. An article by J. A. Golovchenko of Bell Laboratories printed in the Review of Scientific Instruments, Vol. 52, No. 4, April 1981, describes a double crystal monochromator wherein each crystal element is changed by the experimenter to obtain a constant output direction as well as a constant beam position as energy is varied and to make the central ray of the selected beam impinge at the same point on all monochromator crystals independently of the chosen energy. This was achieved by sampling the output beam intensity for angular drifts out of parallelism between two crystals with an electronic drive to correct the beam. SUMMARY OF THE PRESENT INVENTION The double crystal monochromator according to the present invention is provided with a second crystal which is maintained in a parallel relation to a first crystal within a few seconds of arc and can be traversed linearally in a straight line to capture the diffractive rays from the first crystal as it rotates through an angle of up to 60.degree.. The higher energy levels impinging on the first crystal create a need for cooling of that crystal in an ultrahigh vacuum in a range of 10.sup.-10 Torr. The monochromator advantageously provides for crystal rotation from 12.degree. to 72.degree. from the incoming horizontal beam, crystal angular accuracy of .+-.5 seconds of arc for total beam error through the instrument of .+-.20 seconds of arc (.+-.0.97 mm beam movement at a distance of 10 meters downstream of the second crystal). In another aspect of the invention the beam of the second crystal travels parallel to the incoming beam within 0.0005 inch. In a further aspect of the invention the first crystal which is exposed directly to the synchrotron radiation must be cooled to a temperature less than 150.degree. C. since the first crystal cannot exceed 150.degree. C. maximum temperature. In this regard the second crystal is heated to a temperature within .+-.5.degree. C. of the first crystal in order to maintain a crystalline plane structure spacing the same as the first crystal. In another aspect of the invention a piezoelectric actuator is provided to control the pitch rotation of one crystal relative to the other within a range of 40 seconds of arc. The actuator also provides a dither of up to 2 seconds of arc at frequencies up to 30 Hz. One of the primary advantages of the present invention is the ability to adjust the pitch and roll of each crystal to a registration better than 0.01.degree. while under vacuum. The crystals must also be adjustable in the vertical direction to an accuracy of 0.001 mm. The crystal subassemblies advantageously are also easily removeable from the monochromator. A further advantage of the present invention is the development of an all mechanical system that uses computer control only for scan position, scan rate, piezoelectric inputs and temperature control, thus eliminating the requirement for computer program software control so that all interlocks are mechanically built into the monochromator. Other principal features and advantages of the invention will become apparent to those skilled in the art upon review of the following drawings, the detailed description and the appended claims.
	claims	1. A light-shielding mechanism for preventing a sheet-like member from being irradiated with unwanted light while the sheet-like member is being fed and processed in a sheet-like member processing apparatus, comprising: a charge-eliminating brush disposed on a feed path for the sheet-like member for eliminating electric charges from the sheet-like member in contact therewith; said charge-eliminating brush comprising bristles set at a density ranging from 3,000 to 5,000 bristles/inch and having a thickness ranging from 10 denier to 300 denier. 2. A light-shielding mechanism according to claim 1 , wherein said charge-eliminating brush comprises a first charge-eliminating brush and a second charge-eliminating brush which are disposed on said feed path in confronting relationship to each other and one on each side of said sheet-like member in a transverse direction thereof. claim 1 3. A light-shielding mechanism according to claim 2 , wherein said charge-eliminating brush comprises a third charge-eliminating brush and a fourth charge-eliminating brush which are disposed on said feed path in confronting relationship to each other and one on each edge of said sheet-like member. claim 2 4. A light-shielding mechanism according to claim 1 , wherein said charge-eliminating brush comprise bristles set at a density ranging from 3,500 to 4,400 bristles/inch and having a thickness ranging from 50 denier to 200 denier. claim 1 5. A light-shielding mechanism according to claim 4 , wherein said charge-eliminating brush comprises a first charge-eliminating brush and a second charge-eliminating brush which are disposed on said feed path in confronting to relationship to each other and one on each side of said sheet-like member in a transverse direction thereof. claim 4 6. A light-shielding mechanism according to claim 5 , wherein said charge-eliminating brush comprises a third charge-eliminating brush and a fourth charge-eliminating brush which are disposed on said feed path in confronting relationship to each other and one on each edge of said sheet-like member. claim 5 7. A light-shielding mechanism according to claim 1 , wherein said sheet-like member comprises a stimulable phosphor sheet, said sheet-like member processing apparatus having an erasing unit for erasing remaining radiation information from said stimulable phosphor sheet, said erasing unit having a casing with an inlet defined therein, said charge-eliminating brush being mounted on said inlet. claim 1 8. A light-shielding mechanism according to claim 1 , wherein said sheet-like member comprises a stimulable phosphor sheet, said sheet-like member processing apparatus having an erasing unit for erasing remaining radiation information from said stimulable phosphor sheet, said erasing unit having a casing with an outlet defined therein, said charge-eliminating brush being mounted on said outlet. claim 1
	claims	1. A building for wet storage of spent nuclear fuel, the building comprising:a longitudinal axis;a concrete base mat defining a substantially horizontal operating deck;a plurality of substantially vertical perimeter walls supported by the base mat;a roof spanning across the perimeter walls;a spent fuel pool recessed in the base mat below the operating deck and having a first depth, the spent fuel pool containing coolant water having a surface level;a cask pit recessed in the base mat below the operating deck and formed integrally with the spent fuel pool, the cask pit located adjacent the spent fuel pool and having a second depth;a fluid passageway formed through the base mat between the cask pit and spent fuel pool, the cask pit in fluid communication through the passageway with the spent fuel pool wherein liquid coolant is exchangeable between the spent fuel pool and cask pit; andan isolation gate movably disposed in the passageway and operable to fluidly isolate the cask pit from the spent fuel pool, the gate movable between an open position in which liquid coolant can flow between the spent fuel pool and cask pit, and a closed position in which liquid coolant cannot flow between the spent fuel pool and cask pit;a passive cooling system for cooling the water in the spent fuel pool, the cooling system comprising an internal immersion heat exchanger submerged in the spent fuel pool and an external air cooled heat exchanger located outside the perimeter walls, the immersion heat exchanger located below grade and the air cooled heat exchanger located above grade creating an elevation difference;a closed flow loop formed between the immersion and air cooled heat exchangers, the flow loop containing a circulating heat exchange working fluid, wherein the working fluid circulates by natural gravity driven flow between the immersion and air cooled heat exchangers wherein the spent fuel pool is cooled in the absence of electric power to the facility;wherein the air cooled heat exchanger is disposed inside a cavity of a vertically elongated steel reinforced concrete silo, the silo attached to one of the plurality of substantially vertical perimeter walls of the facility and configured so that no part of the closed flow loop is exposed on an exterior of the building;wherein the silo includes an open top defining an air outlet and air intake vent in a sidewall of the silo defining an air inlet, the air inlet and outlet in fluid communication with the air cooled heat exchanger;wherein the silo includes a vertically extending semicircular sidewall which extends below grade, and opposing ends of the semicircular sidewall are attached to the one of the plurality of substantially vertical perimeter walls; andwherein when the gate is in the open position, water from the spent fuel pool flows into the cask pit wherein the surface level of water in the cask pit and pool equalize. 2. The building according to claim 1, wherein the passageway has an open top which penetrates the operating deck, a closed bottom, and opposing vertical stub walls formed by the base mat. 3. The building according to claim 1, wherein the first depth of the pool is substantially the same as the second depth of the cask pit. 4. The building according to claim 2, wherein the passageway has a height greater than one-half of the second depth of the cask pit to allow a fuel assembly comprising a plurality of fuel rods to pass through the passageway from the cask pit into the spent fuel pool. 5. The building according to claim 1, wherein the cask pit and the spent fuel pool each include a horizontal slab defining an upward facing bottom surface, the bottom surfaces being at substantially the same elevation. 6. The building according to claim 1, further comprising an air lock formed between a pair of outer access doors and a pair of inner access doors spaced longitudinally apart from the outer access doors. 7. The building according to claim 6, further comprising a first cask containing a plurality of nuclear spent fuel assemblies located on the operating deck in an access bay formed between the inner access doors and the cask pit. 8. The building according to claim 7, further comprising a first crane configured and operable to lift and move the first cask from the access bay into the cask pit. 9. The building according to claim 8, wherein the cask pit is sized to hold only the first cask and no others at a same time. 10. The building according to claim 9, further comprising a second cask pit sized to hold only a single cask similar in configuration to the first cask. 11. The building according to claim 7, wherein the passageway has a height sufficient to allow a single fuel assembly removed from the first cask by a second crane so that the single fuel assembly remains underwater while moving the single fuel assembly through the passageway into the spent fuel pool. 12. The building according to claim 1, further comprising an air handling system configured to maintain a negative atmospheric pressure inside the building. 13. The building according to claim 1, wherein the roof has a domed shape. 14. The building according to claim 1, wherein the perimeter walls and roof are formed of steel reinforced concrete. 15. The facility according to claim 1, wherein the one of the plurality of substantially vertical perimeter walls forms a second sidewall of the silo. 16. An autonomous impact resistant wet storage facility for spent nuclear fuel, the facility comprising:a longitudinal axis;a concrete base mat defining a substantially horizontal operating deck, the base mat disposed in soil having a grade and extending for a depth below grade;a plurality of substantially vertical concrete perimeter walls supported by the base mat, the perimeter walls extending upwards from the operating deck and rising above grade;a concrete roof spanning across the perimeter walls and collectively forming an enclosed building with the perimeter walls and base mat;a spent fuel pool disposed in the building and recessed in the base mat below the operating deck for storing nuclear fuel assemblies, the spent fuel pool containing water for cooling the fuel assemblies;a cask pit recessed in the base mat below the operating deck and formed integrally with the spent fuel pool, the cask pit located adjacent the spent fuel pool and having a second depth;a vertically elongated fluid passageway formed through the base mat between the cask pit and spent fuel pool, the passageway extending vertically through the operating deck, wherein the cask pit is in fluid communication through the passageway with the spent fuel pool wherein pool water is flowable between the spent fuel pool and cask pit;an isolation gate movably disposed in the passageway which operates to fluidly isolate the cask pit from the spent fuel pool, the gate movable between an open position in which the passageway is unobstructed so pool water can flow between the spent fuel pool and cask pit, and a closed position in which the passageway is obstructed so pool water cannot flow between the spent fuel pool and cask pit;a passive cooling system for cooling the water in the spent fuel pool, the cooling system comprising an internal immersion heat exchanger submerged in the spent fuel pool and an external air cooled heat exchanger located outside the perimeter walls, the immersion heat exchanger located below grade and the air cooled heat exchanger located above grade creating an elevation difference;a closed flow loop formed between the immersion and air cooled heat exchangers, the flow loop containing a circulating heat exchange working fluid;wherein the working fluid circulates by natural gravity driven flow between the immersion and air cooled heat exchangers wherein the spent fuel pool is cooled in the absence of electric power to the facility;wherein the air cooled heat exchanger is disposed inside a cavity of a vertically elongated steel reinforced concrete silo, the silo attached to one of the plurality of substantially vertical perimeter walls of the facility; andwherein the silo includes a vertically extending semicircular sidewall which extends below grade, and opposing ends of the semicircular sidewall are attached to the one of the plurality of substantially vertical perimeter walls. 17. The facility according to claim 16, wherein the silo includes an open top defining an air outlet and air intake vent in a sidewall of the silo defining an air inlet, the air inlet and outlet in fluid communication with the air cooled heat exchanger. 18. The facility according to claim 16, wherein the one of the plurality of substantially vertical perimeter walls forms a second sidewall of the silo.
	abstract	A process for chemical fixation of radionuclides and radioactive compounds present in soils, solid materials, sludges and liquids. Radionuclides and other radioactive compounds are converted to low-temperature Apatite-Group structural isomorphs (general composition: (AB)5(XO4)3Z), usually phosphatic, that are insoluble, non-leachable, non-zeolitic, and pH stable by contacting with a suspension containing a sulfate, hydroxide, chloride, fluoride and/or silicate source and a phosphate anion. The Apatitic-structure end product is chemically altered from the initial material and reduced in volume and mass. The end product can be void of free liquids and exhibits sufficiently high levels of thermal stability to be effective in the presence of heat generating nuclear reactions. The process occurs at ambient temperature and pressure.
040381382	summary	BACKGROUND OF THE INVENTION This invention relates to nuclear reactor fuel elements. The fuel elements of a nuclear reactor core may be arranged in discrete groups, each group being included in a replaceable construction termed a sub-assembly. One such sub-assembly which is used in a fast reactor cooled by liquid sodium comprises a hexagonal bundle of 127 slender fuel elements enclosed by a steel wrapper about the periphery of the bundle. The fuel elements are each helically wrapped with wire spacers which inhibit bowing and vibration and ensure efficient cooling. However, it is expected that in operation of the nuclear reactor the temperature of the fuel element sheath in the regions underneath the wire wraps will be higher than the remaining regions of the sheath due to the presence of stagnant coolant and to accommodate such a temperature differential the maximum operating temperature of the fuel element would be severely limited. SUMMARY OF THE INVENTION According to the invention, a nuclear reactor fuel element has wire wrap spacing means of a form which contacts the fuel element at a series of regularly spaced intermittent points so that coolant can flow between the wire wrap and the fuel element and between the points of contact.
	description	The present application is related to and claims the benefit of the earliest available effective filing date(s) from the following listed application(s) (the “Related Applications”) (e.g., claims earliest available priority dates for other than provisional patent applications or claims benefits under 35 USC §119(e) for provisional patent applications, for any and all parent, grandparent, great-grandparent, etc. applications of the Related Application(s)). For purposes of the USPTO extra-statutory requirements, the present application constitutes a continuation-in-part of United States Patent Application entitled SYSTEM AND METHOD FOR MAINTAINING AND ESTABLISHING OPERATIONAL READINESS IN A FUEL CELL BACKUP SYSTEM OF A NUCLEAR REACTOR SYSTEM, naming RODERICK A. HYDE, CLARENCE T. TEGREENE, AND JOSHUA C. WALTER as inventors, filed Oct. 1, 2010, application ser. no. 12/924,704, which is currently co-pending, or is an application of which a currently co-pending application is entitled to the benefit of the filing date. The United States Patent Office (USPTO) has published a notice to the effect that the USPTO's computer programs require that patent applicants reference both a serial number and indicate whether an application is a continuation or continuation-in-part. Stephen G. Kunin, Benefit of Prior-Filed Application, USPTO Official Gazette Mar. 18, 2003, available at http://www.uspto.gov/web/offices/com/sol/og/2003/week11/patbene.htm. The present Applicant Entity (hereinafter “Applicant”) has provided above a specific reference to the application(s) from which priority is being claimed as recited by statute. Applicant understands that the statute is unambiguous in its specific reference language and does not require either a serial number or any characterization, such as “continuation” or “continuation-in-part,” for claiming priority to U.S. patent applications. Notwithstanding the foregoing, Applicant understands that the USPTO's computer programs have certain data entry requirements, and hence Applicant is designating the present application as a continuation-in-part of its parent applications as set forth above, but expressly points out that such designations are not to be construed in any way as any type of commentary and/or admission as to whether or not the present application contains any new matter in addition to the matter of its parent application(s). All subject matter of the Related Applications and of any and all parent, grandparent, great-grandparent, etc. applications of the Related Applications is incorporated herein by reference to the extent such subject matter is not inconsistent herewith. The present disclosure generally relates to the implementation of a fuel cell backup system in a nuclear reactor system and, more particularly, to maintaining or establishing a state of operational readiness in a fuel cell backup system of a nuclear reactor system. In one aspect, a method includes but is not limited to maintaining a readiness state of a fuel cell system associated with a nuclear reactor system within a set of readiness parameters, the readiness parameters a function of a characteristic of the nuclear reactor system. In another aspect, a method includes but is not limited to monitoring a characteristic of a nuclear reactor system, and, responsive to the monitored characteristic of the nuclear reactor system, establishing a readiness state of a fuel cell system associated with the nuclear reactor system within a set of readiness parameters, the readiness parameters a function of the characteristic of the nuclear reactor system. In addition to the foregoing, other method aspects are described in the claims, drawings, and text forming a part of the present disclosure. In one or more various aspects, related systems include but are not limited to circuitry and/or programming for effecting the herein-referenced method aspects; the circuitry and/or programming can be virtually any combination of hardware, software, and/or firmware configured to effect the herein—referenced method aspects depending upon the design choices of the system designer. In one aspect, an apparatus includes but is not limited to a fuel cell system associated with a nuclear reactor system, and a fuel cell control system configured to maintain a readiness state of the fuel cell system within a set of readiness parameters, the readiness parameters a function of a characteristic of the nuclear reactor system. In another aspect, an apparatus includes but is not limited to a fuel cell system associated with a nuclear reactor system, a monitoring system configured to monitor a characteristic of the nuclear reactor system, and a fuel cell control system configured to establish a readiness state of the fuel cell system within a set of readiness parameters in response to the monitored characteristic of the nuclear reactor system, the readiness parameters a function of the characteristic of the nuclear reactor system. In addition to the foregoing, other system aspects are described in the claims, drawings, and text forming a part of the present disclosure. In addition to the foregoing, various other method and/or system and/or program product aspects are set forth and described in the teachings such as text (e.g., claims and/or detailed description) and/or drawings of the present disclosure. 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. 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. In the following detailed description, reference is made to the accompanying drawings, which form a part hereof. In the drawings, similar symbols typically identify similar components, 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. Referring generally to FIGS. 1A through 1O, a system 100 for maintaining or establishing a readiness state in a fuel cell backup system of a nuclear reactor system is described in accordance with the present disclosure. One or more monitoring systems 102 may monitor one or more characteristics, such as an operational characteristic or a design characteristic, of a nuclear reactor system 104. Then, the monitoring system may transmit a signal 107 indicative of the one or more monitored characteristics of the nuclear reactor system 104 to a fuel cell control system 108. In response to the signal 107 transmitted by the monitoring system, a fuel cell control system 108 (e.g., a fuel cell control module 109, energy transfer system 112, reactant control system 114, or configuration control system 116) may maintain or establish a readiness state (e.g., electrical output state, temperature state, humidity state, or pressure state) of a fuel cell system 110. An acceptable readiness state may be defined by a set of readiness parameters which are a function (e.g., a variable function) of one or more of the monitored characteristics of the nuclear reactor system 104 measured by the monitoring system 102. While the preceding description refers to a system 100 for maintaining or establishing a readiness state in a fuel cell system 110, hereinafter the system 100 will be described in terms of establishing a readiness state in a fuel cell system 110. This should not, however, be interpreted as a limitation as the remainder of the description should be construed as describing the system 100 and its various embodiments for establishing or maintaining a readiness sate in a fuel cell system 110. In some embodiments, the readiness state established by the fuel cell control system 108 may include, but is not limited to, a readiness state of one or more of the fuel cells of the fuel cell system. For example, the fuel cell control system 108 may establish a temperature, a pressure state, a humidity level or an electrical output level within a portion of one or more of the fuel cells of the fuel cell system. For instance, a monitoring system 102 may monitor a characteristic of the nuclear reactor system 104. Then, the monitoring system may transmit a signal indicative of the monitored characteristic of the nuclear reactor system 104 to a fuel cell control system 108. In response to the transmitted signal from the monitoring system 102, the fuel cell control system 108 may establish a temperature level in one or more of the fuel cells of the fuel cell system 110, wherein the established temperature level is specified by the value of the measured characteristic of the nuclear reactor system 104. For instance, the monitoring system 102 may measure an elevated temperature in the nuclear reactor core of the nuclear reactor system 104. In response to that elevated temperature measurement, the fuel cell control system 108 may establish a temperature level in one or more fuel cells of the fuel cell system 110 in order to increase the response time of the fuel cell system in the event of nuclear reactor system malfunction. It is further recognized that the choice of temperature level may be determined by a computer programmed algorithm of the fuel cell control system 108 which relates a monitored characteristic of the nuclear reactor system to an appropriate temperature level in one or more of the fuel cells of the fuel cell system. By way of another example, a monitoring system 102 may monitor a characteristic of the nuclear reactor system 104. Then, the monitoring system may transmit a signal indicative of the monitored characteristic of the nuclear reactor system 104 to a fuel cell control system 108. In response to the transmitted signal from the monitoring system 102, the fuel cell control system 108 may establish an electrical output level (e.g., current output level or voltage output level) in one or more of the fuel cells of the fuel cell system 110, wherein the established electrical output level is specified by the value of the measured characteristic of the nuclear reactor system. For example, the monitoring system 102 may measure an elevated temperature in the nuclear reactor core of the nuclear reactor system 104. In response to that elevated temperature measurement, the fuel cell control system 108 may establish an electrical output level in one or more fuel cells of the fuel cell system 110 in order to increase the response time of the fuel cell system 110 in the event of nuclear reactor system malfunction. It is further recognized that the choice of the electrical output level may be determined by a computer programmed algorithm of the fuel cell control system 108 which relates a monitored characteristic of the nuclear reactor system to an appropriate electrical output level in one or more of the fuel cells of the fuel cell system. In other embodiments, the readiness state established by the fuel cell control system 108 may include, but is not limited to, a readiness state of one or more of the reactant gases of the fuel cell system 110. For example, the fuel cell control system 108 may establish a temperature, a pressure, a humidity level, or a flow rate in the fuel stream or oxidant stream (e.g., air or reservoir supplied oxidant) of the fuel cell system 110. For instance, a monitoring system 102 may monitor a characteristic of the nuclear reactor system 104. Then, the monitoring system may transmit a signal indicative of the monitored characteristic of the nuclear reactor system 104 to a fuel cell control system 108. In response to the transmitted signal from the monitoring system 102, the fuel cell control system 108 may establish a temperature level in one or both of the reactant gases of the fuel cell system 110, wherein the established temperature level is specified by the value of the measured characteristic of the nuclear reactor system. For example, the monitoring system 102 may measure an elevated temperature in the nuclear reactor core of the nuclear reactor system 104. In response to that elevated temperature measurement, the fuel cell control system 108 may establish a temperature level in one or both of the reactant gases of the fuel cell system 110 in order to increase the response time of the fuel cell system in the event of nuclear reactor system malfunction. It is further recognized that the choice of temperature level may be determined by a computer programmed algorithm of the fuel cell control system 108 which relates a monitored characteristic of the nuclear reactor system to an appropriate temperature level in one or both of the reactant gases of the fuel cell system 110. In another instance, a monitoring system 102 may monitor a characteristic of the nuclear reactor system 104. Then, the monitoring system may transmit a signal indicative of the monitored characteristic of the nuclear reactor system 104 to a fuel cell control system 108. In response to the transmitted signal from the monitoring system 102, the fuel cell control system 108 may establish a flow rate in one or both of the reactant gases of the fuel cell system 110, wherein the established flow rate is specified by the value of the measured characteristic of the nuclear reactor system. For example, the monitoring system 102 may measure an elevated temperature in the nuclear reactor core of the nuclear reactor system 104. In response to that elevated temperature measurement, the fuel cell control system 108 may establish a flow rate in one or both of the reactant gases of the fuel cell system 110 in order to increase the response time of the fuel cell system 110 in the event of nuclear reactor system malfunction. It is further recognized that the choice of the flow rate in either the oxidant gas or fuel gas may be determined by a computer programmed algorithm of the fuel cell control system 108 which relates a monitored characteristic of the nuclear reactor system to an appropriate electrical output level in one or more of the fuel cells of the fuel cell system. Referring now to FIG. 1B, the fuel cell control system 108 may include a fuel cell control module 109 communicatively coupled to one or more subsystems (e.g., energy transfer system 114, reactant control system 116, or configuration control system 118) of the fuel cell control system 108. For example, the fuel cell control system 108 may include a fuel cell control module 109 (e.g., computer controlled data management system) communicatively coupled to an energy transfer system 112 of the fuel cell control system 108 by the transmission of a digital or analog signal 113. For instance, the fuel cell control module 109 may be communicatively coupled to an energy transfer control module 145 of energy transfer system 112. In another example, the fuel cell control system 108 may include a fuel cell control module 109 communicatively coupled to a reactant control system 114 of the fuel cell control system 108 by the transmission of a digital or analog signal 115. For instance, the fuel cell control module 109 may be communicatively coupled to a reactant control module 155 of the reactant control system 114. By way of an additional example, the fuel cell control system 108 may include a fuel cell control module 109 communicatively coupled to a configuration control system 116 of the fuel cell control system 108 by the transmission of a digital or analog signal 117. For instance, the fuel cell control system 108 may include a fuel cell control module 109 communicatively coupled to a configuration control module 167 of the configuration control system 116 of the fuel cell control system 108 by the transmission of a digital or analog signal 117 Further, the fuel cell control module 109 may include a fuel cell control module configured to receive an instruction signal 107 from the monitoring system 102. For instance, a monitoring system 102 may monitor one or more characteristics of a nuclear reactor system 104. Then, the monitoring system 102 may transmit an instruction signal 107 indicative of the one or more monitored characteristics of the nuclear reactor system 104 to a fuel cell control module 109 of the fuel cell control system 108. In response to the transmitted signal 107 from the monitoring system, the fuel cell control module 109 may transmit an instruction signal 113 to an energy transfer system 112 (e.g., energy transfer system control module 145) of the fuel cell control system 108 in order to establish a readiness state in the fuel cell system 110. In another instance, a monitoring system 102 may monitor one or more characteristics of a nuclear reactor system 104. Then, the monitoring system 102 may transmit a signal 107 indicative of the one or more monitored characteristics of the nuclear reactor system 104 to a fuel cell control module 109 of a fuel cell control system 108. In response to the transmitted signal 107 from the monitoring system, the fuel cell control module 109 may transmit an instruction signal 115 to a reactant control system 114 of the fuel cell control system 108 in order to establish a readiness state in the fuel cell system 110. Further, a monitoring system 102 may monitor one or more characteristics of a nuclear reactor system 104. Then, the monitoring system may transmit a signal 107 indicative of the one or more monitored characteristics of the nuclear reactor system 104 to a fuel cell control module 109 of a fuel cell control system 108. In response to the transmitted signal 107 from the monitoring system, the fuel cell control module 109 may transmit an instruction signal 117 to a configuration control system 116 of the fuel cell control system 108 in order to establish a readiness state in the fuel cell system 110. It will be appreciated by those skilled in the art that the fuel cell control module 109 may include signal processing and computer data management hardware and/or software configured to receive a signal transmitted from monitoring system 102 and, based upon that signal, determine appropriate instructions (e.g., via a preprogrammed computer algorithm) for the various subsystems. Then, the fuel cell control module 109 may transmit those appropriate instructions to the required fuel cell control subsystems, such as the energy transfer system 112 (e.g., energy transfer control module 145), the reactant control system 114 (e.g., the reactant control module 155), or the configuration control system 116 (e.g., the configuration control module 167). It will be appreciated by those skilled in the art that the communicative coupling between the fuel cell control module 109 and the fuel cell control subsystems 112-116 and the communicative coupling the between fuel cell control module 109 and the monitoring system 102 may be achieved in various manners. For example, the described components may be communicatively coupled via a digital or analog signal transmitted along a transmission line (e.g., copper wire, coaxial cable, or fiber optic cable) or via a digital or analog wireless signal (e.g., radio frequency signal). It should also be appreciated that the communicative coupling may be achieved via a network connection, wherein the fuel cell control module 109, the monitoring system 102, and the various subsystem control modules (i.e., energy transfer control module 145, reactant control module 155 and configuration control module 167) of the fuel cell control system 108 are connected to a common digital network. It should be recognized that communicative coupling described in the preceding description does not represent a limitation, but rather an illustration as one skilled in the art will appreciate that the communicative coupling between the monitoring system 102 and the fuel cell control module 109 and the communicative coupling between the fuel cell control module 109 and the various subsystems of the fuel cell control system 108 may be achieved through a variety of configurations. Referring now to FIG. 1C, the monitoring system 102 may be directly communicatively coupled to a subsystem (e.g. energy transfer system 112, reactant control system 114 or configuration control system 116) of the fuel cell control system 108. For example, a monitoring system 102 may monitor one or more characteristics of a nuclear reactor system 104. Then, the monitoring system 102 may transmit a signal 107 indicative of the one or more monitored characteristics of the nuclear reactor system 104 directly to an energy transfer system 112 (e.g., energy transfer control module 145) of the fuel cell control system 108. In response to the transmitted signal 107 from the monitoring system, the energy transfer system may transfer energy from an energy source to a portion of the fuel cell system 110 in order to establish a readiness state in the fuel cell system 110. In another example, a monitoring system 102 may monitor one or more characteristics of a nuclear reactor system 104. Then, the monitoring system 102 may transmit a signal 107 indicative of the one or more monitored characteristics of the nuclear reactor system 104 to a reactant control system 114 (e.g., reactant control module 155) of the fuel cell control system 108. In response to the transmitted signal 107 from the monitoring system 102, the reactant control system 114 may adjust conditions of the reactants of the fuel cell system 110 in order to establish a readiness state in the fuel cell system 110. By way of an additional example, a monitoring system 102 may monitor one or more characteristics of a nuclear reactor system 104. Then, the monitoring system 102 may transmit a signal 107 indicative of the one or more monitored characteristics of the nuclear reactor system 104 to a configuration control system 116 (e.g., configuration control module 167) of the fuel cell control system 108. In response to the transmitted signal 107 from the monitoring system, the configuration control system 116 may adjust the configuration of the fuel cells of the fuel cell system 110 in order to establish a readiness state in the fuel cell system 110. It should also be appreciated that the communicative coupling may be achieved via a network connection, wherein the monitoring system 102, and the various subsystem control modules (i.e., energy transfer control module 145, reactant control module 155 and configuration control module 167) of the fuel cell control system 108 are connected to a common network. It should be recognized that communicative coupling described in the preceding description does not represent a limitation, but rather an illustration as one skilled in the art will appreciate that the communicative coupling between the monitoring system 102 and the various subsystems of the fuel cell control system 108 may be achieved through a variety of configurations. Referring now to FIG. 1A through 1H, the fuel cell control system 108 may include an energy transfer system 112 configured to transfer energy from one or more energy sources 103 to a portion of the fuel cell system 110. For example, a monitoring system 102 may monitor one or more characteristics of the nuclear reactor system 104. Then, the monitoring system 102 may transmit a signal indicative of the one or more monitored characteristics of the nuclear reactor system 104 to the fuel cell control system 108. In response to the signal 107 transmitted from the monitoring system 102, the fuel cell control system 108 using an energy transfer system 112 configured to transfer energy from an energy source 103 to a portion of the fuel cell system 110 may establish a readiness state in the fuel cell system 110 by transferring energy (e.g., thermal energy or electrical energy) from an energy source 103 (e.g., portion of the nuclear reactor system 104 or an additional energy source 106) to a portion (e.g., a conditioning system 140 or portion of the fuel cell system block 130) of the fuel cell system 110. Referring now to FIG. 1D, the energy source 103 may include, but is not limited to, a portion of the nuclear reactor system 104 associated with the fuel cell system 110. For example, in response to the signal 107 transmitted by the monitoring system 102, the energy transfer system 112 of the fuel cell control system 108 may transfer energy from a portion of the nuclear reactor system 104 to a portion of the fuel cell system 110 in order to establish a readiness state in the fuel cell system 110. In a further embodiment, the portion of the nuclear reactor system 104 may include, but is not limited to, a portion of a coolant system 118 of the nuclear reactor system 104. For example, in response to the transmitted signal 107 from the monitoring system 102, energy transfer system 112 of the fuel cell control system 108 may transfer energy from a portion of the coolant system 118 of the nuclear reactor system 104 to a portion of the fuel cell system 110 in order to establish a readiness state in the fuel cell system 110. In some embodiments, the coolant system may include a primary coolant system 120 of the nuclear reactor system 104. For instance, in response to the transmitted signal 107 from the monitoring system 102, the energy transfer system 112 may transfer thermal energy from a portion of the primary coolant system 120 (e.g., primary coolant loop), of the nuclear reactor system 104 to a portion of the fuel cell system 110 in order to establish a readiness state in the fuel cell system 110. In another embodiment, the coolant system 118 may include a secondary coolant system 122 of the nuclear reactor system 104. For instance, in response to the transmitted signal 107 from the monitoring system 102, the energy transfer system 112 of the fuel cell control system 108 may transfer thermal energy from a portion of the secondary coolant system 122 (e.g., secondary coolant loop) of the nuclear reactor system 104 to a portion of the fuel cell system 110 in order to establish a readiness state in the fuel cell system 110. In another embodiment, the coolant system 118 may include a waste heat rejection loop 124 of the nuclear reactor system. For instance, a monitoring system 102 may monitor one or more characteristics of a nuclear reactor system 104. Then, the monitoring system 102 may transmit a signal indicative of the one or more monitored characteristics of the nuclear reactor system 104 to the fuel cell control system 108. In response to the signal 107 transmitted from the monitoring system 102, the energy transfer system 112 of the fuel cell control system 108 may transfer thermal energy from a portion of the waste heat rejection loop 124 (e.g., waste heat rejection loop transferring heat to cooling towers of the nuclear reactor system 104) of the nuclear reactor system 104 to a portion of the fuel cell system 110 in order to establish a readiness state in the fuel cell system 110. In a further embodiment, the portion of the nuclear reactor may include, but is not limited to, an electrical output of a thermohydraulic system 126 of the nuclear reactor system 104. For example, in response to the transmitted signal 107 from the monitoring system 102, the energy transfer system 112 of the fuel cell control system 108 may transfer electrical energy from an electrical output of a thermohydraulic system 126 (e.g., electrical output of a generator coupled to a turbine of the nuclear reactor system) of the nuclear reactor system 104 to a portion of the fuel cell system 110 in order to establish a readiness state in the fuel cell system 110. It will be appreciated by those skilled in the art that electricity supplied from an external electrical power ‘grid’ to a portion of the fuel cell system 110 in fact represents electricity supplied, in part, by a turbine-generator system of the nuclear reactor system 104 in situations where the nuclear reactor system 104 supplies electricity to the external power grid. Therefore, supplemental electrical power (e.g., power used to maintain or establish temperature in the fuel cell system 110) that is transferred from the external electrical grid to a portion of the fuel cell system 110 (e.g., temperature control system) is in fact, at least in part, supplied by the nuclear reactor system 104. In another embodiment, the energy source 103 may include, but is not limited to, an additional energy source 128. For example, in response to the transmitted signal 107 from the monitoring system 102, the energy transfer system 112 of the fuel cell control system 108 may transfer energy from a portion of an additional non-nuclear energy source 128 to a portion of the fuel cell system 110 in order to establish a readiness state in the fuel cell system 110. In a further embodiment, the additional energy source 128 may include, but is not limited to, a non-nuclear thermohydraulic electrical generator system. For example, in response to the transmitted signal 107 from the monitoring system 102, the energy transfer system 112 of the fuel cell control system 108 may transfer electrical energy from an electrical output of a non-nuclear powered electrical generator (e.g., diesel powered generator or coal powered generator) to a portion of the fuel cell system 110 in order to establish a readiness state in the fuel cell system 110. In another embodiment, the additional energy source 128 may include, but is not limited to, an energy storage system. For example, a monitoring system 102 may monitor one or more characteristics of a nuclear reactor system 104. Then, the monitoring system 102 may transmit a signal 107 indicative of the one or more monitored characteristics of the nuclear reactor system 104 to the fuel cell control system 108. In response to the transmitted signal 107 from the monitoring system 102, the energy transfer system 112 of the fuel cell control system 108 may transfer energy from an energy storage system (e.g., electrical battery, electrical capacitor, or thermal storage system) to a portion of the fuel cell system 110 in order to establish a readiness state in the fuel cell system 110. Referring again to FIG. 1D, the portion of the fuel cell system 110 may include the fuel cell block 130 of the fuel cell system. For example, a monitoring system 102 may monitor one or more characteristics of a nuclear reactor system 104. Then, the monitoring system 102 may transmit a signal 107 indicative of the one or more monitored characteristics of the nuclear reactor system 104 to the fuel cell control system 108. In response to the transmitted signal 107 from the monitoring system 102, the energy transfer system 112 of the fuel cell control system 108 may transfer energy from an energy source 103 to a portion of the fuel cell block 130 of the fuel cell system 110 in order to establish a readiness state in the fuel cell system 110. For instance, energy may be transferred from a portion of the nuclear reactor system 104 to the fuel cell block 130 of the fuel cell system 110 in order to establish a desired operating temperature of the fuel cell system 110. In a further embodiment, the portion of the fuel cell block 130 may include one or more fuel cell stacks 132 of the fuel cell system 110. For example, in response to the transmitted signal 107 from the monitoring system 102, the energy transfer system 112 of the fuel cell control system 108 may transfer energy from an energy source to one or more fuel cell stacks 130 of the fuel cell system 110 in order to establish a readiness state in the fuel cell system 110. For instance, energy may be transferred from a portion of the nuclear reactor system 104 to individual fuel cell stacks 130 of the fuel cell system 108 in order to establish a desired operating temperature of the fuel cell system. In further embodiment, the portion of the fuel cell block 130 may include one or more individual fuel cells of one or more fuel cell stacks of the fuel cell block. For example, in response to the transmitted signal 107 from the monitoring system 102, the energy transfer system 112 of the fuel cell control system 108 may transfer energy from an energy source 103 to an individual fuel cell 134 of a fuel cell stack 132 of the fuel cell system 110 in order to establish a readiness state in the fuel cell system 110. For instance, energy may be transferred from a portion of the nuclear reactor system 104 to the individual fuel cells 134 of the fuel cell stacks 130 of the fuel cell system 110 in order to establish a desired operating temperature of the fuel cell system. It will be recognized by those skilled in the art that heating individual fuel cell stacks and individual fuel cells allows for more precise control of local thermal conditions within the fuel cell system 110 than a global heating system. In a further embodiment, the portion of a fuel cell 134 may include, but is not limited to, the bipolar plates 136 of a fuel cell 134 of a fuel cell system 110. For example, in response to the transmitted signal 107 from the monitoring system 102, the energy transfer system 112 of the fuel cell control system 108 may transfer thermal energy from an energy source 103 to the bipolar plates of one or more fuel cells 134 of the fuel cell system 110 in order to establish a readiness state in the fuel cell system 110. For instance, thermal energy may be transferred from a portion of the heat rejection loop 124 of the nuclear reactor system 104 to the bipolar plates 136 of one or more fuel cells 134 of the fuel cell system 110 in order to establish a desired operating temperature of the fuel cell system. In another instance, thermal energy may be transferred from a portion of primary coolant system 120 of the nuclear reactor system 104 to the bipolar plates 136 of one or more fuel cells 134 of the fuel cell system 110 in order to establish a desired operating temperature of the fuel cell system. Further, the energy transfer system 112 of the fuel cell control system 108 may transfer thermal energy from an energy source 103 to the flow channels 138 of the bipolar plates 136 of one of more fuel cells 134 of the fuel cell system 110 in order to establish a readiness state in the fuel cell system 110. For instance, thermal energy may be transferred from a portion of the heat rejection loop 124 of the nuclear reactor system 104 to the flow channels 138 of the bipolar plates 136 of one or more fuel cells 134 of the fuel cell system 110 in order to establish a desired operating temperature of the fuel cell system 110. It will be appreciate by those skilled in the art that energy may be transferred from an energy source 130 to the fuel cell system 110 in various ways. For instance, electrical energy from an electrical output of the reactor-generator system may be transferred to an electrical heater in thermal communication with a portion of the fuel cell system 110 in order to establish a desired fuel cell operating temperature. In another instance, a heat transfer system may transfer thermal energy directly from a portion of the nuclear reactor system 104 to a portion of the fuel cell system 110 in order to establish a desired fuel cell operating temperature. The preceding description is not to be construed as a limitation but rather merely an illustration as it is recognized that the preferred mechanism for energy transfer is dependent upon the specific context the present invention is implemented. In another embodiment, the portion of the fuel cell system 110 may include a conditioning system 140 of the fuel cell system 110. For example, a monitoring system 102 may monitor one or more characteristics of a nuclear reactor system 104. Then, the monitoring system 102 may transmit a signal 107 indicative of the one or more monitored characteristics of the nuclear reactor system 104 to the fuel cell control system 108. In response to the transmitted signal 107 from the monitoring system 102, the energy transfer system 112 of the fuel cell control system 108 may transfer energy from an energy source 103 to one or more conditioning systems 140 of the fuel cell system 110 in order to establish a readiness state in the fuel cell system 110. For instance, the conditioning system 140 may use the thermal or electrical energy transferred from the energy source 103 to adjust the conditions of the fuel cell system 110 so as to establish a readiness state within the readiness parameters defined by the measured conditions of the nuclear reactor system 104. In a further embodiment, the condition system 140 may include a humidity control system 142 of the fuel cell system 110. For example, in response to the signal 107 transmitted from the monitoring system 102, the energy transfer system 112 of the fuel cell control system 108 may transfer thermal energy from a portion of the nuclear reactor system 104 to a humidity control system 142 of the fuel cell system 110 in order to establish a desired humidity level in the reactant gas streams or the fuel cell membrane of the fuel cell system 110. For instance, the humidity control system 142 (e.g., humidifier) may use the thermal energy transferred from the energy source 103 to adjust the humidity level in the reactant gas (e.g., fuel or oxidant) in order to establish a readiness state within the readiness parameters defined by the measured conditions of the nuclear reactor system 104. In another instance, the humidity control system 142 may use the thermal energy transferred from the energy source 103 to adjust the humidity level in the fuel cell membrane of the fuel cell system 110 in order to establish a readiness state within the readiness parameters defined by the measured conditions of the nuclear reactor system 104. In another embodiment, the conditioning system 140 may include a temperature control system 142 of the fuel cell system 110. For example, in response to the signal 107 transmitted from the monitoring system 102, the energy transfer system 112 of the fuel cell control system 108 may transfer thermal energy from a portion of the nuclear reactor system 104 to a temperature control system 144 of the fuel cell system 110 in order to establish a desired operating temperature of the fuel cell system 110. For instance, the temperature control system 144 (e.g., temperature control feedback system) may use the energy transferred from the energy source 103 to adjust the temperature of a portion (e.g., reactant gas, bipolar plates, or fuel cell membrane) of the fuel cell system 110 in order to establish a readiness state within the readiness parameters defined by the measured conditions of the nuclear reactor system 104. Referring again to FIG. 1D, the energy transfer system 112 of the fuel cell control system 108 may include a heat transfer system 146 configured to transfer thermal energy from one or more energy sources 103 to a portion of the fuel cell system 110. For example, a monitoring system 102 may monitor one or more characteristics of the nuclear reactor system 104. Then, the monitoring system 102 may transmit a signal 107 indicative of the one or more monitored characteristics of the nuclear reactor system 104 to the fuel cell control system 108. In response to the signal 107 transmitted from the monitoring system 102, the heat transfer system 146 configured to transfer thermal energy from one or more energy sources 103 to a portion of the fuel cell system 110 may establish a readiness state in the fuel cell system 110 by transferring thermal energy from a portion of the nuclear reactor system 104 (e.g., heat rejection loop, portion of the primary coolant system or portion, of secondary coolant system) to a portion of the fuel cell system 110, such as the bipolar plates 138 of one or more of fuel cells 134, the flow channels 136 of one or more fuel cells 134, or one or more conditioning systems 140 (e.g., humidity control system 142 or temperature control system 144). Further, the heat transfer system 146 of the fuel cell control system 108 may be configured to transfer thermal energy from an energy source 103 to a portion of the fuel cell system 110 via thermal convection (e.g., natural convection or forced convection via fluid pumps(s)). Additionally, the heat transfer system 146 of the fuel cell control system 108 may be configured to transfer thermal energy from an energy source 103 to a portion of the fuel cell system 110 via thermal conduction. It will be appreciated by those skilled in the art that the heat transfer system 146 may be configured to transfer thermal energy from a portion of an energy source 103 to the fuel cell system 110 using both thermal conduction and thermal convection. Referring now to FIGS. 1D through 1H, the heat transfer system 146 may include a heat supply loop 152. For example, in response to a signal 107 transmitted by the monitoring system 102, the heat transfer system 146 of the fuel cell control system 108 may establish a readiness state in the fuel cell system 110 by transferring thermal energy from an energy source 103 to a portion of the fuel cell system 110 using one or more heat supply loops 152. For instance, as illustrated in FIG. 1E, in response to a signal 107 transmitted by the monitoring system 102, the heat transfer system 146 of the fuel cell control system 108 may establish a readiness state in the fuel cell system 110 by transferring thermal energy from a portion of the nuclear reactor system 104 (e.g., waste heat rejection loop 124, primary coolant system 120 or secondary coolant system 122) to a portion of the fuel cell system 110 (e.g., conditioning system 140 or bipolar plates 136 of a fuel cell) using one or more heat supply loops 152. In a further embodiment, illustrated in FIG. 1E, the heat supply loop 152 may comprise a heat supply loop having a first portion in thermal communication with a portion of the nuclear reactor system 104 (e.g., primary coolant loop, secondary coolant loop, or a heat rejection loop) and a second portion in thermal communication with a portion of the fuel cell system 110 (e.g., condition system 140 or portion of fuel cell block 130). For instance, in response to a signal 107 transmitted by the monitoring system 102, the heat transfer system 146 of the fuel cell control system 108 may establish a readiness state in the fuel cell system 110 by transferring thermal energy from a portion of the nuclear reactor system 104 to a portion of the fuel cell system 110 using one or more heat supply loops 152 having a first portion in thermal communication with a heat rejection loop 124 of the nuclear reactor system 104 and a second portion in thermal communication with the bipolar plates 136 of one or more fuel cells 134 of the fuel cell system 110. In another instance, in response to a signal 107 transmitted by the monitoring system 102, the heat transfer system 146 of the fuel cell control system 108 may establish a readiness state in the fuel cell system 110 by transferring thermal energy from a portion of the nuclear reactor system 104 to a portion of the fuel cell system 110 using one or more heat supply loops 152 having a first portion in thermal communication with a heat rejection loop 124 of the nuclear reactor system 104 and a second portion in thermal communication with a conditioning system 140 of the fuel cell system 110. In another embodiment, illustrated in FIG. 1F, the heat transfer system 146 may include one or more heat exchangers 154. For example, in response to the signal 107 transmitted by monitoring system 102, the heat transfer system 146 of the fuel cell control system 108 may establish a readiness state in the fuel cell system 112 by transferring thermal energy from a portion of the nuclear reactor system 104 to a portion of the fuel cell system 110 using one or more heat exchangers 154. For instance, the heat exchanger 154 may comprise a heat exchanger having a first portion in thermal communication with a portion of the nuclear reactor system 104 (e.g., primary coolant loop) and a second portion in thermal communication with a portion of the fuel cell system 110 (e.g., flow channels 138 of one or more fuel cells 134). In a further embodiment, the heat transfer system 146 of the fuel cell control system 108 may include a combination of one or more heat exchange loops 152 and one or more heat exchangers 154. For example, as illustrated in FIG. 1F, a first portion of a first heat exchanger 154 may be in thermal communication with a portion of the nuclear reactor system 104, while a second portion of the first heat exchanger 154 may be in thermal communication with the heat supply loop 152. Further, a first portion of a second heat exchanger 154 may be in thermal communication with a portion of the fuel cell system 110, while a second portion of the second heat exchanger 154 may be in thermal communication with the heat supply loop 152. Collectively, the first heat exchanger-heat supply loop-second heat exchanger system acts to transfer thermal energy from a portion of the nuclear reactor system 104 to a portion of the fuel cell system 110 in order to establish a readiness state in the fuel cell system 110 in response to a signal 107 transmitted from the monitoring system 102 to the fuel cell control system 108. By way of another example, illustrated in FIG. 1G, a first portion of a heat exchanger 154 may be in thermal communication with a portion of the nuclear reactor system 104, while a second portion of the heat exchanger 154 may be in thermal communication with a first portion of the heat supply loop 152. In addition, a second portion of the heat supply loop 152 may be in direct thermal communication with a portion of the fuel cell system 110 with no interposed heat exchanger. For instance, the second portion of the heat supply loop 152 may be coupled to a portion of the fuel cell system 110 so that the heat supply loop fluid may be in direct thermal communication (i.e., heat supply fluid is allowed to flow through a portion of the fuel cell system) with a portion of the fuel cell system 110, thus transferring thermal energy directly from the fluid circulated in the heat supply loop to the fuel cell system 110. In an additional example, illustrated in FIG. 1H, a first portion of the heat supply loop 152 may be in direct thermal communication with a portion of the nuclear reactor system 104. Further, a first portion of a heat exchanger 154 may be in thermal communication with a second portion of the heat supply loop 152, while a second portion of the heat exchanger 154 is in thermal communication with a portion of the fuel cell system 110. For instance, the first portion of heat supply loop 152 may be coupled to a heat rejection loop 124 of the nuclear reactor system 104 so that a portion of the fluid (e.g., water) transferred in the heat rejection loop 124 is allowed to flow through the heat supply loop 152. Thermal energy may then be transferred from the heat rejection loop fluid diverted through the heat supply loop 153 to a portion of the fuel cell system 110 via the heat exchanger 154 connected between the second portion of the heat supply loop 152 and the portion of the fuel cell system 110. In another embodiment, the heat transfer system 146 may include a direct fluid exchange system. For example, the heat transfer system 146 may include a heat supply loop 152 configured to transfer fluid from a portion of the nuclear reactor system 104 (e.g., heat rejection loop 124) to a portion of the fuel cell system 110. For instance, a first portion of a heat supply loop 152 may be operably coupled to a heat rejection loop 124 of the nuclear reactor system 104 so that a portion of the heat rejection fluid (e.g., water) is allowed to flow through the heat supply loop 152. Additionally, a second portion of the heat supply loop 152 may be coupled to a portion of the fuel cell system 110 so that the heat rejection fluid may be circulated through a portion of the fuel cell system 110 via the heat supply loop 152. As a result, thermal energy from the fluid circulated in the heat rejection loop 124 may be transferred from the heat rejection fluid to a portion of the fuel cell system 110. It is further contemplated that in order to achieve effective thermal energy transfer via the heat supply loop 152 one or more fluid pumps and one or more valve systems may be utilized in order to circulate the heat rejection fluid through the nuclear reactor system-heat supply loop-fuel cell system circuit. For instance, a fluid carrying heat supply loop 152 may couple a portion of the nuclear reactor system 104 and a portion of the fuel cell system 110, allowing the heat rejection liquid to flow through a portion of the fuel cell system 110. The rate of fluid flow may be controlled by the heat transfer system 146 of the fuel cell control system 108. For instance, a valve system and/or fluid pumps (e.g., mechanical pumps) may be controlled to volumetrically limit the flow through the heat supply circuit It is further contemplated that the fuel cell control module 109 of the fuel cell control system 108 may transmit an instruction signal to the heat transfer system 146 (e.g. via the energy transfer module 145). In addition, it is further recognized that polymer electrolyte membrane (PEM) fuel cells are particularly useful in implementing the present invention as PEM fuel cells have been shown to have an optimal operating temperature (approximately 60 to 160° C.) near the waste heat temperatures of a variety of nuclear reactor systems (e.g., PWR system or BWR system). It is further contemplated that solid oxide fuel cells, which have an optimal operating temperature (approximately 600 to 1000° C.) much higher than PEM fuel cells, may be implemented in the context of a high temperature gas reactor, wherein the heat rejection occurs at a higher temperature than in PWR and BWR reactor systems. Referring again to FIG. 1D, the energy transfer system 112 configured to transfer energy from one or more energy sources 103 to a portion of the fuel cell system 110 may include an electrical transfer system 148 configured to transfer electrical energy form one or more energy sources 103 to a portion of the fuel cell system 110. For example, a monitoring system 102 may monitor one or more characteristics of the nuclear reactor system 104. Then, the monitoring system 102 may transmit a signal 107 indicative of the one or more monitored characteristics of the nuclear reactor system 104 to the fuel cell control system 108. In response to the signal 107 transmitted from the monitoring system 102 the electrical transfer system 148 configured to transfer electrical energy from one or more energy sources 103 to a portion of the fuel cell system 110 may establish a readiness state in the fuel cell system 110 by transferring electrical energy from a portion of the nuclear reactor system 104 (e.g., electrical output of reactor thermohydraulic system) to a portion of the fuel cell system 110, such as a conditioning system 140 (e.g., temperature control system 144 or humidity control system 142) of the fuel cell system 110. In a further embodiment, the electrical transfer system 148 configured to transfer electrical energy form one or more energy sources 103 to a portion of the fuel cell system 110 may include an electrical energy-to-thermal energy conversion system 150. For example, the electrical energy-to-thermal energy conversion system 150 may include, but not limited to, a resistive heating coil or a thermoelectric device configured to convert a portion of the electrical energy produced by the reactor thermohydraulic system to thermal energy. For instance, in response to the signal 107 transmitted by the monitoring system 102, the electrical-to-thermal conversion system 150 of the fuel cell control system 108 may establish a readiness state in the fuel cell system 110 by converting electrical energy from the electrical output of a thermohydraulic system to thermal energy using a resistive heating coil and transferring that thermal energy to a portion of the fuel cell system 110. It will be recognized by those skilled in the art that electrical energy may be used to supplement the heating of a given fuel cell system in instances where the employed fuel cells of the fuel cell system have an optimal operating temperature above the waste heat temperature of the associated nuclear reactor system 104. For example, in a molten carbonate fuel cell (MCFC) system associated with a light water reactor having a heat rejection temperature of 80° C., additional energy must be supplied to the MCFC system in order to reach the system's optimal operating temperature (approximately 600 to 700° C.). It is contemplated that electrical energy may be transferred from an electrical output of a thermohydraulic system of the associated nuclear reactor system 104 to a portion of the MCFC system in order to provide supplemental energy to the MCFC system so that the MCFC system's optimal operating temperature may be achieved and maintained. It should be recognized that the preceding description is not a limitation but merely an illustration as a variety of fuel cell types and nuclear reactor types may be implemented in the context of the present of invention. Referring now to FIG. 1I, the fuel cell control system 108 may include a reactant control system 114 configured to adjust one or more conditions of one or more of the reactant gases of the fuel cell system 110. For example, a monitoring system 102 may monitor one or more characteristics of the nuclear reactor system 104. Then, the monitoring system 102 may transmit a signal indicative of the one or more monitored characteristics of the nuclear reactor system 104 to the fuel cell control system 108. In response to the signal 107 transmitted from the monitoring system 102, the fuel cell control system 108 using a reactant control system 114 configured to adjust a condition (e.g., mass flow rate or pressure) of one or more of the reactant gases (e.g., fuel or oxidant) of the fuel cell system 110 may establish a readiness state in the fuel cell system 110. In a further embodiment, the reactant control system 114 may include, but is not limited to, a reactant pump control system 156 or a reactant valve control system 158. For example, a monitoring system 102 may monitor one or more characteristics of the nuclear reactor system 104. Then, the monitoring system 102 may transmit a signal indicative of the one or more monitored characteristics of the nuclear reactor system 104 to the fuel cell control system 108. In response to the signal 107 transmitted from the monitoring system 102, a reactant pump control system 156 of the fuel cell control system 108 may establish a readiness state in the fuel cell system by adjusting a condition (e.g., mass flow rate or pressure) of one or more of the reactant gases (e.g., fuel or oxidant) of the fuel cell system 110. For instance, in response to a signal 107 transmitted from the monitoring system 102, a reactant pump control system 156 of the reactant control system 114 of the fuel cell control system 108 may adjust (e.g., increase or decrease) the pumping rate of the reactant pumps of the fuel cell system 110. In another instance, in response to a signal 107 transmitted from the monitoring system 102, a reactant pump control system 156 of the reactant control system 114 of the fuel cell control system 108 may activate or deactivate one or more of the reactant pumps of the fuel cell system 110. By way of another example, in response to the signal 107 transmitted by the monitoring system 102, a reactant valve control system 158 of the fuel cell control system 108 may establish a readiness state in the fuel cell system 110 by adjusting a condition (e.g., mass flow rate or pressure) of one or more of the reactant gases (e.g., fuel or oxidant) of the fuel cell system 110. For instance, in response to a signal 107 transmitted by the monitoring system 102, a reactant valve control system 158 of the reactant control system 114 of the fuel cell control system 108 may adjust the flow rate of one or more of the reactant gases by controlling one or more reactant valves of the fuel cell control system 110. It will be recognized by those skilled in the art that reactant pump control system 156 and the reactant valve control system 158 may be used independently or in conjunction with one another to adjust the flow rate or pressure of the fuel gas or oxidant gas of the fuel cell system 110. In addition, it should be recognized that by adjusting the pressure or flow rate of the reactant gases a fuel cell control system 108 may establish a readiness state within the readiness parameters. For example, the voltage and current output levels of a given fuel cell system 110 may be adjusted by increasing or decreasing the reactant pressure in one or more fuel cells of the fuel cell system 110. By way of another example, the temperature of one or more fuel cells may be adjusted by changing the flow rate of the reactant gases. For instance, given a reactant gas held at ambient temperatures, the fuel cell control system 108 may decrease the temperature of a fuel cell membrane of one or more fuel cells at elevated temperatures by increasing the flow rate of the reactant gases being fed into the fuel cell. By way of an additional example, the humidity level of one or more fuel cells may be adjusted by changing the flow rate of the reactant gases. For instance, given a reactant having a first humidity level, the fuel cell control system 108 may decrease or increase the humidity level in a fuel cell membrane by increasing or decreasing the flow rate of the reactant gas being fed into the fuel cell. The preceding description should not be interpreted as a limitation but rather an illustration as it is contemplated that a number of other implementations of the present invention may be applicable in related contexts. In another embodiment, the reactant control system 114 of the fuel cell control system 108 may be used to pre-load a reactant into one or more fuel cells of the fuel cell system 110. For example, a monitoring system 102 may monitor one or more characteristics of the nuclear reactor system 104. Then, the monitoring system 102 may transmit a signal indicative of the one or more monitored characteristics of the nuclear reactor system 104 to the fuel cell control system 108. In response to the signal 107 transmitted from the monitoring system 102, a reactant control system 114 of the fuel cell control system 108 may establish a readiness state in the fuel cell system by pre-loading a reactant into the fuel cell system 110. For instance, a monitoring system 102 may monitor a heightened temperature level in the core of the nuclear reactor system 104. In response, to that temperature level measurement, the reactant control system 114 may pre-load fuel into the fuel cells of the fuel cell system 110. By pre-loading fuel into the fuel cell system 110 the response time required for the fuel cell system 110 to respond to a nuclear reactor malfunction may be shortened. In another embodiment, the reactant control system 114 of the fuel cell control system 108 may be used to unload a reactant from one or more fuel cells of the fuel cell system 110. For example, a monitoring system 102 may monitor one or more characteristics of the nuclear reactor system 104. Then, the monitoring system 102 may transmit a signal indicative of the one or more monitored characteristics of the nuclear reactor system 104 to the fuel cell control system 108. In response to the signal 107 transmitted from the monitoring system 102, a reactant control system 114 of the fuel cell control system 108 may establish a readiness state in the fuel cell system by unloading a reactant from the fuel cell system 110. For instance, a monitoring system 102 may monitor a lowered temperature level in the core of the nuclear reactor system 104. The response time required for a given fuel cell system at lower nuclear reactor core temperatures is smaller than the response time required for the fuel cell system at higher temperature. In response to a lowered nuclear reactor core temperature level measurement, the reactant control system 114 may unload fuel from the fuel cells of the fuel cell system 110. In another embodiment, the reactant control system 114 of the fuel cell control system 108 may include a reactant supply control system 160 configured to adjust one or more supply conditions of one or more of the reactant gases of the fuel cell system 110. For example, a reactant supply control system 160 may include a reactant supply control system configured to control the number of reactant supply tanks supplying reactant gas to the fuel cell system. For example, a monitoring system 102 may monitor one or more characteristics of the nuclear reactor system 104. Then, the monitoring system 102 may transmit a signal indicative of the one or more monitored characteristics of the nuclear reactor system 104 to the fuel cell control system 108. In response to the signal 107 transmitted from the monitoring system 102, the reactant supply control system 160 of the fuel cell control system 108 may establish a readiness state in the fuel cell system 110 by increasing or decreasing the number of reactant reservoir tanks supplying reactant gas to the fuel cells of the fuel cell system. It is further contemplated that the reactant control system 114 may include a reactant control module 155 suitable for controlling the subsystems of the reactant control system (e.g., reactant pump control system 156, reactant valve control system 158 or reactant supply control system 160) in response to a signal transmitted from a fuel cell control module 109 or the monitoring system 102. The reactant control module 155 may include a computer data processing system equipped with signal processing and transmission hardware and software configured to receive a signal transmitted by the fuel cell control module 109 or the monitoring system 102. It is also contemplated that the reactant supply control system 160 may include pump 164 and valve 166 control subsystems that are controlled by a reactant supply control module 162 configured to respond to a signal transmitted from the reactant control module 155, the fuel cell control module 109, or the monitoring system 102. The reactant supply control module 162 may include a computer data processing system equipped with signal processing and transmission hardware and software configured to receive a signal transmitted by the reactant control module 155, the fuel cell control module 109 or the monitoring system 102. Referring now to FIG. 1J, the fuel cell control system 108 may include a configuration control system 116 configured to adjust (i.e., reconfigure) an electrical coupling configuration of two or more of the fuel cells of the fuel cell system 110. For example, a monitoring system 102 may monitor one or more characteristics of the nuclear reactor system 104. Then, the monitoring system 102 may transmit a signal indicative of the one or more monitored characteristics of the nuclear reactor system 104 to the fuel cell control system 108. In response to the signal 107 transmitted from the monitoring system 102, the configuration control system 116 of the fuel cell control system 108 may establish a readiness state in the fuel cell system 110 by adjusting the electrical coupling configuration (e.g., adjusting the electrical circuit arrangement) of two or more of the fuel cells of the fuel cell system 110. For example, the configuration control system may be used to switch the electrical configuration of the fuel cell system 110 from a first configuration to a second configuration in order to adjust the electrical output characteristics (e.g., output current level or voltage level) of the fuel cell control system 110. In a further embodiment, the configuration control system 116 may include configuration control circuitry 168. For example, the configuration control circuitry may include, but is not limited to, switching circuitry 170. For example, a monitoring system 102 may monitor one or more characteristics of the nuclear reactor system 104. Then, the monitoring system 102 may transmit a signal indicative of the one or more monitored characteristics of the nuclear reactor system 104 to the fuel cell control system 108. In response to the signal 107 transmitted from the monitoring system 102, the configuration control system 116 of the fuel cell control system 108 may establish a readiness state in the fuel cell system 110 by adjusting the electrical coupling configuration of two or more of the fuel cells of the fuel cell system 110 using switching circuitry 170. Further, the switching circuitry 170 may include, but is not limited to, one or more transistors 171 (e.g., NPN transistor or PNP transistor) or one or more relay systems. For example, the relay system 172 may include, but is not limited to, an electromagnetic relay system 173 (e.g., a solenoid based relay system), a solid state relay system 174, a transistor switched electromagnetic relay system 175, or a microprocessor controlled relay system 176. For instance, a monitoring system 102 may monitor one or more characteristics of the nuclear reactor system 104. Then, the monitoring system 102 may transmit a signal indicative of the one or more monitored characteristics of the nuclear reactor system 104 to the fuel cell control system 108. In response to the signal 107 transmitted from the monitoring system 102, the configuration control system 116 of the fuel cell control system 108 may establish a readiness state in the fuel cell system 110 by adjusting the electrical coupling configuration of two or more of the fuel cells of the fuel cell system 110 using a transistor switched relay system 175. It is further contemplated that the configuration control system 116 may include a configuration control module 167 suitable for controlling the configuration circuitry 168 in response to a signal transmitted from a fuel cell control module 109 or directly from the monitoring system 102. The configuration control module 167 may include a computer data processing system equipped with signal processing and transmission hardware and software configured to receive a signal transmitted by the fuel cell control module 109 or the monitoring system 102. By way of an additional example, the microprocessor controlled relay system, may include, but is not limited to a microprocessor controlled relay system programmed to respond to one or more conditions 174 (e.g., a signal transmitted from fuel cell control module 109 or a signal transmitted directly from the monitoring system 102). For instance, a monitoring system 102 may monitor one or more characteristics of the nuclear reactor system 104. Then, the monitoring system 102 may transmit a signal indicative of the one or more monitored characteristics of the nuclear reactor system 104 to the fuel cell control system 108. In response to the signal 107 transmitted from the monitoring system 102, the configuration control system 116 of the fuel cell control system 108 may establish a readiness state in the fuel cell system 110 by adjusting the electrical coupling configuration of two or more of the fuel cells of the fuel cell system 110 using a microprocessor controlled relay system programmed to respond to a signal transmitted from the configuration control module 167, fuel cell control module 109, or the monitoring system 102. By way of another example, the switching circuitry 170 may adjust the electrical coupling configuration of two or more of the fuel cells of the fuel cell system 110 by switching a parallel configuration of two or more fuel cells (or fuel cell stacks or fuel cell modules) to a series configuration. Conversely, the switching circuitry 170 may adjust the electrical coupling configuration of two or more of the fuel cells of the fuel cell system 110 by switching a series configuration of two or more fuel cells (or fuel cell stacks or fuel cell modules) to a parallel configuration. It should be appreciated that the switching circuitry 170 may include a number of switching circuitry components which can be controlled independently such that a portion of the switching circuitry components can used to adjust the overall fuel cell system 110 electrical coupling configuration by adjusting the electrical configuration of fuel cells (or fuel cell stacks or fuel cell modules) on an individual basis. In addition, the configuration control circuitry 168 may adjust the electrical configuration of the fuel cell system 110 by adjusting the quantity of fuel cells operating within the fuel cell system 110. For example, the configuration circuitry may be used to couple additional fuel cells (or fuel cell stacks or fuel cell modules) to the fuel cell system 110. Conversely, the configuration circuitry 168 may be used to disconnect fuel cells (or fuel cell stacks or fuel cell modules) from the fuel cell system 110. Referring now to FIG. 1K, the one or characteristics of the nuclear reactor system 104 monitored by the monitoring system may include, but are not limited to, operational characteristics, design characteristics, or nuclear reactor operation system characteristics. For example, the monitoring system 102 may include a monitoring system 178 configured to monitor an operational characteristic of the nuclear reactor system 104. For instance, a monitoring system 178 configured to monitor an operational characteristic of the nuclear reactor system may monitor one or more operational characteristics of the nuclear reactor system 104. Then, the monitoring system 178 configured to monitor an operational characteristic may transmit a signal indicative of the monitored operational characteristic of the nuclear reactor system 104 to the fuel cell control system 108. In response to the signal 107 transmitted from the monitoring system 178 configured to monitor an operational characteristic, the fuel cell control system 108 may establish a readiness state in the fuel cell system 110, where the readiness state is within a set of readiness parameters defined by the operational characteristic of the nuclear reactor system 104. In a further embodiment, the monitoring system 178 configured to monitor an operation characteristic of the nuclear reactor system 104 may monitor one or more characteristics of the nuclear reactor core. For example, an operational characteristic of the nuclear reactor core may include, but is not limited to, thermal characteristics, such as core temperature or the rate of change of the core temperature (e.g., local or average). In another example, the operational characteristic of the nuclear reactor core may include, but is not limited to, the power level of the nuclear reactor core or the reactivity of the nuclear reactor core. Additionally, the operational characteristic of the nuclear reactor core may include, but is not limited to, the pressure in the nuclear reactor core or the rate of change of the pressure in the nuclear reactor core. In a further example, the operational characteristic of the nuclear reactor core may include, but is not limited to, the void fraction in the nuclear reactor. For instance, the monitoring system 178 configured to monitor an operation characteristic of the nuclear reactor system 104 may monitor the void fraction of the nuclear reactor by measuring the coolant flow through the nuclear reactor core. In another instance, the monitoring system 178 configured to monitor an operation characteristic of the nuclear reactor system 104 may monitor the void fraction of the nuclear reactor by measuring a pressure drop in the nuclear reactor core. In an additional instance, the monitoring system 178 configured to monitor an operation characteristic of the nuclear reactor system 104 may monitor the void fraction of the nuclear reactor by measuring the heat output of the nuclear reactor core. In another instance, the monitoring system 178 configured to monitor an operation characteristic of the nuclear reactor system 104 may monitor the void fraction of the nuclear reactor by measuring a pressure drop in the nuclear reactor core. In another instance, the monitoring system 178 configured to monitor an operation characteristic of the nuclear reactor system 104 may monitor the projected afterheat in the nuclear reactor core. In another embodiment, the monitoring system 102 may include a monitoring system 179 configured to monitor a design characteristic of the nuclear reactor system 104. For instance, a monitoring system 179 configured to monitor a design characteristic of the nuclear reactor system may monitor one or more design characteristics of the nuclear reactor system 104. Then, the monitoring system 179 configured to monitor a design characteristic may transmit a signal indicative of the monitored design characteristic of the nuclear reactor system 104 to the fuel cell control system 108. In response to the signal 107 transmitted from the monitoring system 179 configured to monitor a design characteristic, the fuel cell control system 108 may establish a readiness state in the fuel cell system 110, where the readiness state is within a set of readiness parameters defined by the design characteristic of the nuclear reactor system 104. In a further embodiment, the monitoring system 179 configured to monitor a design characteristic of the nuclear reactor system 104 may monitor one or more characteristics of the nuclear reactor core. For example, a design characteristic of the nuclear reactor core may include, but is not limited to, the responsiveness of a safety system of the nuclear reactor system to a design basis accident. A design basis accident may include, but is not limited to, loss of off-site power, reactivity initiated events (e.g., rod withdrawal), loss of flow transients (e.g., pump malfunction), or loss of coolant (e.g., guillotine break or blowdown malfunction). Further, the monitoring system 179 configured to monitor a design characteristic of the nuclear reactor system 104 may monitor the safety system's ability to reestablish coolant flow in the event of a coolant flow loss or the time necessary for the safety system to shut down the nuclear reactor core. By way of another example, a design characteristic of the nuclear reactor core may include, but is not limited to, the time required for a fuel element of the nuclear reactor system to reach a specified temperature upon loss of coolant flow. For instance, the monitoring system 179 configured to monitor a design characteristic of the nuclear reactor system 104 may monitor the time necessary for a portion of a fuel pin assembly to heat to a specified temperature in the event of fuel pump malfunction. Further, the monitoring system 179 configured to monitor a design characteristic of the nuclear reactor system 104 may monitor the time necessary for a a collection of fuel pin assemblies to heat to a specified temperature in the event of fuel pump malfunction. In another embodiment, the monitoring system 102 may include a monitoring system 180 configured to monitor a characteristic of an operation system of the nuclear reactor system 104. For instance, a monitoring system 180 configured to monitor a characteristic of an operation system of the nuclear reactor system may monitor one or more characteristics of an operation system of the nuclear reactor system 104. Then, the monitoring system 180 configured to monitor a a characteristic of an operation system of the nuclear reactor system 104 may transmit a signal indicative of the monitored characteristic of an operation system of the nuclear reactor system 104 to the fuel cell control system 108. In response to the signal 107 transmitted from the monitoring system 180 configured to monitor a characteristic of an operation system of the nuclear reactor system 104, the fuel cell control system 108 may establish a readiness state in the fuel cell system 110, where the readiness state is within a set of readiness parameters defined by the characteristic of the operation system of the nuclear reactor system 104. In a further embodiment, the monitoring system 180 configured to monitor a characteristic of an operation system of the nuclear reactor system 104 may monitor one or more characteristics of a control system of the nuclear reactor system, a coolant system of the nuclear reactor system, a shutdown system of the nuclear reactor system, a monitoring system of the nuclear reactor system, or a safety system of the nuclear reactor. Further, the monitoring system 180 configured to monitor a characteristic of an operation system of the nuclear reactor system 104 may be responsive to a signal transmitted by an operation system of the nuclear reactor system 104. For instance, the monitoring system 180 configured to monitor a characteristic of an operation system of the nuclear reactor system 104 may receive a signal transmitted from the safety system of the nuclear reactor system 104. Then, in response to the signal transmitted from the safety system of the nuclear reactor system 104 the monitoring system 102 may in turn transmit an instruction signal 107 to the fuel cell control system 108. Referring now to FIG. 1L, one or more of the fuel cells 134 of the fuel cell system 110, may include, but are not limited to, a polymer electrolyte fuel cell 182, a solid oxide fuel cell 183, an alkaline fuel cell 184, or a molten carbonate fuel cell 185. For example, one or more monitoring systems 102 may monitor one or more characteristics of a nuclear reactor system 104. Then, the monitoring system may transmit a signal indicative of the one or more monitored characteristics of the nuclear reactor system 104 to a fuel cell control system 108. In response to the transmitted signal from the monitoring system, a fuel cell control system 108 may establish a readiness state in a fuel cell system 110 having one or more polymer electrolyte fuel cells 182. By way of another example, one or more monitoring systems 102 may monitor one or more characteristics of a nuclear reactor system 104. Then, the monitoring system may transmit a signal indicative of the one or more monitored characteristics of the nuclear reactor system 104 to a fuel cell control system 108. In response to the transmitted signal from the monitoring system, a fuel cell control system 108 may establish a readiness state in a fuel cell system 110 having one or more solid oxide fuel cells 183. Referring now to FIG. 1M, the nuclear reactor of the nuclear reactor system 104, may include, but is not limited to, a thermal spectrum nuclear reactor 186, a fast spectrum nuclear reactor 187, a multi-spectrum nuclear reactor 18S, a breeder nuclear reactor 189, or a traveling wave reactor 190. For example, one or more monitoring systems 102 may monitor one or more characteristics of a thermal spectrum nuclear reactor system 186. Then, the monitoring system may transmit a signal indicative of the one or more monitored characteristics of the thermal spectrum nuclear reactor system 186 to a fuel cell control system 108. In response to the transmitted signal 107 from the monitoring system 102, a fuel cell control system 108 may establish a readiness state in the fuel cell system 110. By way of another example, one or more monitoring systems 102 may monitor one or more characteristics of a traveling wave nuclear reactor system 190. Then, the monitoring system may transmit a signal indicative of the one or more monitored characteristics of the traveling wave nuclear reactor system 190 to a fuel cell control system 108. In response to the transmitted signal 107 from the monitoring system 102, a fuel cell control system 108 may establish a readiness state in the fuel cell system 110. Referring now to FIG. 1N, an energy supply system 191 may transfer electrical energy from the electrical output of the fuel cell system 110 to one or more operation systems of the nuclear reactor system 104. For example, the energy supply system 191 may transfer electrical energy from the electrical output of the fuel cell system 110 to a portion of a coolant system (e.g., coolant pump) of the nuclear reactor system. By way of another example, the energy supply system 191 may transfer electrical energy from the electrical output of the fuel cell system 110 to a portion of a shutdown system of the nuclear reactor system 104. It will be recognized by those skilled in the art that the electrical output of the fuel cell system 110 may be used to supplement or augment one or more operation systems of the nuclear reactor system 104 in the event of total or partial malfunction of the nuclear reactor system 104. The operation systems 193 driven or partially driven by the electrical energy transferred from the output of the fuel cell system 110 may include, but are not limited to, a control system, a monitoring system, a warning system, a shutdown system, or a coolant system (e.g., primary coolant system or secondary coolant system). In a further embodiment, the energy supply system 191 may include an energy supply system 192 configured to supply electrical energy to an operation system 193 of the nuclear reactor system 104 in response to a condition. For example, the condition may include, but is not limited to, a signal transmitted by the fuel cell control system 108, a signal from an operation system 193 of the nuclear reactor system 104, a signal from an operator of the nuclear reactor system 104, or a shutdown event of the nuclear reactor system 104. For instance, in response to a signal transmitted from the fuel cell control system 108, the energy supply system 191 may initiate transfer of electrical energy from the output of the fuel cell system 110 to an operation system of the nuclear reactor system. In another instance, in response to a signal transmitted from a safety system of the nuclear reactor system 104, the energy supply system 191 may initiate transfer of electrical energy from the output of the fuel cell system 110 to an operation system of the nuclear reactor system. I should be appreciated by those skilled in the art that the energy supply system 192 may include condition response circuitry configured to initiate transfer of electrical energy from the fuel cell system 110 to an operation system of the nuclear reactor system in response to a condition. For example, the condition response circuitry may include, but is not limited to, one or more transistors (e.g., NPN transistor or PNP transistor) or one or more relay systems. Further, the relay system may include, but is not limited to, an electromagnetic relay system (e.g., a solenoid based relay system), a solid state relay system, a transistor switched electromagnetic relay system, or a microprocessor controlled relay system. Referring now to FIG. 1O, an electrical output of the fuel cell system 110 may be modified using an output modification system 194. For example, the output modification system 194 may include, but is not limited to, power management circuitry 195. For instance, the power management circuitry 195 used to modify the electrical output of the fuel cell system 110 may include, but is not limited to, a power converter, voltage converter (e.g., a DC-DC converter or a DC-AC inverter), or voltage regulation circuitry. Further, the voltage regulation circuitry used to modify the electrical output of the fuel cell system 110 may include, but is not limited to, a Zener diode, a series voltage regulator, a shunt regulator, a fixed voltage regulator or an adjustable voltage regulator. In a further embodiment, the output modification system 194 may include, but is not limited to, control circuitry 196. For instance, the control circuitry 194 may include control circuitry configured to modify the electrical output of the fuel cell system 110 by adjusting the electrical output of the fuel cell system. For example, the control circuitry may be configured to simulate an A.C. electrical output of the fuel cell system 110 by sequentially staging the D.C. outputs of at least two fuel cells of the fuel cell system 110. For instance, the control circuitry may include a plurality of solid state switching devices suitable for sequentially staging the D.C. outputs of two or more fuel cells of the fuel cell system in order to simulate an A.C. signal from the electrical output of the fuel cell system 110. Referring generally to FIG. 2, a system 200 for maintaining a readiness state in a fuel cell backup system of a nuclear reactor system is described in accordance with the present disclosure. One or more monitoring systems 102 may monitor one or more characteristics of a nuclear reactor system 104. Then, the monitoring system 102 may transmit a signal indicative of the one or more monitored characteristics of the nuclear reactor system 104 to a fuel cell control system 108 configured to maintain a readiness state in a fuel cell system 110. In response to the transmitted signal 107 from the monitoring system 102, the fuel cell control system 108 (e.g., a fuel cell control module 109, energy transfer system 112, reactant control system 114, or configuration control system 116) may maintain a readiness state (e.g., electrical output state, temperature state, humidity state, or pressure state) in the fuel cell system 110. For instance, the fuel cell control system 108 may transfer energy from an energy source 103 (e.g., portion of the nuclear reactor system 104 or an additional energy source 112) to a portion of the fuel cell system 110 in order to maintain a readiness state of the fuel cell system 110. An acceptable readiness state is defined by a set of readiness parameters which are a function of one or more of the monitored characteristics of the nuclear reactor system 104 measured by the monitoring system 102. Following 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 alternate 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. FIG. 3 illustrates an operational flow 300 representing example operations related to maintaining a readiness state in a fuel cell backup system of a nuclear reactor system. In FIG. 3 and in following figures that include various examples of operational flows, discussion and explanation may be provided with respect to the above-described examples of FIGS. 1A through 2, and/or with respect to other examples and contexts. However, it should be understood that the operational flows may be executed in a number of other environments and contexts, and/or in modified versions of FIGS. 1A through 2. Also, although the various operational flows are presented in the sequence(s) illustrated, it should be understood that the various operations may be performed in other orders than those which are illustrated, or may be performed concurrently. After a start operation, the operational flow 300 moves to a maintaining operation 310. The maintaining operation 310 depicts maintaining a readiness state of a fuel cell system associated with a nuclear reactor system within a set of readiness parameters, the readiness parameters a function of a characteristic of the nuclear reactor system. For example, as shown in FIGS. 1A through 2, a fuel cell control system 108 (e.g., energy transfer system 112, reactant control system 114, or configuration control system 116) may maintain a readiness state of a fuel cell system 110 within a set of readiness parameters, wherein the readiness parameters are a function of one or more characteristics of the nuclear reactor system 104. By way of another example, a fuel cell module 109 of a fuel cell control system 108 may transmit an instruction signal 113 to an energy transfer system 112 of the fuel cell control system 108 in order to maintain a readiness state of a fuel cell system 110 within a set of readiness parameters. FIG. 4A illustrates alternative embodiments of the example operational flow 300 of FIG. 3. FIG. 4A illustrates example embodiments where the maintaining operation 310 may include at least one additional operation. Additional operations may include an operation 401. The operation 401 illustrates maintaining a readiness state of a fuel cell system associated with a nuclear reactor system within a set of readiness parameters, the readiness parameters a variable function of the characteristic of a nuclear reactor system. For example, as shown in FIGS. 1A through 2, a fuel cell control system 108 may maintain a readiness state within a set of readiness parameters which are a variable function of a characteristic of the nuclear reactor system 104. FIG. 4B illustrates alternative embodiments of the example operational flow 300 of FIG. 3. FIG. 4B illustrates example embodiments where the maintaining operation 310 may include at least one additional operation. Additional operations may include an operation 402, and/or an operation 404. The operation 402 illustrates maintaining a readiness state of a fuel cell system associated with a nuclear reactor system within a set of readiness parameters by transferring energy from an energy source to a portion of the fuel cell system, the readiness parameters a function of the characteristic of a nuclear reactor system. For example, as shown in FIGS. 1A through 2, an energy transfer system 112 of a fuel cell control system 108 may maintain a readiness state within a set of readiness parameters by transferring energy (e.g., thermal or electrical) from an energy source 103 to a portion of the fuel cell system 110. Further, the operation 404 illustrates maintaining a readiness state of a fuel cell system associated with a nuclear reactor system within a set of readiness parameters by transferring energy from the nuclear reactor system to a portion of the fuel cell system, the readiness parameters a function of a characteristic of the nuclear reactor system. For example, as shown in FIGS. 1A through 2, an energy transfer system 112 of a fuel cell control system 108 may maintain a readiness state within a set of readiness parameters by transferring energy (e.g., thermal or electrical) from a portion of the nuclear reactor system 104 to a portion of the fuel cell system 110. FIG. 5 illustrates alternative embodiments of the example operational flow 300 of FIG. 3. FIG. 5 illustrates example embodiments where the maintaining operation 310 may include at least one additional operation. Additional operations may include an operation 502, an operation 504, and/or an operation 506. Further, the operation 502 illustrates maintaining a readiness state of a fuel cell system associated with a nuclear reactor system within a set of readiness parameters by transferring thermal energy from an energy source to a portion of the fuel cell system, the readiness parameters a function of a characteristic of the nuclear reactor system. For example, as shown in FIGS. 1A through 2, a energy transfer system 112 of a fuel cell control system 108 may maintain a readiness state within a set of readiness parameters by transferring thermal energy from an energy source 103 to a portion of the fuel cell system 110 (e.g., bipolar plates of one or more fuel cells). Further, the operation 504 illustrates maintaining a readiness state of a fuel cell system associated with a nuclear reactor system within a set of readiness parameters by transferring thermal energy from an energy source to a portion of the fuel cell system using a heat transfer system, the readiness parameters a function of a characteristic of the nuclear reactor system. For example, as shown in FIGS. 1A through 2, a heat transfer system 146 of a fuel cell control system 108 may maintain a readiness state within a set of readiness parameters by transferring thermal energy from an energy source 103 to a portion of the fuel cell system 110 (e.g., condition system). Further, the operation 506 illustrates maintaining a readiness state of a fuel cell system associated with a nuclear reactor system within a set of readiness parameters by transferring thermal energy from an energy source to a conditioning system of the fuel cell system using a heat transfer system, the readiness parameters a function of a characteristic of the nuclear reactor system. For example, as shown in FIGS. 1A through 2, a heat transfer system 146 of a fuel cell control system 108 may maintain a readiness state within a set of readiness parameters by transferring thermal energy from an energy source 103 to a humidity control system 142 of the fuel cell system 110. FIG. 6 illustrates alternative embodiments of the example operational flow 300 of FIG. 3. FIG. 6 illustrates example embodiments where the maintaining operation 310 may include at least one additional operation. Additional operations may include an operation 602, an operation 604, and/or an operation 606. Further, the operation 602 illustrates maintaining a readiness state of a fuel cell system associated with a nuclear reactor system within a set of readiness parameters by transferring electrical energy from an energy source to a portion of the fuel cell system, the readiness parameters a function of a characteristic of the nuclear reactor system. For example, as shown in FIGS. 1A through 2, an energy transfer system 112 of a fuel cell control system 108 may maintain a readiness state within a set of readiness parameters by transferring electrical energy from an energy source 103 to a temperature control system 144 of the fuel cell system 110. Further, the operation 604 illustrates maintaining a readiness state of a fuel cell system associated with a nuclear reactor system within a set of readiness parameters by transferring electrical energy from an energy source to a portion of the fuel cell system using an electrical energy transfer system, the readiness parameters a function of a characteristic of the nuclear reactor system. For example, as shown in FIGS. 1A through 2, an electrical energy transfer system 148 of a fuel cell control system 108 may maintain a readiness state within a set of readiness parameters by transferring electrical energy from an energy source 103 to a temperature control system 144 of the fuel cell system 110. Further, the operation 606 illustrates maintaining a readiness state of a fuel cell system associated with a nuclear reactor system within a set of readiness parameters by transferring electrical energy from an energy source to a portion of the fuel cell system using an electrical-to-thermal energy conversion system, the readiness parameters a function of a characteristic of the nuclear reactor system. For example, as shown in FIGS. 1A through 2, an electrical-to-thermal conversion system 150 of a fuel cell control system 108 may maintain a readiness state within a set of readiness parameters by transferring electrical energy from an energy source 103 to a portion (e.g., one or more fuel cells) of the fuel cell system 110. FIG. 7 illustrates alternative embodiments of the example operational flow 300 of FIG. 3. FIG. 7 illustrates example embodiments where the maintaining operation 310 may include at least one additional operation. Additional operations may include an operation 702, an operation 704, an operation 706, and/or an operation 708. The operation 702 illustrates maintaining a readiness state of a fuel cell system associated with a nuclear reactor system within a set of readiness parameters by adjusting a condition of at least one reactant of the fuel cell system, the readiness parameters a function of a characteristic of the nuclear reactor system. For example, as shown in FIGS. 1A through 2, the reactant control system 114 of the fuel cell control system 108 may maintain a readiness state of a fuel cell system associated with a nuclear reactor system within a set of readiness parameters by adjusting a condition (e.g., pressure of reactant gas or flow rate of reactant gas) of at least one reactant of the fuel cell system. Further, the reactant pump control system 156 of the reactant control system 114 of the fuel cell control system 108 may maintain a readiness state of a fuel cell system 110 associated with a nuclear reactor system 104 within a set of readiness parameters by adjusting a condition of at least one reactant of the fuel cell system. The operation 704 illustrates maintaining a readiness state of a fuel cell system associated with a nuclear reactor system within a set of readiness parameters by reconfiguring a portion of an electrical configuration of the fuel cell system, the readiness parameters a function of a characteristic of the nuclear reactor system. For example, as shown in FIGS. 1A through 2, the configuration control system 116 (e.g., switching circuitry) of the fuel cell control system 108 may maintain a readiness state of a fuel cell system 110 associated with a nuclear reactor system 104 within a set of readiness parameters by reconfiguring an electrical configuration (e.g., circuit arrangement) of the fuel cell system 110. The operation 706 illustrates maintaining a readiness state of a fuel cell system associated with a nuclear reactor system within a set of readiness parameters, the readiness parameters a function of an operational characteristic of the nuclear reactor system. For example, as shown in FIGS. 1A through 2, the fuel cell control system 108 may maintain a readiness state of a fuel cell system 110 associated with a nuclear reactor system 104 within a set of readiness parameters, the readiness parameters a function of an operational characteristic of the nuclear reactor system (e.g., thermal characteristics). Further, the operation 708 illustrates maintaining a readiness state of a fuel cell system associated with a nuclear reactor system within a set of readiness parameters, the readiness parameters a function of an operational characteristic of a nuclear reactor core of the nuclear reactor system. For example, as shown in FIGS. 1A through 2, the fuel cell control system 108 may maintain a readiness state of a fuel cell system 110 associated with a nuclear reactor system 104 within a set of readiness parameters, the readiness parameters a function of an operational characteristic of the nuclear reactor core (e.g., temperature, power level, pressure, or void fraction). FIG. 8 illustrates alternative embodiments of the example operational flow 300 of FIG. 3. FIG. 8 illustrates example embodiments where the maintaining operation 310 may include at least one additional operation. Additional operations may include an operation 802, an operation 804, and/or an operation 806. The operation 802 illustrates maintaining a readiness state of a fuel cell system associated with a nuclear reactor system within a set of readiness parameters, the readiness parameters a function of a design characteristic of the nuclear reactor system. For example, as shown in FIGS. 1A through 2, the fuel cell control system 108 may maintain a readiness state of a fuel cell system 110 associated with a nuclear reactor system 104 within a set of readiness parameters, the readiness parameters a function of a design characteristic of the nuclear reactor system. Further, the operation 804 illustrates maintaining a readiness state of a fuel cell system associated with a nuclear reactor system within a set of readiness parameters, the readiness parameters a function of the responsiveness of a safety system of the nuclear reactor system to a design basis accident. For example, as shown in FIGS. 1A through 2, the fuel cell control system 108 may maintain a readiness state of a fuel cell system 110 associated with a nuclear reactor system 104 within a set of readiness parameters, the readiness parameters a function of the responsiveness of a safety system of the nuclear reactor system to a design basis accident (e.g., guillotine break). Further, the operation 806 illustrates maintaining a readiness state of a fuel cell system associated with a nuclear reactor system within a set of readiness parameters, the readiness parameters a function of the time required for a fuel element of the nuclear reactor system to reach a specified temperature upon loss of coolant flow. For example, as shown in FIGS. 1A through 2, the fuel cell control system 108 may maintain a readiness state of a fuel cell system 110 associated with a nuclear reactor system 104 within a set of readiness parameters, the readiness parameters a function of the time required for a fuel element, such as a fuel pin assembly or a collection of fuel pin assemblies, of the nuclear reactor system to reach a specified temperature upon loss of coolant flow. FIG. 9 illustrates alternative embodiments of the example operational flow 300 of FIG. 3. FIG. 9 illustrates example embodiments where the maintaining operation 310 may include at least one additional operation. Additional operations may include an operation 902, and/or an operation 904. The operation 902 illustrates maintaining a readiness state of a fuel cell system associated with a nuclear reactor system within a set of readiness parameters, the readiness parameters a function of a characteristic of an operation system of the nuclear reactor system. For example, as shown in FIGS. 1A through 2, the fuel cell control system 108 may maintain a readiness state of a fuel cell system 110 associated with a nuclear reactor system 104 within a set of readiness parameters, the readiness parameters a function of a characteristic of an operation system (e.g., safety system, coolant system, monitoring system or shutdown system) of the nuclear reactor system 104. Further, the operation 904 illustrates maintaining a readiness state of a fuel cell system associated with a nuclear reactor system within a set of readiness parameters, the readiness parameters a function of a signal transmitted from an operation system of the nuclear reactor system. For example, as shown in FIGS. 1A through 2, the fuel cell control system 108 may maintain a readiness state of a fuel cell system 110 associated with a nuclear reactor system 104 within a set of readiness parameters, the readiness parameters a function of a signal (e.g., digital or analog signal) transmitted from an operation system (e.g., safety system, coolant system, monitoring system or shutdown system) of the nuclear reactor system 104. FIG. 10 illustrates alternative embodiments of the example operational flow 300 of FIG. 3. FIG. 10 illustrates example embodiments where the maintaining operation 310 may include at least one additional operation. Additional operations may include an operation 1002, an operation 1004, and/or an operation 1006. The operation 1002 illustrates maintaining an electrical output level of a fuel cell system within an acceptable electrical output range, the acceptable electrical output range a function of a characteristic of the nuclear reactor system. For example, as shown in FIGS. 1A through 2, the fuel cell control system 108 may maintain an electrical output level of a fuel cell system 110 associated with a nuclear reactor system 104 within acceptable electrical output range, the acceptable electrical output range a function of a characteristic of the nuclear reactor system. For instance, the fuel cell control system 108 may transfer thermal energy (via the heat transfer system) to the fuel cell system 110 in order to heat one or more of the fuel cells of the fuel cell system 110 so as to maintain the electrical output level of the fuel cell system within in an acceptable output range. Further, the operation 1004 illustrates maintaining an electrical current output level of a fuel cell system within an acceptable electrical current output range, the acceptable electrical current output range a function of a characteristic of the nuclear reactor system. For example, as shown in FIGS. 1A through 2, the fuel cell control system 108 may maintain an electrical current output level of a fuel cell system 110 associated with a nuclear reactor system 104 within acceptable electrical current output range, the acceptable electrical current output range a function of a characteristic of the nuclear reactor system 104. For instance, the configuration control system 116 of the fuel cell control system 108 may reconfigure (e.g., decouple parallel coupled fuel cells and recouple them in a serial configuration or vice-versa) the electrical coupling configuration of two or more fuel cells of the fuel cell system 110 in order to maintain the electrical current output level of the fuel cell system within in an acceptable output range. Further, the operation 1006 illustrates maintaining a voltage level of a fuel cell system within an acceptable voltage range, the acceptable voltage range a function of a characteristic of the nuclear reactor system. For example, as shown in FIGS. 1A through 2, the fuel cell control system 108 may maintain an electrical voltage output level of a fuel cell system 110 associated with a nuclear reactor system 104 within acceptable electrical voltage output range, the acceptable electrical voltage output range a function of a characteristic of the nuclear reactor system 104. For instance, the configuration control system 116 of the fuel cell control system 108 may reconfigure (e.g., decouple parallel coupled fuel cells and recouple them in a serial configuration or vice-versa) the electrical coupling configuration of two or more fuel cells of the fuel cell system 110 in order to maintain the electrical voltage output level of the fuel cell system within in an acceptable output range. FIG. 11 illustrates alternative embodiments of the example operational flow 300 of FIG. 3. FIG. 11 illustrates example embodiments where the maintaining operation 310 may include at least one additional operation. Additional operations may include an operation 1102, an operation 1104, an operation 1106, an operation 1108 and/or an operation 1110. The operation 1102 illustrates maintaining temperature of a portion of a fuel cell system within an acceptable temperature range, the acceptable temperature range a function of a characteristic of the nuclear reactor system. For example, as shown in FIGS. 1A through 2, the fuel cell control system 108 may maintain a temperature of a portion of a fuel cell system 110 associated with a nuclear reactor system 104 within acceptable temperature range, the acceptable temperature range a function of a characteristic of the nuclear reactor system 104. For instance, the energy transfer system 112 of the fuel cell control system 108 may transfer energy (e.g., thermal or electrical) from an energy source 103 to the fuel cell system 110 in order to heat or cool one or more of the fuel cells of the fuel cell system 110 so as to maintain the temperature of the fuel cell system 110 within in an acceptable temperature range. The operation 1104 illustrates maintaining pressure in a portion of a fuel cell system within an acceptable pressure range, the acceptable pressure range a function of a characteristic of the nuclear reactor system. For example, as shown in FIGS. 1A through 2, the fuel cell control system 108 may maintain a pressure in a portion of a fuel cell system 110 associated with a nuclear reactor system 104 within acceptable pressure range, the acceptable pressure range a function of a characteristic of the nuclear reactor system 104. For instance, the energy transfer system 112 of the fuel cell control system 108 may transfer energy (e.g., thermal or electrical) from an energy source 103 to the fuel cell system 110 in order to maintain the pressure of the fuel cell system 110 within in an acceptable pressure range. The operation 1106 illustrates maintaining a humidity level in a portion of a fuel cell system within an acceptable humidity range, the acceptable humidity range a function of a characteristic of the nuclear reactor system. For example, as shown in FIGS. 1A through 2, the fuel cell control system 108 may maintain a humidity level in a portion of a fuel cell system 110 associated with a nuclear reactor system 104 within acceptable humidity range, the acceptable humidity range a function of a characteristic of the nuclear reactor system 104. For instance, the energy transfer system 112 of the fuel cell control system 108 may transfer energy (e.g., thermal or electrical) from an energy source 103 to the humidity control system of the fuel cell system 110 in order to maintain the humidity level of the fuel cell system 110 within in an acceptable humidity range. The operation 1108 illustrates maintaining temperature of a reactant stream of a fuel cell system within an acceptable temperature range, the acceptable temperature range a function of a characteristic of the nuclear reactor system. For example, as shown in FIGS. 1A through 2, the fuel cell control system 108 may maintain a temperature of a reactant stream (e.g., fuel stream or oxidant stream) of a fuel cell system 110 associated with a nuclear reactor system 104 within acceptable temperature range, the acceptable temperature range a function of a characteristic of the nuclear reactor system 104. For instance, the energy transfer system 112 of the fuel cell control system 108 may transfer energy (e.g., thermal or electrical) from an energy source 103 to the reactant conditioning system of the fuel cell system 110 in order to heat or cool one or more of the reactants of the fuel cell system 110 so as to maintain the temperature of one or both of the reactant streams of the fuel cell system 110 within in an acceptable temperature range. The operation 1110 illustrates maintaining pressure in a reactant stream of a fuel cell system within an acceptable pressure range, the acceptable pressure range a function of a characteristic of the nuclear reactor system. For example, as shown in FIGS. 1A through 2, the fuel cell control system 108 may maintain pressure of a reactant stream (e.g., fuel stream or oxidant stream) of a fuel cell system 110 associated with a nuclear reactor system 104 within an acceptable pressure range, the acceptable pressure range a function of a characteristic of the nuclear reactor system 104. For instance, the reactant control system 114 of the fuel cell control system 108 may control reactant valves and/or pumps of the fuel cell system 110 in order to increase or decrease the flow of one or more of the reactant streams of the fuel cell system 110 so as to maintain the pressure of one or both of the reactant streams of the fuel cell system 110 within in an acceptable pressure range. FIG. 12 illustrates alternative embodiments of the example operational flow 300 of FIG. 3. FIG. 12 illustrates example embodiments where the maintaining operation 310 may include at least one additional operation. Additional operations may include an operation 1202, an operation 1204, an operation 1206, an operation 1208, and/or an operation 1210. The operation 1202 illustrates maintaining humidity of a reactant stream of a fuel cell system within an acceptable humidity range, the acceptable humidity range a function of a characteristic of the nuclear reactor system. For example, as shown in FIGS. 1A through 2, the fuel cell control system 108 may maintain a humidity level of a reactant stream (e.g., fuel stream or oxidant stream) of a fuel cell system 110 associated with a nuclear reactor system 104 within acceptable humidity range, the acceptable humidity range a function of a characteristic of the nuclear reactor system 104. For instance, the energy transfer system 112 of the fuel cell control system 108 may transfer energy (e.g., thermal or electrical) from an energy source 103 to the reactant conditioning system, such as a humidifier, of the fuel cell system 110 in order to maintain the humidity level of one or both of the reactant streams of the fuel cell system 110 within in an acceptable humidity range. The operation 1204 illustrates maintaining a readiness state of a polymer electrolyte membrane fuel cell system within a set of readiness parameters, the readiness parameters a function of a characteristic of the nuclear reactor system. For example, as shown in FIGS. 1A through 2, a fuel cell control system 108 may maintain a readiness state of a polymer electrolyte membrane fuel cell system 182 within a set of readiness parameters, wherein the readiness parameters are a function of one or more characteristics of the nuclear reactor system 104. The operation 1206 illustrates maintaining a readiness state of a solid oxide fuel cell system within a set of readiness parameters, the readiness parameters a function of a characteristic of the nuclear reactor system. For example, as shown in FIGS. 1A through 2, a fuel cell control system 108 may maintain a readiness state of a solid oxide fuel cell system 183 within a set of readiness parameters, wherein the readiness parameters are a function of one or more characteristics of the nuclear reactor system 104. The operation 1208 illustrates maintaining a readiness state of an alkaline fuel cell system within a set of readiness parameters, the readiness parameters a function of a characteristic of the nuclear reactor system For example, as shown in FIGS. 1A through 2, a fuel cell control system 108 may maintain a readiness state of an alkaline fuel cell system 184 within a set of readiness parameters, wherein the readiness parameters are a function of one or more characteristics of the nuclear reactor system 104. The operation 1210 illustrates maintaining a readiness state of a molten carbonate fuel cell system within a set of readiness parameters, the readiness parameters a function of a characteristic of the nuclear reactor system. For example, as shown in FIGS. 1A through 2, a fuel cell control system 108 may maintain a readiness state of a molten carbonate fuel cell system 185 within a set of readiness parameters, wherein the readiness parameters are a function of one or more characteristics of the nuclear reactor system 104. FIG. 13 illustrates an operational flow 1300 representing example operations related to maintaining a readiness state in a fuel cell backup system of a nuclear reactor system. FIG. 13 illustrates an example embodiment where the example operational flow 300 of FIG. 3 may include at least one additional operation. Additional operations may include an operation 1310, and/or an operation 1312. After a start operation and a maintaining operation 310, the operational flow 1300 moves to a transferring operation 1310. Operation 1310 illustrates transferring electrical energy from a fuel cell system to an operation system of the nuclear reactor system. For example, as shown in FIGS. 1A through 2, an energy supply system 191 may transfer electrical energy from the electrical output of the fuel cell system 110 to an operation system (e.g., coolant system or shutdown system) of the nuclear reactor system. The operation 1312 illustrates, responsive to at least one condition, transferring electrical energy from a fuel cell system to an operation system of the nuclear reactor system. For example, as shown in FIGS. 1A through 2, an energy supply system 191 may transfer electrical energy from the electrical output of the fuel cell system 110 to an operation system (e.g., coolant system or shutdown system) of the nuclear reactor system in response to a condition, such as a signal from an operation system of the nuclear reactor system 104, or a shutdown event of the nuclear reactor system 104. FIG. 14 illustrates an operational flow 1400 representing example operations related to maintaining a readiness state in a fuel cell backup system of a nuclear reactor system. FIG. 25 illustrates an example embodiment where the example operational flow 300 of FIG. 3 may include at least one additional operation. Additional operations may include an operation 1410, an operation 1412, an operation 1414, and/or an operation 1416. After a start operation and a maintaining operation 310, the operational flow 1400 moves to a modifying operation 1410. Operation 1410 illustrates modifying an electrical output of the fuel cell system. For example, as shown in FIGS. 1A through 2, the output modification system 194 may modify the characteristics of the electrical output of the fuel cell system 110. The operation 1412 illustrates modifying an electrical output of the fuel cell system using power management circuitry. For example, as shown in FIGS. 1A through 2, power management circuitry 195 (e.g., voltage regulation circuitry) may modify the characteristics of the electrical output of the fuel cell system 110. The operation 1414 illustrates modifying the electrical output of the fuel cell system by adjusting the electrical output of at least one fuel cell of the fuel cell system using control circuitry. For example, as shown in FIGS. 1A through 2, control circuitry 196 may modify the characteristics of the electrical output of the fuel cell system 110 by adjusting the electrical output of one or more fuel cells of the fuel cell system. Further, the operation 1416 illustrates simulating an A.C. electrical output of the fuel cell system by sequentially staging a D.C. output of at least two fuel cells of the fuel cell system. For example, as shown in FIGS. 1A through 2, control circuitry 196 may include solid state switches configured to simulate an A.C. electrical output of the fuel cell system 110 by sequentially staging the D.C. electrical outputs of two or more fuel cells of the fuel cell system 110. FIG. 15 illustrates an operational flow 1500 representing example operations related to establishing a readiness state in a fuel cell backup system of a nuclear reactor system. In FIG. 15 and in following figures that include various examples of operational flows, discussion and explanation may be provided with respect to the above-described examples of FIGS. 1 through 2, and/or with respect to other examples and contexts. However, it should be understood that the operational flows may be executed in a number of other environments and contexts, and/or in modified versions of FIGS. 1A through 2. Also, although the various operational flows are presented in the sequence(s) illustrated, it should be understood that the various operations may be performed in other orders than those which are illustrated, or may be performed concurrently. After a start operation, the operational flow 1500 moves to a monitoring operation 1510. The monitoring operation 1510 depicts monitoring characteristics of a nuclear reactor system. For example, as shown in FIGS. 1A through 2, a monitoring system 102 may monitor one or more characteristics (e.g., operation characteristics of the nuclear reactor, design characteristics of the nuclear reactor, or operational characteristics of an operation system of the nuclear reactor). Then, the establishing operation 1520 depicts, responsive to the monitored characteristics of the nuclear reactor system, establishing a readiness state of a fuel cell system associated with the nuclear reactor system within a set of readiness parameters, the readiness parameters a function of a characteristic of the nuclear reactor system. For example, as shown in FIGS. 1A through 2, in response to a signal 107 (e.g., digital or analog signal transmitted wirelessly or by wireline) transmitted by the monitoring system 102, a fuel cell control system 108 (e.g., energy transfer system 112, reactant control system 114, or configuration control system 116) may establish a readiness state of a fuel cell system 110 within a set of readiness parameters, wherein the readiness parameters are a function of one or more of the characteristics of the nuclear reactor system 104. By way of another example, in response to the signal 107 transmitted by the monitoring system 102, a fuel cell module 109 of a fuel cell control system 108 may transmit an instruction signal 113 to an energy transfer system 112 of the fuel cell control system 108 in order to maintain a readiness state of a fuel cell system 110 within a set of readiness parameters. FIG. 16A illustrates alternative embodiments of the example operational flow 1500 of FIG. 15. FIG. 16A illustrates example embodiments where the establishing operation 1520 may include at least one additional operation. Additional operations may include an operation 1601. The operation 1601 illustrates, responsive to the monitored characteristics of the nuclear reactor system, establishing a readiness state of a fuel cell system associated with the nuclear reactor system within a set of readiness parameters, the readiness parameters a variable function of a characteristic of the nuclear reactor system. For example, as shown in FIGS. 1A through 2, in response to the signal 107 transmitted by the monitoring system 102, a fuel cell control system 108 may establish a readiness state within a set of readiness parameters which are a variable function of a characteristic of the nuclear reactor system 104. FIG. 16B illustrates alternative embodiments of the example operational flow 1500 of FIG. 15. FIG. 16 illustrates example embodiments where the establishing operation 1520 may include at least one additional operation. Additional operations may include an operation 1602, and/or an operation 1604. The operation 1602 illustrates, responsive to the monitored characteristics of the nuclear reactor system, establishing a readiness state of a fuel cell system within a set of readiness parameters by transferring energy from an energy source to a portion of the fuel cell system, the readiness parameters a function of a characteristic of the nuclear reactor system. For example, as shown in FIGS. 1A through 2, in response to the signal 107 transmitted by the monitoring system 102, an energy transfer system 112 of a fuel cell control system 108 may establish a readiness state within a set of readiness parameters by transferring energy (e.g., thermal or electrical) from an energy source 103 to a portion of the fuel cell system 110. Further, the operation 1604 illustrates, responsive to the monitored characteristics of the nuclear reactor system, establishing a readiness state of a fuel cell system within a set of readiness parameters by transferring energy from a portion of the nuclear reactor system to a portion of the fuel cell system, the readiness parameters a function of a characteristic of the nuclear reactor system. For example, as shown in FIGS. 1A through 2, in response to the signal 107 transmitted by the monitoring system 102, an energy transfer system 112 of a fuel cell control system 108 may establish a readiness state within a set of readiness parameters by transferring energy (e.g., thermal or electrical) from a portion of the nuclear reactor system 104 (e.g., portion of the coolant system of the nuclear reactor system 104) to a portion of the fuel cell system 110. FIG. 17 illustrates alternative embodiments of the example operational flow 1500 of FIG. 15. FIG. 17 illustrates example embodiments where the establishing operation 1520 may include at least one additional operation. Additional operations may include an operation 1702, an operation 1704, and/or an operation 1706. Further, the operation 1702 illustrates, responsive to the monitored characteristics of the nuclear reactor system, establishing a readiness state of a fuel cell system within a set of readiness parameters by transferring thermal energy from an energy source to a portion of the fuel cell system, the readiness parameters a function of a characteristic of the nuclear reactor system. For example, as shown in FIGS. 1A through 2, in response to the signal 107 transmitted from the monitoring system 102, a energy transfer system 112 of a fuel cell control system 108 may establish a readiness state within a set of readiness parameters by transferring thermal energy from an energy source 103 to a portion of the fuel cell system 110 (e.g., bipolar plates of one or more fuel cells). Further, the operation 1704 illustrates, responsive to the monitored characteristics of the nuclear reactor system, establishing a readiness state of a fuel cell system within a set of readiness parameters by transferring thermal energy from an energy source to a portion of the fuel cell system using a heat transfer system, the readiness parameters a function of a characteristic of the nuclear reactor system. For example, as shown in FIGS. 1A through 2, in response to the signal 107 transmitted by the monitoring system 107, a heat transfer system 146 of a fuel cell control system 108 may establish a readiness state within a set of readiness parameters by transferring thermal energy from an energy source 103 to a portion of the fuel cell system 110 (e.g., condition system or a portion of one or more fuel cells). Further, the operation 1706 illustrates, responsive to the monitored characteristics of the nuclear reactor system, establishing a readiness state of a fuel cell system within a set of readiness parameters by transferring thermal energy from an energy source to a conditioning system of the fuel cell system using a heat transfer system, the readiness parameters a function of a characteristic of the nuclear reactor system. For example, as shown in FIGS. 1A through 2, in response to the signal 107 transmitted by the monitoring system 102, a heat transfer system 146 of a fuel cell control system 108 may establish a readiness state within a set of readiness parameters by transferring thermal energy from an energy source 103 to a humidity control system 142 of the fuel cell system 110. FIG. 18 illustrates alternative embodiments of the example operational flow 1500 of FIG. 15. FIG. 18 illustrates example embodiments where the establishing operation 1520 may include at least one additional operation. Additional operations may include an operation 1802, an operation 1804, and/or an operation 1806. Further, the operation 1802 illustrates, responsive to the monitored characteristics of the nuclear reactor system, establishing a readiness state of a fuel cell system within a set of readiness parameters by transferring electrical energy from an energy source to a portion of the fuel cell system, the readiness parameters a function of a characteristic of the nuclear reactor system. For example, as shown in FIGS. 1A through 2, in response to the signal 107 from the monitoring system 102, an energy transfer system 112 of a fuel cell control system 108 may establish a readiness state within a set of readiness parameters by transferring electrical energy from an energy source 103 to a temperature control system 144 of the fuel cell system 110. Further, the operation 1804 illustrates, responsive to the monitored characteristics of the nuclear reactor system, establishing a readiness state of a fuel cell system within a set of readiness parameters by transferring electrical energy from an energy source to a portion of the fuel cell system using an electrical energy transfer system, the readiness parameters a function of a characteristic of the nuclear reactor system. For example, as shown in FIGS. 1A through 2, an electrical energy transfer system 148 of a fuel cell control system 108 may establish a readiness state within a set of readiness parameters by transferring electrical energy from an energy source 103 to a temperature control system 144 of the fuel cell system 110. Further, the operation 1806 illustrates, responsive to the monitored characteristics of the nuclear reactor system, establishing a readiness state of a fuel cell system within a set of readiness parameters by transferring electrical energy from an energy source to a portion of the fuel cell system using an electrical-to-thermal energy conversion system, the readiness parameters a function of a characteristic of the nuclear reactor system. For example, as shown in FIGS. 1A through 2, in response to the signal 107 transmitted by the monitoring system 102, an electrical-to-thermal conversion system 150 of a fuel cell control system 108 may establish a readiness state within a set of readiness parameters by transferring electrical energy from an energy source 103 to a portion (e.g., one or more fuel cells) of the fuel cell system 110. FIG. 19 illustrates alternative embodiments of the example operational flow 1500 of FIG. 15. FIG. 19 illustrates example embodiments where the establishing operation 1520 may include at least one additional operation. Additional operations may include an operation 1902 and/or an operation 1906. The operation 1902 illustrates, responsive to the monitored characteristics of the nuclear reactor system, establishing a readiness state of a fuel cell system associated with the nuclear reactor system within a set of readiness parameters by adjusting a condition of at least one reactant of the fuel cell system, the readiness parameters a function of a characteristic of the nuclear reactor system. For example, as shown in FIGS. 1A through 2, in response to the signal 107 transmitted by the monitoring system 102, the reactant control system 114 of the fuel cell control system 108 may establish a readiness state of a fuel cell system associated with a nuclear reactor system 104 within a set of readiness parameters by adjusting a condition (e.g., pressure of reactant gas or flow rate of reactant gas) of at least one reactant (e.g., fuel or oxidant) of the fuel cell system 110. Further, the reactant pump control system 156 of the reactant control system 114 of the fuel cell control system 108 may establish a readiness state of a fuel cell system 110 associated with a nuclear reactor system 104 within a set of readiness parameters by adjusting a condition of at least one reactant of the fuel cell system 110. The operation 1904 illustrates, responsive to the monitored characteristics of the nuclear reactor system, establishing a readiness state of a fuel cell system associated with the nuclear reactor system within a set of readiness parameters by reconfiguring a portion of an electrical configuration of the fuel cell system, the readiness parameters a function of a characteristic of the nuclear reactor system. For example, as shown in FIGS. 1A through 2, in response to the signal 107 transmitted by the monitoring system 102, the configuration control system 116 (e.g., switching circuitry) of the fuel cell control system 108 may establish a readiness state of a fuel cell system 110 associated with a nuclear reactor system 104 within a set of readiness parameters by reconfiguring an electrical configuration (e.g., circuit arrangement) of the fuel cell system 110. FIG. 20 illustrates alternative embodiments of the example operational flow 1500 of FIG. 15. FIG. 20 illustrates example embodiments where the establishing operation 1520 may include at least one additional operation. Additional operations may include an operation 2002, an operation 2004, and/or an operation 2006. The operation 2002 illustrates, responsive to the monitored characteristics of the nuclear reactor system, establishing an electrical output level of a fuel cell system within an acceptable electrical output range, the acceptable electrical output range a function of a characteristic of the nuclear reactor system. For example, as shown in FIGS. 1A through 2, in response to the signal 107 transmitted by the monitoring system 102, the fuel cell control system 108 may establish an electrical output level of a fuel cell system 110 associated with a nuclear reactor system 104 within acceptable electrical output range, the acceptable electrical output range a function of a characteristic of the nuclear reactor system 104. For instance, the fuel cell control system 108 may transfer thermal energy (via the heat transfer system) to the fuel cell system 110 in order to heat one or more of the fuel cells of the fuel cell system 110 so as to establish an electrical output level of the fuel cell system 110 within in an acceptable output range. Further, the operation 2004 illustrates, responsive to the monitored characteristics of the nuclear reactor system, establishing an electrical current output level of a fuel cell system within an acceptable electrical current output range, the acceptable electrical current output range a function of a characteristic of the nuclear reactor system. For example, as shown in FIGS. 1A through 2, in response to the signal 107 transmitted by the monitoring system 102, the fuel cell control system 108 may establish an electrical current output level of a fuel cell system 110 associated with a nuclear reactor system 104 within an acceptable electrical current output range, the acceptable electrical current output range a function of a characteristic of the nuclear reactor system 104. For instance, the configuration control system 116 of the fuel cell control system 108 may reconfigure (e.g., decouple parallel coupled fuel cells and recouple them in a serial configuration or vice-versa) the electrical coupling configuration of two or more fuel cells of the fuel cell system 110 in order to establish an electrical current output level of the fuel cell system within in an acceptable output range. Further, the operation 2006 illustrates, responsive to the monitored characteristics of the nuclear reactor system, establishing a voltage level of a fuel cell system within an acceptable voltage range, the acceptable voltage range a function of a characteristic of the nuclear reactor system. For example, as shown in FIGS. 1A through 2, in response to the signal 107 transmitted from the monitoring system 102, the fuel cell control system 108 may establish an electrical voltage output level of a fuel cell system 110 associated with a nuclear reactor system 104 within acceptable electrical voltage output range, the acceptable electrical voltage output range a function of a characteristic of the nuclear reactor system 104. For instance, the configuration control system 116 of the fuel cell control system 108 may reconfigure (e.g., decouple parallel coupled fuel cells and recouple them in a serial configuration or vice-versa) the electrical coupling configuration of two or more fuel cells of the fuel cell system 110 in order to establish an electrical voltage output level of the fuel cell system 110 within in an acceptable output range. FIG. 21 illustrates alternative embodiments of the example operational flow 1500 of FIG. 15. FIG. 21 illustrates example embodiments where the establishing operation 1520 may include at least one additional operation. Additional operations may include an operation 2102, an operation 2104, and/or an operation 2106. The operation 2102 illustrates, responsive to the monitored characteristics of the nuclear reactor system, establishing a temperature of a portion of a fuel cell system within an acceptable temperature range, the acceptable temperature range a function of a characteristic of the nuclear reactor system. For example, as shown in FIGS. 1A through 2, in response to the signal 107 transmitted from the monitoring system 102, the fuel cell control system 108 may establish a temperature in a portion of a fuel cell system 110 associated with a nuclear reactor system 104 within acceptable temperature range, the acceptable temperature range a function of a characteristic of the nuclear reactor system 104. For instance, the energy transfer system 112 of the fuel cell control system 108 may transfer energy (e.g., thermal or electrical) from an energy source 103 to the fuel cell system 110 in order to heat or cool one or more of the fuel cells of the fuel cell system 110 so as to establish a temperature of the fuel cell system 110 within in an acceptable temperature range. The operation 2104 illustrates, responsive to the monitored characteristics of the nuclear reactor system, establishing a pressure in a portion of a fuel cell system within an acceptable pressure range, the acceptable pressure range a function of a characteristic of the nuclear reactor system. For example, as shown in FIGS. 1A through 2, in response to the signal 107 transmitted from the monitoring system 102, the fuel cell control system 108 may establish a pressure in a portion of a fuel cell system 110 associated with a nuclear reactor system 104 within acceptable pressure range, the acceptable pressure range a function of a characteristic of the nuclear reactor system 104. For instance, the energy transfer system 112 of the fuel cell control system 108 may transfer energy (e.g., thermal or electrical) from an energy source 103 to the fuel cell system 110 in order to establish a pressure in the fuel cell system 110 within in an acceptable pressure range. The operation 2106 illustrates, responsive to the monitored characteristics of the nuclear reactor system, establishing a humidity level in a fuel cell system within an acceptable humidity range, the acceptable humidity range a function of a characteristic of the nuclear reactor system. For example, as shown in FIGS. 1A through 2, in response to the signal 107 transmitted by the monitoring system 102, the fuel cell control system 108 may establish a humidity level in a portion of a fuel cell system 110 associated with a nuclear reactor system 104 within acceptable humidity range, the acceptable humidity range a function of a characteristic of the nuclear reactor system 104. For instance, the energy transfer system 112 of the fuel cell control system 108 may transfer energy (e.g., thermal or electrical) from an energy source 103 to the humidity control system of the fuel cell system 110 in order to establish a humidity level in the fuel cell system 110 within in an acceptable humidity range. FIG. 22 illustrates alternative embodiments of the example operational flow 1500 of FIG. 15. FIG. 21 illustrates example embodiments where the establishing operation 1520 may include at least one additional operation. Additional operations may include an operation 2202, an operation 2204, and/or an operation 2206. The operation 2202 illustrates, responsive to the monitored characteristics of the nuclear reactor system, establishing a temperature of a reactant stream of a fuel cell system within an acceptable temperature range, the acceptable temperature range a function of a characteristic of the nuclear reactor system. For example, as shown in FIGS. 1A through 2, in response to the signal 107 transmitted by the monitoring system 102, the fuel cell control system 108 may establish a temperature of a reactant stream (e.g., fuel stream or oxidant stream) of a fuel cell system 110 associated with a nuclear reactor system 104 within acceptable temperature range, the acceptable temperature range a function of a characteristic of the nuclear reactor system 104. For instance, the energy transfer system 112 of the fuel cell control system 108 may transfer energy (e.g., thermal or electrical) from an energy source 103 to the reactant conditioning system of the fuel cell system 110 in order to heat or cool one or more of the reactants of the fuel cell system 110 so as to establish a temperature of one or both of the reactant streams of the fuel cell system 110 within in an acceptable temperature range. The operation 2204 illustrates, responsive to the monitored characteristics of the nuclear reactor system, establishing a pressure in a reactant stream of a fuel cell system within an acceptable pressure range, the acceptable pressure range a function of a characteristic of the nuclear reactor system. For example, as shown in FIGS. 1A through 2, in response to the signal 107 transmitted by the monitoring system 102, the fuel cell control system 108 may establish a pressure in a reactant stream (e.g., fuel stream or oxidant stream) of a fuel cell system 110 associated with a nuclear reactor system 104 within an acceptable pressure range, the acceptable pressure range a function of a characteristic of the nuclear reactor system 104. For instance, the reactant control system 114 of the fuel cell control system 108 may control reactant valves and/or pumps of the fuel cell system 110 in order to increase or decrease the flow of one or more of the reactant streams of the fuel cell system 110 so as to establish a pressure of one or both of the reactant streams of the fuel cell system 110 within in an acceptable pressure range. The operation 2206 illustrates, responsive to the monitored characteristics of the nuclear reactor system, establishing a humidity level of a reactant stream of a fuel cell system within an acceptable humidity range, the acceptable humidity range a function of a characteristic of the nuclear reactor system. For example, as shown in FIGS. 1A through 2, in response to the signal 107 transmitted by the monitoring system 102, the fuel cell control system 108 may establish a humidity level of a reactant stream (e.g., fuel stream or oxidant stream) of a fuel cell system 110 associated with a nuclear reactor system 104 within acceptable humidity range, the acceptable humidity range a function of a characteristic of the nuclear reactor system 104. For instance, the energy transfer system 112 of the fuel cell control system 108 may transfer energy (e.g., thermal or electrical) from an energy source 103 to the reactant conditioning system, such as a humidifier, of the fuel cell system 110 in order to establish a humidity level of one or both of the reactant streams of the fuel cell system 110 within in an acceptable humidity range. FIG. 23 illustrates alternative embodiments of the example operational flow 1500 of FIG. 15. FIG. 23 illustrates example embodiments where the establishing operation 1520 may include at least one additional operation. Additional operations may include an operation 2302, an operation 2304, an operation 2306, and/or an operation 2308. The operation 2302 illustrates, responsive to the monitored characteristics of the nuclear reactor system, establishing a readiness state of a polymer electrolyte membrane fuel cell system within a set of readiness parameters, the readiness parameters a function of a characteristic of the nuclear reactor system. For example, as shown in FIGS. 1A through 2, in response to the signal 107 from the monitoring system 102, a fuel cell control system 108 may establish a readiness state of a polymer electrolyte membrane fuel cell system 182 within a set of readiness parameters, wherein the readiness parameters are a function of one or more characteristics of the nuclear reactor system 104. The operation 2304 illustrates, responsive to the monitored characteristics of the nuclear reactor system, establishing a readiness state of a solid oxide fuel cell system within a set of readiness parameters, the readiness parameters a function of a characteristic of the nuclear reactor system. For example, as shown in FIGS. 1A through 2, in response to the signal 107 from the monitoring system 102, a fuel cell control system 108 may establish a readiness state of a solid oxide fuel cell system 183 within a set of readiness parameters, wherein the readiness parameters are a function of one or more characteristics of the nuclear reactor system 104. The operation 2306 illustrates, responsive to the monitored characteristics of the nuclear reactor system, establishing a readiness state of an alkaline fuel cell system within a set of readiness parameters, the readiness parameters a function of a characteristic of the nuclear reactor system. For example, as shown in FIGS. 1A through 2, in response to the signal 107 from the monitoring system 102, a fuel cell control system 108 may establish a readiness state of an alkaline fuel cell system 184 within a set of readiness parameters, wherein the readiness parameters are a function of one or more characteristics of the nuclear reactor system 104. The operation 2308 illustrates, responsive to the monitored characteristics of the nuclear reactor system, establishing a readiness state of a molten carbonate fuel cell system within a set of readiness parameters, the readiness parameters a function of a characteristic of the nuclear reactor system. For example, as shown in FIGS. 1A through 2, in response to the signal 107 from the monitoring system 102, a fuel cell control system 108 may establish a readiness state of a molten carbonate fuel cell system 185 within a set of readiness parameters, wherein the readiness parameters are a function of one or more characteristics of the nuclear reactor system 104. FIG. 24 illustrates alternative embodiments of the example operational flow 1500 of FIG. 15. FIG. 24 illustrates example embodiments where the monitoring operation 1510 may include at least one additional operation. Additional operations may include an operation 2402, an operation 2404, and/or an operation 2406. The operation 2402 illustrates monitoring characteristics of a nuclear reactor system using a nuclear reactor monitoring system. For example, as shown in FIGS. 1A through 2, the monitoring system 102 may monitor one or more characteristics, such as an operation characteristic or a design characteristic, of the nuclear reactor system 104. Further, the operation 2404 illustrates transmitting a signal from the nuclear reactor monitoring system to a computer data management system. For example, as shown in FIGS. 1A through 2, upon monitoring a characteristic of the nuclear reactor system 104, the monitoring system 102 may transmit a signal indicative of the monitored characteristic to a computer data management system (e.g., a computer system configured to archive and analyze monitored characteristic data). Further, the operation 2406 illustrates transmitting a signal from the nuclear reactor monitoring system to a fuel cell control system. For example, as shown in FIGS. 1A through 2, upon monitoring a characteristic of the nuclear reactor system 104, the monitoring system 102 may transmit a signal indicative of the monitored characteristic to the fuel cell control system 108. For instance, the monitoring system 102 may transmit a signal indicative of the monitored characteristic to the fuel cell control module 109 of the fuel cell control system 108. FIG. 25 illustrates alternative embodiments of the example operational flow 1500 of FIG. 15. FIG. 25 illustrates example embodiments where the monitoring operation 1510 may include at least one additional operation. Additional operations may include an operation 2502, an operation 2504, an operation 2506, an operation 2508, and operation 2510, and operation 2512, and/or an operation 2514. The operation 2502 illustrates monitoring an operational characteristic of a nuclear reactor system. For example, as shown in FIGS. 1A through 2, a monitoring system 178 configured to monitor an operational characteristic of the nuclear reactor system may monitor one or more operational characteristics of the nuclear reactor system 104, such as temperature or pressure of a portion (e.g., coolant fluid of a coolant loop) of the nuclear reactor system 104. Further, the operation 2504 illustrates monitoring an operational characteristic of the nuclear reactor core nuclear reactor system. For example, as shown in FIGS. 1A through 2, a monitoring system 178 configured to monitor an operational characteristic of the nuclear reactor system may monitor one or more operational characteristics of the nuclear reactor core of the nuclear reactor system 104, such as temperature, pressure, or void fraction of the nuclear reactor core. The operation 2506 illustrates monitoring a design characteristic of a nuclear reactor system. For example, as shown in FIGS. 1A through 2, a monitoring system 179 configured to monitor a design characteristic of the nuclear reactor system may monitor one or more design characteristics of the nuclear reactor system 104. Further, the operation 2508 illustrates monitoring the responsiveness of a safety system of a nuclear reactor system to a design basis accident. For example, as shown in FIGS. 1A through 2, a monitoring system 179 configured to monitor a design characteristic of the nuclear reactor system may monitor the responsiveness of a safety system of a nuclear reactor system to a design basis accident, such as guillotine break. Further, the operation 2510 illustrates monitoring the responsiveness of a safety system of a nuclear reactor system to a design basis accident. For example, as shown in FIGS. 1A through 2, a monitoring system 179 configured to monitor a design characteristic of the nuclear reactor system may monitor the time required for a fuel element, such as a fuel pin assembly or a collection of fuel pin assemblies, of a nuclear reactor system to reach a specified temperature upon loss of coolant flow. The operation 2512 illustrates monitoring a characteristic of an operation system of a nuclear reactor system. For example, as shown in FIGS. 1A through 2, a monitoring system 180 configured to monitor a characteristic of an operation system of the nuclear reactor system may monitor one or more characteristics of an operation system (e.g., coolant system, safety system, shutdown system, or warning system) of the nuclear reactor system 104. Further, the operation 2514 illustrates monitoring a signal transmitted by an operation system of a nuclear reactor system. For example, as shown in FIGS. 1A through 2, a monitoring system 180 configured to monitor a characteristic of an operation system of the nuclear reactor system may monitor one or more signals transmitted from an operation system (e.g., coolant system, safety system, shutdown system, or warning system) of the nuclear reactor system 104. For instance, the monitoring system 180 configured to monitor a characteristic of an operation system of the nuclear reactor system may monitor a digital signal transmitted by a safety system of the nuclear reactor system 104. FIG. 26 illustrates an operational flow 2600 representing example operations related to establishing a readiness state in a fuel cell backup system of a nuclear reactor system. FIG. 26 illustrates an example embodiment where the example operational flow 1500 of FIG. 15 may include at least one additional operation. Additional operations may include an operation 2610, and/or an operation 2612. After a start operation, a monitoring operation 1510, and an establishing operation 1520, the operational flow 2600 moves to a transferring operation 2610. The transferring operation 2610 illustrates transferring electrical energy from a fuel cell system to an operation system of the nuclear reactor system. For example, as shown in FIGS. 1A through 2, an energy supply system 191 may transfer electrical energy from the electrical output of the fuel cell system 110 to an operation system (e.g., coolant system or shutdown system) of the nuclear reactor system 104 The operation 2612 illustrates, responsive to at least one condition, transferring electrical energy from the fuel cell system to an operation system of the nuclear reactor system. For example, as shown in FIGS. 1A through 2, an energy supply system 191 may transfer electrical energy from the electrical output of the fuel cell system 110 to an operation system (e.g., coolant system or shutdown system) of the nuclear reactor system in response to a condition, such as a signal from an operation system of the nuclear reactor system 104, or a shutdown event of the nuclear reactor system 104. FIG. 27 illustrates an operational flow 2700 representing example operations related to establishing a readiness state in a fuel cell backup system of a nuclear reactor system. FIG. 27 illustrates an example embodiment where the example operational flow 1500 of FIG. 15 may include at least one additional operation. Additional operations may include an operation 2710 and/or operation 2712. After a start operation, a monitoring operation 1510, and an establishing operation 1520, the operational flow 2700 moves to a modifying operation 2710. The modifying operation 2710 illustrates modifying an electrical output of the fuel cell system. For example, as shown in FIGS. 1A through 2, the output modification system 194 may modify the characteristics of the electrical output of the fuel cell system 110. Further, the operation 2712 illustrates modifying an electrical output of the fuel cell system using power management circuitry. For example, as shown in FIGS. 1A through 2, power management circuitry 195 (e.g., voltage regulation circuitry) may modify the electrical characteristics of the electrical output of the fuel cell system 110. FIG. 28 illustrates alternative embodiments of the example operational flow 2700 of FIG. 27. FIG. 28 illustrates example embodiments where the modifying operation 2710 may include at least one additional operation. Additional operations may include an operation 2810, and/or an operation 2812. The operation 2810 illustrates modifying an electrical output of the fuel cell system by adjusting the electrical output of at least one fuel cell of the fuel cell system using control circuitry. For example, as shown in FIGS. 1A through 2, control circuitry 196 may modify the characteristics of the electrical output of the fuel cell system 110 by adjusting the electrical output of one or more fuel cells of the fuel cell system. Further, the operation 3012 illustrates simulating an A.C. electrical output of the fuel cell system by sequentially staging the D.C. output of at least two fuel cells of the fuel cell system. For example, as shown in FIGS. 1A through 2, control circuitry 196 may include solid state switches configured to simulate an A.C. electrical output of the fuel cell system 110 by sequentially staging the D.C. electrical outputs of two or more fuel cells of the fuel cell system 110. Those having skill in the art will recognize that the state of the art has progressed to the point where there is little distinction left between hardware, software, and/or firmware implementations of aspects of systems; the use of hardware, software, and/or firmware is generally (but not always, in that in certain contexts the choice between hardware and software can become significant) a design choice representing cost vs. efficiency tradeoffs. Those having skill in the art will appreciate that there are various vehicles by which processes and/or systems and/or other technologies described herein can be effected (e.g., hardware, software, and/or firmware), and that the preferred vehicle will vary with the context in which the processes and/or systems and/or other technologies are deployed. For example, if an implementer determines that speed and accuracy are paramount, the implementer may opt for a mainly hardware and/or firmware vehicle; alternatively, if flexibility is paramount, the implementer may opt for a mainly software implementation; or, yet again alternatively, the implementer may opt for some combination of hardware, software, and/or firmware. Hence, there are several possible vehicles by which the processes and/or devices and/or other technologies described herein may be effected, none of which is inherently superior to the other in that any vehicle to be utilized is a choice dependent upon the context in which the vehicle will be deployed and the specific concerns (e.g., speed, flexibility, or predictability) of the implementer, any of which may vary. Those skilled in the art will recognize that optical aspects of implementations will typically employ optically-oriented hardware, software, and or firmware. In some implementations described herein, logic and similar implementations may include software or other control structures. Electronic circuitry, for example, may have one or more paths of electrical current constructed and arranged to implement various functions as described herein. In some implementations, one or more media may be configured to bear a device-detectable implementation when such media hold or transmit device-detectable instructions operable to perform as described herein. In some variants, for example, implementations may include an update or modification of existing software or firmware, or of gate arrays or programmable hardware, such as by performing a reception of or a transmission of one or more instructions in relation to one or more operations described herein. Alternatively or additionally, in some variants, an implementation may include special-purpose hardware, software, firmware components, and/or general-purpose components executing or otherwise invoking special-purpose components. Specifications or other implementations may be transmitted by one or more instances of tangible transmission media as described herein, optionally by packet transmission or otherwise by passing through distributed media at various times. Alternatively or additionally, implementations may include executing a special-purpose instruction sequence or invoking circuitry for enabling, triggering, coordinating, requesting, or otherwise causing one or more occurrences of virtually any functional operations described herein. In some variants, operational or other logical descriptions herein may be expressed as source code and compiled or otherwise invoked as an executable instruction sequence. In some contexts, for example, implementations may be provided, in whole or in part, by source code, such as C++, or other code sequences. In other implementations, source or other code implementation, using commercially available and/or techniques in the art, may be compiled/implemented/translated/converted into a high-level descriptor language (e.g., initially implementing described technologies in C or C++ programming language and thereafter converting the programming language implementation into a logic-synthesizable language implementation, a hardware description language implementation, a hardware design simulation implementation, and/or other such similar mode(s) of expression). For example, some or all of a logical expression (e.g., computer programming language implementation) may be manifested as a Verilog-type hardware description (e.g., via Hardware Description Language (HDL) and/or Very High Speed Integrated Circuit Hardware Descriptor Language (VHDL)) or other circuitry model which may then be used to create a physical implementation having hardware (e.g., an Application Specific Integrated Circuit). Those skilled in the art will recognize how to obtain, configure, and optimize suitable transmission or computational elements, material supplies, actuators, or other structures in light of these teachings. The foregoing detailed description has set forth various embodiments of the devices and/or processes via the use of block diagrams, flowcharts, and/or examples. Insofar as such block diagrams, flowcharts, and/or examples contain one or more functions and/or operations, it will be understood by those within the art that each function and/or operation within such block diagrams, flowcharts, or examples can be implemented, individually and/or collectively, by a wide range of hardware, software, firmware, or virtually any combination thereof. In one embodiment, several portions of the subject matter described herein may be implemented via Application Specific Integrated Circuits (ASICs), Field Programmable Gate Arrays (FPGAs), digital signal processors (DSPs), or other integrated formats. However, those skilled in the art will recognize that some aspects of the embodiments disclosed herein, in whole or in part, can be equivalently implemented in integrated circuits, as one or more computer programs running on one or more computers (e.g., as one or more programs running on one or more computer systems), as one or more programs running on one or more processors (e.g., as one or more programs running on one or more microprocessors), as firmware, or as virtually any combination thereof, and that designing the circuitry and/or writing the code for the software and or firmware would be well within the skill of one of skill in the art in light of this disclosure. In addition, those skilled in the art will appreciate that the mechanisms of the subject matter described herein are capable of being distributed as a program product in a variety of forms, and that an illustrative embodiment of the subject matter described herein applies regardless of the particular type of signal bearing medium used to actually carry out the distribution. Examples of a signal bearing medium include, but are not limited to, the following: a recordable type medium such as a floppy disk, a hard disk drive, a Compact Disc (CD), a Digital Video Disk (DVD), a digital tape, a computer memory, etc.; and a transmission type medium such as a digital and/or an analog communication medium (e.g., a fiber optic cable, a waveguide, a wired communications link, a wireless communication link (e.g., transmitter, receiver, transmission logic, reception logic, etc.), etc.). In a general sense, those skilled in the art will recognize that the various embodiments described herein can be implemented, individually and/or collectively, by various types of electro-mechanical systems having a wide range of electrical components such as hardware, software, firmware, and/or virtually any combination thereof; and a wide range of components that may impart mechanical force or motion such as rigid bodies, spring or torsional bodies, hydraulics, electro-magnetically actuated devices, and/or virtually any combination thereof. Consequently, as used herein “electro-mechanical system” includes, but is not limited to, electrical circuitry operably coupled with a transducer (e.g., an actuator, a motor, a piezoelectric crystal, a Micro Electro Mechanical System (MEMS), etc.), electrical circuitry having at least one discrete electrical circuit, electrical circuitry having at least one integrated circuit, electrical circuitry having at least one application specific integrated circuit, electrical circuitry forming a general purpose computing device configured by a computer program (e.g., a general purpose computer configured by a computer program which at least partially carries out processes and/or devices described herein, or a microprocessor configured by a computer program which at least partially carries out processes and/or devices described herein), electrical circuitry forming a memory device (e.g., forms of memory (e.g., random access, flash, read only, etc.)), electrical circuitry forming a communications device (e.g., a modem, communications switch, optical-electrical equipment, etc.), and/or any non-electrical analog thereto, such as optical or other analogs. Those skilled in the art will also appreciate that examples of electro-mechanical systems include but are not limited to a variety of consumer electronics systems, medical devices, as well as other systems such as motorized transport systems, factory automation systems, security systems, and/or communication/computing systems. Those skilled in the art will recognize that electro-mechanical as used herein is not necessarily limited to a system that has both electrical and mechanical actuation except as context may dictate otherwise. In a general sense, those skilled in the art will recognize that the various aspects described herein which can be implemented, individually and/or collectively, by a wide range of hardware, software, firmware, and/or any combination thereof can be viewed as being composed of various types of “electrical circuitry.” Consequently, as used herein “electrical circuitry” includes, but is not limited to, electrical circuitry having at least one discrete electrical circuit, electrical circuitry having at least one integrated circuit, electrical circuitry having at least one application specific integrated circuit, electrical circuitry forming a general purpose computing device configured by a computer program (e.g., a general purpose computer configured by a computer program which at least partially carries out processes and/or devices described herein, or a microprocessor configured by a computer program which at least partially carries out processes and/or devices described herein), electrical circuitry forming a memory device (e.g., forms of memory (e.g., random access, flash, read only, etc.)), and/or electrical circuitry forming a communications device (e.g., a modem, communications switch, optical-electrical equipment, etc.). Those having skill in the art will recognize that the subject matter described herein may be implemented in an analog or digital fashion or some combination thereof. Those skilled in the art will recognize that at least a portion of the devices and/or processes described herein can be integrated into a data processing system. Those having skill in the art will recognize that a data processing system generally includes one or more of a system unit housing, a video display device, memory such as volatile or non-volatile memory, processors such as microprocessors or digital signal processors, computational entities such as operating systems, drivers, graphical user interfaces, and applications programs, one or more interaction devices (e.g., a touch pad, a touch screen, an antenna, etc.), and/or control systems including feedback loops and control motors (e.g., feedback for sensing position and/or velocity; control motors for moving and/or adjusting components and/or quantities). A data processing system may be implemented utilizing suitable commercially available components, such as those typically found in data computing/communication and/or network computing/communication systems. 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. Although a user is shown/described herein as a single illustrated figure, those skilled in the art will appreciate that the user may be representative of a human user, a robotic user (e.g., computational entity), and/or substantially any combination thereof (e.g., a user may be assisted by one or more robotic agents) unless context dictates otherwise. Those skilled in the art will appreciate that, in general, the same may be said of “sender” and/or other entity-oriented terms as such terms are used herein unless context dictates otherwise. 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,” “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.
	summary
039490270	summary	This invention relates to nuclear fuel and more particularly to ceramic fuel tablets or pellets used in nuclear reactors. These ceramic pellets generally consist of uranium, plutonium, thorium, etc.. in oxide, nitride, carbide or other form. The manufacturing process for these pellets comprises the compression of appropriate powder in a cylindrical mould, followed by sintering to confer to the pellets the necessary characteristics, such as mechanical resistance and density. The sintering cycle however gives rise to shrinkage of the raw material, which may attain about 50 % of the green volume. Thus the sintering of the pellet influences its final form; more particularly, the pellet will not shrink uniformly on sintering and will shrink more in the less dense areas if the density of the raw pellet is not homogeneous after compression. The peculiar friction conditions on compression do not allow a homogeneously dense pellet to be obtained. Therefore the pellets are at present rectified after sintering, in order to correct their form, so that the final product obtained is properly cylindrical. The diametrical tolerances set by the users of nuclear fuel pellets are generally very strict, so that rectification, mostly effected by grinding, can hardly be avoided. It has been noted that density differences in the raw pellets on compression are a function of the relative movement of the punches in the compaction chamber. Indeed, when using a double-acting press, i.e. compressing by simultaneous shifting of the upper and lower punches, the lateral surface of the obtained cylindrical pellet presents, after sintering, an inwardly incurved configuration or "diabolo" configuration. If, on the other hand, one uses a single-acting pelletizing machine, i.e. with only one punch moving on compression, the obtained pellet will, after sintering, present a configuration resembling a cone trunk, the broad base of which corresponds to the part of the pellet in contact with the mobile punch. In order to make these pellets comply with the specifications, their "diabolo" or conical form must be rectified by grinding their lateral side so as to obtain a cylindrical surface. The invention proposes an improved compression phase, allowing avoidance of rectification of the pellets after sintering. According to the invention, the pellet is compressed solely by shifting the lower punch of a pelletizing machine in a compaction chamber of which at least the upper part of the walls is inclined in the direction of a diameter increase towards the die-bearing table, on a height corresponding at least to the height of the compressed pellet. The inclination angle, i.e. the angle formed by the vertical axis of the die and the intersection line between an axial plan and the inclined wall, will generally lie between 1.degree.30' and 0.degree.6' as a function of the compaction conditions, the diameter of the pellets and the nature of the powder. The walls are preferably inclined on their whole height, but it is evident that, if the pellet to be compressed is rather high and the compression ratio rather elevated, the walls may have a lower cylindrical part and an upper inclined part. Generally it is enough to have walls inclined on a height corresponding to the tablet to be compressed to obtain the advantages according to the invention. The inclined part should then be situated on the upper part of the wall. A pellet, compressed according to the invention, will thus have the configuration of a cone trunk, the large base of which will be the less dense part. After sintering, this pellet will generally have a substantially cylindrical form and rectification will be superfluous.
	description	This application claims priority to, and incorporates by reference, U.S. Provisional Patent Application Ser. No. 60/428,165 which was filed Nov. 21, 2002. 1. Field of the Invention The invention relates generally to the field of inspection systems. More particularly, the invention relates to a remote sensing device to detect materials of varying atomic numbers. 2. Discussion of the Related Art The identification of weapons-grade materials (such as uranium, plutonium, or radiation dispersion devices known as “dirty bombs”) concealed within cargo containers is of growing importance worldwide. Typically, instruments such as Geiger counters and gamma ray detectors are employed at ports-of-entry to scan such containers. Nevertheless, these technologies have limited applications. For example, highly-enriched uranium (235U) does not emit a significant flux of gamma rays, and can be easily shielded by a thin layer of lead. Meanwhile, it is known that by measuring photon attenuation, one can identify materials with large atomic numbers. In order to accurately interrogate a cargo container, a high-energy beam of photons with high penetrating power may be used. Further, a detection system that can identify materials of varying atomic number is needed. Until now, the requirements of a method and/or apparatus for probing closed containers for weapons-grade fissile materials of varying atomic number with a high-energy photon beam, and resolving the energy and attenuation of the outgoing flux of photons from the container has not been met. There is a need for the following embodiments. Of course, the invention is not limited to these embodiments. According to an aspect of the invention, a method for identifying a material includes casting an incident photon beam on the material and detecting an emerging photon beam with an array of fission-fragment detectors, a first set of scintillator paddles, and a second set of scintillator paddles, wherein the array of fission-fragment detectors, the first set of scintillator paddles, and the second set of scintillator paddles are sensitive to different ranges of photon beam energy. According to another aspect of the invention, a photon beam flux monitor for resolving a high-energy beam includes an array of fission-fragment detectors for measuring a first range of photon energies, a first set of scintillator paddles coupled to the array of fission-fragment detectors for measuring a second range of photon energies, a convertor coupled to the first set of scintillator paddles, and a second set of scintillator paddles coupled to the convertor for measuring a third range of photon energies. These, and other, embodiments of the invention will be better appreciated and understood when considered in conjunction with the following description and the accompanying drawings. It should be understood, however, that the following description, while indicating various embodiments of the invention and numerous specific details thereof, is given by way of illustration and not of limitation. Many substitutions, modifications, additions and/or rearrangements may be made within the scope of the invention without departing from the spirit thereof, and the invention includes all such substitutions, modifications, additions and/or rearrangements. The invention and the various features and advantageous details thereof are explained more fully with reference to the nonlimiting embodiments that are illustrated in the accompanying drawings and detailed in the following description. It should be understood that the detailed description and the specific examples, while indicating specific embodiments of the invention, are given by way of illustration only and not by way of limitation. Various substitutions, modifications, additions and/or rearrangements within the spirit and/or scope of the underlying inventive concept will become apparent to those of ordinary skill in the art from this disclosure. The invention includes a method and/or apparatus for identifying the presence of substances concealed in closed containers or inaccessible areas by using a beam of high-energy X-rays produced from an electron accelerator. The invention also includes a method and/or apparatus for measuring fluxes of transmitted photons in the regime of high energies, thereby determining the atomic number of the material on the photon beam path. Further, the invention can include an energy-sensitive photon beam flux monitor (BFM) to analyze properties of materials by measuring the energy-dependent attenuation of the transmitted beam of photons. In the passage of photons through matter, a photon interacts with atoms or nuclei in an energy-dependent way. Specifically, high atomic number (Z) materials tend to absorb higher energy photons, and low Z materials tend to absorb lower energy photons. The invention includes a method and/or apparatus for measuring the attenuation of a photon beam flux, therefore yielding a measure of the density and distribution of the interrogated material. The invention may be used to identify and distinguish high and low density materials concealed within a vessel, including weapons-grade materials such as, for example, uranium, plutonium, or radiation dispersion devices (known as “dirty bombs”). Further, the invention can include using a detector with a natural uranium target to measure the fission fragments induced from photons. In one embodiment, the detector has a high degree of photon-energy selectivity in the range of 10.0 to 20.0 MeV. In another embodiment, the invention includes a photon beam flux monitor including a detector for resolving photon energies up to about 6 MeV and another detector for resolving photon energies above about 6 MeV. In yet another embodiment, the invention includes using three detectors, each detector being sensitive to a different range of energies. These energy ranges may overlap. Referring to FIG. 1, a block diagram of a photon interrogation system 100 is depicted, representing an embodiment of the invention. An electron beam generator (accelerator) 105 directs a beam upon a radiator 110 to produce a photon beam through the process of bremsstrahlung. In one embodiment, the electron beam generator 105 produces a flux of about 107 photons per second. In another embodiment, the electron beam generator 105 produces a photon beam with energies between about 1 to 15 MeV. The radiator 110 may be, for example, a thin tungsten foil. The radiator 110 is coupled to an electron stopping block 115, which interrogates a cargo container 125 with an incident photon beam 120. A emerging photon beam 121 is monitored with a photon beam flux monitor 130. Still referring to FIG. 1, the photon beam flux monitor 130 includes three detection devices including an array of fission-fragment detectors (Parallel-Plate Avalanche Detectors or PPADs) 135 followed by two sets of scintillator paddles (telescopes) 140 and 150 with a convertor 145 in between, wherein each scintillator is sensitive to a different range of photon energies. In one embodiment, the convertor 145 is a lead (Pb) convertor. The first set of scintillator paddles 140 may detect materials of low atomic number (low Z) by resolving photon energies up to about 6 MeV, and the second set of scintillator paddles 150 may detect materials of high atomic number (high Z) by resolving photon energies exceeding about 6 MeV. In other embodiments, different energy ranges may be desirable. Depending upon complementary detection techniques and the desired penetration power of the photon beam 120, the electron beam energies of the emerging photon beam 121 may be as high as 50 MeV, and its energy distribution may range between 0 and 50 MeV with a characteristic 1/Eγ falloff (bremsstrahlung photons). Still referring to FIG. 1, the three sets of detectors 135, 140 and 150 can be used to measure the beam of photons 121 emerging from the cargo container 125. By resolving the energy of the beam 121, the effective density distribution of the matter within the container 125 may be revealed. Material concealed within the cargo container 125 may selectively absorb the various parts of the bremsstrahlung spectrum of the incident photon beam 120 depending upon its atomic number. The photon flux monitor 130 may register a drop in the emerging photon beam 121 intensity in the energy regime where the interrogated material has preferentially absorbed the photon beam. Still referring to FIG. 1, in one embodiment, low-Z detectors may be formed of a telescoping array of approximately 1 inch thick scintillator paddles 140, wherein a first layer blocks out charged particles. Each scintillator paddle may be instrumented on one end with a photomultiplier tube (PMT). The low-Z detector array may be segmented to minimize pile up of the signal. Low-Z materials such as water, chemical explosives, and plastic interact primarily with the lower energy portion of the emerging photon beam 121. The variation of the PMT current may give a measurement of the distribution of low-Z materials within the interrogated vessel 125. In one embodiment the scintillator paddles 140 are sensitive to photon energies less than about 6 MeV. In other embodiments, different energy ranges may be desirable. Still referring to FIG. 1, high-Z detectors may be formed of a grouping of thin scintillator paddles 150. Placed in front of these scintillators may be a thin lead-convertor foil 145 for producing electron/positron (e−/e+) pairs. When a photon strikes the convertor 145 (which may be, for example, a tungsten or lead foil), the photon converts into the electron/positron pair. In one embodiment, the thickness of the convertor 145 is between about 1% to 5% radiation lengths. Next, the electron and positron travel into the second set of scintillators 150, where they are detected. The e−/e+ pairs may be measured, for example, by placing a sweeping dipole magnet (not shown) in between the convertor 145 and a bilaterally-symmetric arrangement of the scintillator paddles 150. The e−/e+ pairs may also be measured by directly measuring the double ionization peak. In one embodiment, the scintillator paddles 150 are sensitive to photon energies exceeding about 6 MeV. In other embodiments, different energy ranges may be desirable. Still referring to FIG. 1, the array of fission-fragment detectors (PPADs) 135 may be ionization detectors that operate in the avalanche regime, which is defined by a combination of gas pressure and electric field such that a single free electron can start an exponential ionization process. Typical gas pressures vary from 1 Torr to about 25 Torr, while the corresponding electric field varies from about 100 V/mm to 400 V/mm. The array of fission-fragment detectors 135 may be tuned to the photofission cross section of the fissile material to be interrogated in container 125. In one embodiment, the array of fission-fragment detectors 135 is sensitive to photon energies in the range of about 10 to 20 MeV. In other embodiments, a different range of energies may be desirable. Referring to FIG. 2, a diagram of fission-fragment detector 200 (PPAD) is depicted according to one embodiment of the invention. The fission-fragment detector 200 may be used as an element of the array of fission-fragment detectors 135 detailed in FIG. 1. The fission-fragment detector 200 is a two-parallel-plate capacitor immersed in a gas at low pressure. A voltage is applied between the plates to establish the conditions for an avalanche regime to be generated across the gap. When a free electron is created inside the detector by an ionizing particle, it generates an avalanche of electron pairs. The number of avalanche electrons is proportional to the distance they travel. In order to minimize the probability of electric breakdown in the form of sparks and glow discharges when the PPAD 200 is in the avalanche regime, a gas with high self-quenching properties may be used such as, for example, isobutane. Referring to FIG. 3, an exploded view of the fission-fragment detector 200 detailed in FIG. 2 is depicted according to one sample embodiment of the invention. A target holder 205 including a photofission target 210 is coupled to a collimator 235. The collimator is coupled to an anode plane 215, and the anode plane 215 is coupled to a cathode plane 230. The anode plane 215 includes a grid of gold-plated tungsten/rhenium wires 220, and the cathode plane 230 includes an aluminized mylar foil 225. Both electrodes (anode 215 and cathode 230) may include rectangular frames of PC-Board material (such as fiberglass), with windows cut inside. The thickness of each frame may be approximately 1.5 mm, so that when placed back to back, they generate a gap between the electrodes 215 and 230 of approximately 3 mm. Part of the copper on the external side of the PC boards may be removed to provide space for connecting resistors, capacitors, high-voltage (HV) connectors, and signal connectors. The copper may also be removed from the edges of the windows to minimize the probability of electric breakdown along the limits of the active region. In order to further reduce the probability of electric breakdown, the cathode 230 window may be made approximately 10 mm×10 mm larger in area than the anode 215 window. Still referring to FIG. 3, the grid wires 220 may be attached to the anode frame 225 with epoxy glue. Exposed areas are covered with a layer of epoxy glue, and the high voltage connectors are encapsulated in plastic cases. The PPAD 200 may be tuned to uranium by having the target 210 made of, for example, a thin film of 238U deposited on one side of an approximately 100 μm thick aluminum foil. The target 210 may also be, for example, an approximately 178 micron thick film of 238U. In one embodiment, the invention includes using targets 210 of different materials to tune the PPAD 200 to a corresponding range of energies. The ability to tune the PPAD 200 allows detection of materials of varying atomic numbers. Still referring to FIG. 3, fission fragments are generally slow moving and have high charge, hence they may be readily stopped within thick targets. Typically, only fission fragments produced from the outer 5-um layer of the target 210 emerge, and the rest is absorbed within the target 210. Thick targets 210 may serve as a relative flux monitor, since the rate of photofission production in the outer layer scale with the intensity of the beam. The absolute flux can be calibrated with an empty vessel. In one embodiment, an absolute measurement may be made by having thin films of 238U sputtered onto an aluminum substrate. Still referring to FIG. 3, the target 210 is sandwiched between two frames with approximately 5 mm×10 mm windows and may be connected to the electrodes 215 and 230 by teflon screws, where the distance between the targets and the detectors is set by teflon spacers. The angles of the particles coming into the detectors electrodes 215 and 230 are constrained by the collimator 235. The collimator 235 may be made of fiberglass, approximately 1 mm thick, with a circular hole (approximately 40 mm in diameter) in the center. Referring to FIG. 4, a diagram of an array of fission-fragment detectors 400 (PPADs) is depicted according to one aspect of the invention. The array of fission-fragment detectors 400 may be used as the array 135 of FIG. 1. A pair of collimators 405 is coupled to a target surface 410 and to a fission-fragment detector 415 (PPAD). The PPAD 415 is coupled to a rail 420 through a holder 425. In one embodiment, the array of fission-fragment detectors 400 includes a plurality of target-detector assemblies. Still referring to FIG. 4, the array of fission-fragment detectors 400 may be operated in a low-pressure gas atmosphere such as, for example, isobutane, and placed inside a hermetically sealed reaction chamber. In one embodiment, in order to maintain the avalanche regime and to keep the gain of avalanche detectors constant, the pressure and purity of the gas is maintained stable by flowing the gas through the chamber using a pressure and flow control system. Referring to FIG. 5, a block diagram of a data acquisition and processing system 500 is depicted according to one exemplary embodiment of the invention. The data acquisition and processing system 500 may be used to read and process a signal from a detector (135, 140, or 150), detailed in FIG. 1. Still referring to FIG. 5, a detection device signal is amplified by a pre amp circuit 505. A quad linear fan-in fan-out circuit 510 takes the amplified device signal and generates four identical output signals with unit gain, while providing control over the polarity of these output signals. The quad linear fan-in fan-out circuit 510 is coupled to a time-to-digital converter circuit 515, an analog-to-digital converter circuit 520, and to a discriminator circuit 525. The discriminator circuit 525 outputs a digital NIM (nuclear instrumentation module) pulse when its input is above a threshold. In another embodiment, the discriminator circuit 525 outputs Fastbus pulses. The discriminator 525 is coupled to a scaler circuit 535 and to the first input of an AND gate 530. Still referring to FIG. 5, an accelerator signal, indicating whether the beam generator 105 detailed in FIG. 1 is in operation, is coupled to the second input of the AND gate 530. The output of the AND gate 530 is coupled to the analog-to-digital converter 520, the time-to-digital converter 515, and to the scaler circuit 535. The time-to-digital converter circuit 515, the analog-to-digital converter circuit 520, and the scaler circuit 535 are coupled to a standard bus backplane 540. In one embodiment, the backplane 540 is a VME (Versa Module Europe) backplane. The backplane 540 is coupled to a computer 545 including a data acquisition system board or system. The computer 545 is coupled to a program storage media 550. The program storage media 550 may be any type of readable memory including, for example, a magnetic or optical media such as a card, tape or disk, a semiconductor memory such as a PROM or FLASH memory, or any other available media. In one embodiment, three readout logic circuits such as the one detailed in FIG. 5 may be used to process each signal from the three detectors 135, 140, and 150 detailed in FIG. 1. The data acquisition system 500 may be used to collect signals from each device of the photon interrogation system 100. Each of the three measurements can be buffered with an identification tag at the backplane 540 and read out with data acquisition software stored at the program storage media 550 and used by the computer 545. The three measurements may be combined to create, for example, a histogram or an energy distribution graph. Referring to FIG. 6, a simulated photon energy distribution graph 600 of the bremsstrahlung spectra resulting from the interaction of the electron beam directed upon the radiator is plotted on a log-log scale. The horizontal axis is the photon energy in MeV and the vertical axis is the photon yield binned in units of dN/dEgamma. This photon energy distribution result can be obtained by a data acquisition system such as the one depicted in FIG. 5 when the container 125 detailed in FIG. 1 is absent. Referring to FIGS. 1 and 6, The graph 600 is created by combining a low-Z detector 140 signal 605, a PPADs detector 135 signal 610, and a high-Z detector 150 signal 615. In the presence of a radiological device composed of, for example, uranium, plutonium or neptunium concealed within the interrogated vessel 125, the measured spectrum from the beam flux monitor 130 reflects a precipitous drop in intensity between about 10 and 20 MeV. The radiological material selectively absorbs the photon beam within its energy regime. The terms a or an, as used herein, are defined as one or more than one. The term plurality, as used herein, is defined as two or more than two. The term another, as used herein, is defined as at least a second or more. The terms including and/or having, as used herein, are defined as comprising (i.e., open language). The term coupled, as used herein, is defined as connected, although not necessarily directly, and not necessarily mechanically. The term approximately, as used herein, is defined as at least close to a given value (e.g., preferably within 10% of, more preferably within 1% of, and most preferably within 0.1% of). The term program or software, as used herein, is defined as a sequence of instructions designed for execution on a computer system. A program, or computer program, may include a subroutine, a function, a procedure, an object method, an object implementation, an executable application, an applet, a servlet, a source code, an object code, a shared library/dynamic load library and/or other sequence of instructions designed for execution on a computer system. The appended claims are not to be interpreted as including means-plus-function limitations, unless such a limitation is explicitly recited in a given claim using the phrase(s) “means for” and/or “step for.” Subgeneric embodiments of the invention are delineated by the appended independent claims and their equivalents. Specific embodiments of the invention are differentiated by the appended dependent claims and their equivalents.
	description	The invention relates to a lead substitute material for radiation protection purposes in the energy range of an X-ray tube with a voltage of 60–125 kV. Conventional radiation protection clothing for use in X-ray diagnosis usually contains lead or lead oxide as the protective material. Substitution of other materials for this protective material is desirable for the following reasons, in particular: On the one hand, lead and its processing entail a significant environmental impact and owing to its very large weight, on the other hand, lead necessarily entails a very large heavy weight of the protective clothing and therefore a great physical burden on the user. For years, therefore, attempts have been made to find a substitute material for lead in radiation protection. The use of chemical elements with atomic numbers of from 50 to 76, or their compounds, has predominantly been proposed for this. DE 199 55 192 A1 describes a method for producing a radiation protection material from a polymer as the matrix material and the powder of a metal with a high atomic number. DE 201 00 267 U1 describes a highly elastic, lightweight, flexible, rubber-like radiation protection material, wherein additives of chemical elements with an atomic number greater than or equal to 50, and their oxides, are mixed with a special polymer. To reduce the weight compared with conventional lead shields, EP 0 371 699 A1 proposes a material which likewise contains elements with high atomic numbers, in addition to a polymer as the matrix. A large number of metals are mentioned in this case. Depending on the elements which are used, the attenuation factor or lead equivalent value (International Standard IEC 61331-1, Protective devices against diagnostic medical X-radiation) of the material in question shows a sometimes very pronounced dependency on the radiation energy, which is a function of the voltage of the X-ray tube. The known radiation protection clothing made of lead-free material therefore has a more or less significant reduction in absorption compared with lead below 70 kV and above 110 kV. This means that, in order to achieve the same screening effect as for a material containing lead, a higher weight per unit area of the protective clothing is necessary for this range of the X-ray voltage. The application range of commercially available lead-free protective clothing is therefore generally restricted. In order to be able to substitute lead for radiation protection purposes, an absorption performance as close as possible to that of lead is necessary over a larger energy range, radiation protection materials usually being categorised according to the lead equivalent value and the radiation protection calculations often being based on lead equivalent values. It is an object of the invention to replace lead as a radiation protection material with respect to its screening properties over an energy range of an X-ray tube with a voltage of 60–125 kV, that is to say over a larger energy range, and over a larger thickness range of the nominal lead equivalent values, while simultaneously achieving a weight reduction which is as great as possible. Only materials which are more environmentally friendly than lead are intended to be used in this case. The object of the invention is achieved by a lead substitute material for radiation protection purposes in the energy range of an X-ray tube with a voltage of 60–125 kV, which is characterised in that the lead substitute material comprises Sn, Bi and optionally W, or compounds of these metals, and the composition of the lead substitute material is a function of the nominal lead equivalent value. In order to achieve the object it was therefore necessary, on the one hand, to find a material selection for optimum screening properties over a larger energy range and, on the other hand, to find a material selection for a larger thickness range of the protective layer. Preferred compounds of Sn, Bi and W are their oxides. It is a fundamentally new and surprising discovery that, in order to achieve an optimum result, the composition of lead substitute materials varies as a function of the thickness of the protective material. A lead-free screening material with the extended application range can now be achieved by a combination of tin with bismuth and optionally tungsten, which is matched to the respective nominal lead equivalent value. In a preferred embodiment of the invention, the lead substitute material is characterised in that it has 10–20% by weight of a matrix material, 50–75% by weight of Sn, or Sn compounds, and 20–35% by weight of Bi, or Bi compounds, for nominal lead equivalent values of up to 0.15 mm, and 40–60% by weight of Sn, or Sn compounds, 15–30% by weight of Bi, or Bi compounds and 0–30% by weight of W, or W compounds, for nominal lead equivalent values of 0.15–0.60 mm. In a particularly preferred embodiment of the invention, the lead substitute material is characterised in that it has 52–70% by weight of Sn, or Sn compounds, and 21–32% by weight of Bi, or Bi compounds, for nominal lead equivalent values of up to 0.15 mm, and 42–57% by weight of Sn, or Sn compounds, 15–30% by weight of Bi, or Bi compounds, and 5–27% by weight of W, or W compounds, for nominal lead equivalent values of 0.15–0.60 mm. The matched combination of tin and bismuth and optionally tungsten, or compounds of these metals, now makes it possible to provide an environmentally friendly lead substitute material which is substantially more lightweight than conventional lead or lead oxide material, and which can substitute for the latter in the energy range of an X-ray tube with a voltage of 60–125 kV. This energy range is the essential range for X-ray diagnosis. The criterion when substituting for lead is a 10% deviation of the lead equivalent value from the nominal value, as stipulated in DIN 6813. Radiation protection clothing which is made of the substitute material according to the invention can therefore be worn without restrictions in all applications of X-ray diagnosis. This constitutes a substantial advantage over all known lead substitute materials. In another particularly preferred embodiment of the invention, the lead substitute material is characterised in that it comprises a structure made up of layers with differing composition. The lead substitute material may comprise a structure made up of at least two layers with differing composition, which are separate or connected together, the layer further away from the body comprising predominantly Sn and the layer(s) near the body comprising predominantly Bi and optionally W. The invention will be explained in more detail with reference to the following examples and comparative examples. The measurements of the weight- and energy-related radiation protection effects were based on the IEC 61331-1 standards; particular points to note in this regard are the measurement geometry and the prefiltering mentioned therein for the X-radiation. The results of the measurements are collated in Table 1 and in FIG. 1. TABLE 1Weight per unit area (kg/m2) of the various radiationprotection materials, expressed in terms of the absorptionby pure lead, under measurement conditions accordingto IEC 61331-1 as a function of energy.Protective material60 kV80 kV100 kV125 kV150 kVAbsorption of the97.289.380.874.469.7primary radiationin %0.25 mm of pure2.832.832.832.832.83lead (withoutmatrix) -reference valueLead with matrix3.593.593.593.593.59Commercially3.462.882.963.634.41available lead-freematerial(Optimit R-100A)Commercially3.793.093.204.134.51available lead-freematerial(Xenolite ® NL)Lead substitute2.932.832.833.073.53material accordingto the invention,with the composition:15 wt. % matrix,54 wt. % Sn,12 wt. %W, 19 wt. % Bi Table 1 shows that, for an equal protective effect in the range of 60–125 kV, the lead substitute material according to the invention has the most advantageous weight per unit area of all the lead-free materials. A radiation protection shield with the nominal lead equivalent value 0.25 mm, made of the novel material, therefore weighs around 21% less than a conventional shield with lead as the protective material. The measurements which were carried out furthermore show that the radiation-physical properties of the lead substitute material are dependent both on the energy of the incident radiation and on the layer thickness, that is to say the composition of the lead substitute material needs to be modified for each layer thickness, in order to match it to the absorption performance of lead. The results are shown in Table 2, where the compositions are indicated for conventional lead equivalent values with the corresponding values measured according to IEC 61331-1. WeightNominal leadCompositionper unit60 kV80 kV100 kV125 kVequivalent valueM = MatrixareaBeam qualities according to IEC 61331-1(mm)material(kg/m2)Measured lead equivalent value (mm)0.02565 wt. % Sn + 22 wt. %0.250.0230.0250.0250.023Bi + 15 wt. % M0.0555 wt. % Sn + 30 wt. %0.510.0450.0500.0500.045Bi + 15 wt. % M0.12555 wt. % Sn + 30 wt. %1.250.1200.1250.1250.120Bi + 15 wt. % M0.2554 wt. % Sn + 12 wt. %2.80.240.250.250.23W + 19 wt. %Bi + 15 wt. % M0.3548 wt. % Sn + 20 wt. %3.90.330.350.360.32W + 17 wt. %Bi + 15 wt. % M0.0544 wt. % Sn + 25 wt. %5.50.480.500.500.45W + 16 wt. %Bi + 15 wt. % M As can be seen from Table 2, for example, the substitute material comparable with 0.2 mm of lead consists of 15% by weight matrix material, 54% by weight Sn, 12% by weight W and 19% by weight Bi, with a mass per unit area of 2.8 kg/m2 in total. The matrix material is the substrate and may, for example, consist of rubber or latex. Large deviations from the composition according to the invention detrimentally effect either the allowable application range and/or the weight. If a protective layer with a lead equivalent value of 0.5 mm is required, however, then the composition needs to be modified according to Table 2 in order to achieve the performance corresponding to lead over an energy range of from 60 to 125 kV. In terms of radiation physics, the embodiment of the invention to which claim 5 relates can make it possible to further reduce the user's radiation exposure. For example, the radiation exposure at an X-ray voltage of 100 kV can be reduced by about 15% if the outer layer consists exclusively of tin and the inner layer consists of bismuth and optionally tungsten. The weight of the protective clothing can advantageously be reduced further by taking this relationship into account.
048470406	summary	BACKGROUND OF THE INVENTION 1. Field of the Invention The invention relates to a nuclear power plant with a gas cooled high temperature reactor, and more particularly to a reactor with a spherical fuel element core and a prestressed concrete pressure vessel surrounding the high temperature reactor. The reactor cavity is clad with a liner and contains a plurality of heat exchangers, preferably steam generators for operational heat removal. At least two auxiliary heat exchangers for decay heat removal are also arranged in the reactor cavity. The prestressed concrete pressure vessel has a thermal protection system including a thermal insulating layer and a liner cooling system comprising a plurality of cooling pipes through which water flows and which, together with intermediate heat exchangers and cooling water pumps, a closed intermediate cooling loop used for the removal of the decay heat in case of an auxiliary heat exchanger failure. 2. Description of the Related Technology U.S. Pat. No. 4,554,129 shows a gas cooled nuclear reactor installed in the cavity of a prestressed concrete pressure vessel. The heat generated in the reactor core is transferred to several heat exchangers located in the cavity above the nuclear reactor. No auxiliary heat exchangers are provided for the removal of the decay heat. Decay heat is removed by a liner cooling system made up of cooling pipes welded to the cavity liner inlet and return lines connected to the cooling pipes and a plurality of circulating pumps. Heat is transported from the reactor core to the liner cooling system by central pipes provided inside the heat exchanger and by shut-off valves closed in normal operation in a annular space bordering on the liner. A natural downward directed flow of heat is established in the annular space, so that all areas of the liner are exposed to the cooling gas. DE-OS No. 33 35 268 shows a high temperature reactor with spherical fuel elements. Steam generators and blowers for operational removal of heat and auxiliary cooling systems for the decay heat removal in case of accidents are located together with the other components of the primary loop in the cavity of the prestressed concrete pressure vessel. The cavity is provided with a liner exhibiting a cooling system. This liner cooling system may be used for the removal of the decay heat, should the auxiliary cooling systems fail. It has also been proposed to conduct the decay heat to the liner cooling system by natural convection without coolant loss in case of accidents. DE-OS No. 31 21 377 shows a liner cooling system for a prestressed concrete pressure vessel intended to house a nuclear reactor and containing an interior thermal insulating layer. The liner cooling system is redundant, i.e. it consists of several mutually independent water circulating loops. SUMMARY OF THE INVENTION It is an object of the invention to provide a nuclear power plant with an adequate mass flow of water in the liner cooling system for removal of decay heat in any situation. According to the invention, this object is attained by the following characteristics: (a) Arranging intermediate heat exchangers above the upper edge of the prestressed concrete pressure vessel (b) Dividing the liner cooling system into several cooling pipe zones according to vertical or height positions wherein at least one cooling pipe zone in an upper zone range (by height) is always connected to a zone in a lower location; (c) Arranging a supplemental pump connected to an emergency power system in parallel with each primary cooling water pump, where the supplemental pumps may have a significantly lower capacity; (d) Providing every cooling water pump with a bypass line with an actively or passively controlled check valve, where the check valve opens if the cooling water pump is inoperative and; (e) Providing an installation for removal of the decay heat by natural convection in case of a failure of the intermediate heat exchangers connected to the liner cooling system. The natural convection decay heat removal system includes: a water reservoir with a nitrogen cushion connected to the forward or feed line of the liner cooling system, a vertical boiling tube, placed in the intermediate loop, connected to the liner cooling system return line and located geodesically higher than the liner cooling system, a water separator located in the intermediate cooling loop, connected to the vertical boiling tube and including a nitrogen cushion, and a safety blow-off valve connected to the water separator. Adequate flow through the liner cooling system requires an appropriate pressure forcing the water through the system. The amount of pressure is referred to as a "driving pressure difference" or .DELTA.P.sub.tr. The decisive factor for the presence of an adequately high water mass flow in the liner cooling system without the use of active aggregates, such as pumps or blowers, is a sufficiently large driving pressure difference .DELTA.P.sub.tr in the intermediate cooling loop. The driving pressure difference is proportional to the product .DELTA..rho..g.H (.DELTA..rho.=difference in density, H=height). According to invention, the necessary driving pressure difference .DELTA.P.sub.tr is provided by acting on the factor H, i.e. by the geometric layout of the intermediate heat exchangers, and by an appropriate interconnection of cooling pipe zones of differing heights. The latter measure is required, as the reactor protection building enclosing the prestressed concrete pressure vessel only has a limited height available for the intermediate heat exchangers. According to the invention, in case the cooling water pumps are not operating due to a power failure, the necessary driving pressure difference .DELTA.P.sub.tr is assured by supplemental pumps connected in parallel to the cooling water pumps and operated with emergency power. No problems arise from the connection to the emergency power system as the supplemental pumps have capacities lower by one to two orders of magnitude than the normal operation cooling water pumps in view of a permissibly higher cooling water heating range and correspondingly lower water flow rates. The reduction of pressure losses in the intermediate cooling loops leads to an increased water flow rate. High pressure losses can occur in case of a failure of a the cooling water pumps; that is exactly a situation where an adequate flow of water is absolutely necessary since heat may be removed from the liner cooling system only by natural convection. High pressure losses of inactive cooling water pumps are reduced to a minimum by the proposed bypass lines with check valves. In case the intermediate heat exchangers of the intermediate cooling loop fail, a device connected to the liner cooling system is activated, whereby the decay heat is removed by the heat of evaporation contained in a water reservoir through a safety blow-off valve. The advantage of the invention lies in that decay heat can be removed safely even in case of a failure of the auxiliary heat exchangers and the intermediate heat exchangers and also in the case of a power failure, so that any impermissible heating of the primary loop components is prevented and the risk of the release of activities (radioactivity) reduced. The higher density differences of cooling water present in the liner cooling system may be utilized advantageously to increase the water flow rate by natural convection in the rare case of accidents, as a result of the high, acceptable, concrete and liner temperatures during removal of decay heat. An increase in temperature of 200.degree.-300.degree. C. over normal operating temperature due the thermal capacity. The temperature rise will result in a considerable rise in density difference .DELTA.P.sub.tr of the cooling water above the normal operation level. The driving pressure difference .DELTA.P.sub.tr is also raised in the process. In a single phase heat removal process in the liner cooling system the temperature of the cooling water may be raised to slightly below the boiling point. A further increase of .DELTA..rho. may be obtained by two-phase heat removal (bubble boiling) in the liner cooling system. An increase in the flow of water in the liner cooling system may be achieved by reducing pressure losses in the intermediate cooling loop. Pressure losses in the operation of decay heat removal is defined by: ##EQU1## wherein N=nominal operation NWA=decay heat removal operation PA1 .DELTA..delta.=heating range of the cooling water PA1 Q=volume of heat removed by the liner cooling system .DELTA.P.sub.P =pressure loss of the inactive cooling water pumps. It follows from this relationship that it is advantageous to choose a high heating range for the cooling water (to the boiling temperature or slightly thereunder). The ratio of the volumes of heat removed in nominal operation and in a decay heat removal operations are determined by the temperatures generated. The nominal pressure losses .DELTA.P.sub.N may be reduced by the appropriate layout of the liner cooling system and choice of the components of the intermediate cooling loop. The following measures may be effected to reduce nominal pressure loss extensive equalization of pressure losses in the individual cooling pipes, in particular reduction of peak values, or selection of intermediate heat exchangers with low pressure losses on the liner cooling system side. The composition of different cooling pipe zones of the liner cooling system may be established advantageously by direct coupling during the layout of the liner cooling system. Alternatively, the cooling pipe zones may be connected to each other in the decay heat removal operation by short circuiting the forward and return lines of the zones by externally accessible valves. The valves may be manually actuated. This so-called external coupling is taken into consideration in the layout of the liner cooling system. Check valves located in the cooling water pump bypass lines may be actuated in a number of ways: actively, controlled by the speed of the cooling water pump involved (for example, the check valve concerned remains closed at pump speed of equal to or higher than 100 rpm and opens if the velocity drops below that rate), or passively by the pressure difference applied to the cooling water pump involved. In the latter case the check valve concerned, which in normal operation is closed, may be opened by its own weight or the release of a spring. According to a further embodiment of the invention the decay heat removal capacity of the liner cooling system by natural and forced convection must be optimized against the thermal resistance of the thermal insulating layer so that the maximum permissible liner, fuel element and reactor installation temperature is not exceeded. Heat resistance corresponding to insulation thickness of the thermal insulating layer decisively affects a rise in temperature of structural parts located within the prestressed concrete pressure vessel in case of failure of decay heat removal by the auxiliary heat exchangers and the liner cooling system. The aforementioned component temperatures are calculated as a function of the thermal resistivity of the thermal insulating layer during layout of the thermal insulating layer for the case of a "failure of the decay heat removal installation". The optimum is achieved when the same safety margin from the maximally permissible limiting temperature is established for all of the structural components. The two extreme cases described below serve to demonstrate this condition in case of failure of the auxiliary heat exchangers and the removal of the decay heat through natural convection in the liner cooling system (this is also valid for forced convection, decay heat removal by the liner cooling system. If the insulating layer thickness is inadequate: Excessive heat is transported to the liner and in the liner cooling system and leads to exceeding the maximum permissible liner and concrete temperatures, however, in view of the effective removal of heat from the primary loop through the liner cooling system the temperature of the fuel elements and of the reactor installations remain far below its limiting value. If the thickness of the thermal insulating layer is too large: Inadequate decay heat removal from the primary loop results in fuel element and the reactor installation temperatures exceeding their failure limits due to an excessive cooling gas temperature while the liner temperature remains far below the permissible limiting value (and the heat removal capacity of the liner cooling system remains unutilized). According to a feature of the invention an optimization is effected for the liner cooling system alone (i.e. not for the entire thermal protection system) relative to the maximum removal of decay heat by natural convection and of a minimalization of the liner temperature respectively. The parameters used in for this optimization are the cooling pipe diameter and the cooling pipe spacing. The flow of cooling water and thus the amount of heat that may be removed in a natural convection liner cooling system depends on the choice of the cooling pipe diameter and the existing cooling pipe spacing. The flow of cooling water in a liner cooling system (flow per m.sup.2 of cooling water) is predetermined by the heating range. In order to increase the natural convection flow of the cooling pipe cross section must be enlarged proportionally to the spacing of the cooling pipes. This measure is limited by the maximum temperature of the liner which rises with increasing cooling pipe spacing. A variation of the cooling pipe cross section and the cooling pipe spacing leads to optimalization of minimizing the liner temperature and to maximizing the cooling water flow and liner cooling system natural convection heat removal capacity resulting in a reduction of the residual risk in case of a decay heat removal failure. This optimalization leads to a reduction in liner cooling pipe spacing (while maintaining cooling pipe diameter) or an increase in cooling pipe diameter and a correspondingly larger cooling pipe spacing or a combination of the two measures. Advantageously, the installation for the removal of decay heat, in case of a failure of the intermediate heat exchangers, the pressure of the nitrogen cushion in the water separator and the actuating pressure of the safety blow-off valve may be correlated in a manner such that heat removal from the intermediate cooling loop to a predetermined cooling water temperature is effected initially in a single phase range and passes into a two-phase range only when the saturation pressure in the water separator attains the actuating pressure the safety blow-off valve. Additionally the water reservoir may have an open configuration and be located geodesically high enough so that upon a rise of the pressure in the intermediate cooling loop to the actuating pressure of the safety blow-off valve a blow-off through the open water reservoir is prevented. Alternatively the water container may have a closed configuration and is pressure connects by a connecting line to the water separator and that its geodesic height is determined so that the same water level may be established in the water separator and the water reservoir.
	summary
050383701	claims	1. Apparatus for generating an X-ray or gamma beam with small cross-section and variable direction comprising an X-ray or gamma emitter, from the focus of which a bundle of rays emerges, and a diaphragm arrangement, which cuts out a beam from the bundle of rays and comprises a hollow-cylindrical first diaphragm body which is rotatable about its axis of symmetry and has two mutually offset helical slits on the circumference, said slits winding around the diaphragm body in at least one turn each and are shaped in such a way that at least one straight line runs through the slits towards the focus, the position of which line can be varied by rotation of the diaphragm body. 2. Apparatus according to claim 1 wherein each slit has an integral number of turns. 3. Apparatus according to claim 1 including a second diaphragm body which only allows there through a primary beam of the bundle of rays, and in that the second diaphragm body is arranged such that the primary beam always coincides with said at least one straight line. 4. Apparatus according to claim 3 wherein the second diaphragm body has the form of a hollow cylinder, the axis of which lies in the plane containing the axis of symmetry and the focus and the cross-section of which is circular and in that the second diaphragm body is provided with two helical slits mutually offset by 180.degree. on the circumference. 5. Apparatus according to claim 4, wherein the slits on the circumference of the second diaphragm body describe in angle of 180.degree.. 6. Apparatus according to claim 5 including a drive device which drives the first diaphragm body at 2n times the angular velocity as the second diaphragm body. 7. Apparatus according to claim 6 wherein the angle at the circumference which a slit on the first diaphragm body describes is greater by a factor of 2n than the angle at circumference described by a slit on the second diaphragm body where n is an integer. 8. Apparatus according to claim 4 wherein the two diaphragm bodies are arranged concentrically to each other and one encloses the other and in that the slits of the second diaphragm body are wider than at least one of the slits in the first diaphragm body. 9. Apparatus according to claim 8 wherein the first diaphragm body encloses the second diaphragm body. 10. Apparatus according to claim 1 wherein the slits of the diaphragm body have pitches differing from each other. 11. Apparatus according to claim 10 wherein of the slits in the diaphragm body, the one with the greater pitch is narrower than the other one. 12. Apparatus according to claim 1 including a slit diaphragm, the slit of which coincides with the axis of rotation of the diaphragm body and which determines the dimensions of the cut-out beam in the direction perpendicular to its longitudinal direction. 13. Apparatus according to claim 3 wherein the second diaphragm body has the form of a hollow cylinder, the axis of which lies in the plane containing the axis of symmetry and the focus and the cross-section of which is semicircular and in that the second diaphragm body is provided with one slit of semicircular cross-section. 14. Apparatus according to claim 13 wherein the slit on the circumference of the second diaphragm body describes an angle of 180.degree.. 15. Apparatus according to claim 2 including a second diaphragm body which only allows there through a primary beam of the bundle of the rays, and in that the second diaphragm body is arranged such that the primary beam always coincides with said at least one straight line. 16. Apparatus according to claim 13 wherein the two diaphragm bodies are arranged concentrically to each other an one encloses the other and in that the slit of the second diaphragm body is wider than at least one of the slits in the first diaphragm body. 17. Apparatus according to claim 16 wherein the first diaphragm body encloses the second diaphragm body. 18. Apparatus according to claim 17 wherein the slits of the first mentioned diaphragm body have pitches differing from each other. 19. Apparatus according to claim 18 wherein of the slits in the first diaphragm body, the one with the greater pitch is narrower than the other one. 20. Apparatus according to claim 19 including a slit diaphragm the slit of which coincides with the axis of rotation of the first diaphragm body and which determines the dimensions of the cut-out beam in the direction perpendicular to its longitudinal direction.
	abstract	A method for manufacturing a core barrel according to the embodiment includes: welding one end part of a short ring to a lower core support plate; and machining the lower core support plate to which the short ring is welded. The machining of the lower core support plate includes forming a placement surface on which the fuel assembly is to be placed; and forming a fuel alignment pin hole, in which a fuel alignment pin for positioning the fuel assembly is to be inserted. After the machining of the lower core support plate, a main body barrel is welded to the other end part of the short ring, where the main body barrel covers the reactor core including the fuel assembly to be placed on the placement surface.
	claims	1. A composition comprising a polymer having a degree of polymerization n, ranging from 2 to 10000, and containing 2 to 10000 monomer units, said monomer units being:either monomer units derived from 4-vinylpyridine, in which the carbons in position 2 and 6 are substituted by one of the substituents of the following group: carboxylic acid, or ester of 1 to 20 carbons,or monomer units derived from a co-monomer, provided that said monomer units derived from 4-vinylpyridine represent at least 20% of the degree of polymerization n, said polymer being optionally complexed with a metal, said polymer being linear or cross-linked. 2. The composition according to claim 1, comprising a linear or cross-linked polymer consisting of a monomer unit, of Formula I,in which:R4 is a compound allowing the propagation of the polymerization, whether or not originating from a polymerization initiator, or allowing the termination of the polymerization, or a transfer agent,A is a compound derived from said polymerization initiator or a fragment derived from the polymerization method,a and r are identical or different and are 0 or 1,i is a strictly positive, indexed integer, varying from 1 to n, itself comprised from 2 to 10000,and for each i:Ri,1 and Ri,2 are substituents chosen from the following group: carboxylic acid, or ester of 1 to 20 carbons,Bi is a monomer unit derived from a co-monomer, whether or not forming a cross-linking bridge,Mi is a metal,ni et mi, are integers equal to 0 or 1,ni+mi=1,xi is a number comprised from 0 to 6,pi is the electrical charge of the metal complex ranging from −6 to +6,said polymer being linear when there is no Bi forming a cross-linking bridge,said polymer being cross-linked when there is at least one Bi forming a cross-linking bridge between two linear polymers,or comprising a non-cross-linked linear polymer consisting of monomer unit, of Formula II,in which:the definitions of Ri,1, Ri,2, A, R4, a, r, Mi, xi, ni, mi, pi and i are as described for Formula I,Bi is a monomer unit derived from a co-monomer which does not form a cross-linking bridge. 3. Composition according to claim 1, comprising a polymer of Formula III,in which:R4 is a compound allowing the propagation of the polymerization, whether or not originating from a polymerization initiator, or allowing the termination of the polymerization, or a transfer agent,A is a compound derived from said polymerization initiator or a fragment derived from the polymerization method,a and r are identical or different and are 0 or 1,i is a strictly positive, indexed integer, varying from 1 to n, itself comprised from 2 to 10000,and for each i:Ri,1 and Ri,2 are substituents chosen from the following group: carboxylic acid, or ester of 1 to 20 carbons,Mi is a metal,xi is a number comprised from 0 to 6,pi is the electrical charge of the metal complex ranging from −6 to +6. 4. The composition according to claim 1, comprising a homopolymer of 2,6-dicarboxyl-4-vinylpyridine of Formula V,in which:R4 is a compound allowing the propagation of the polymerization, whether or not originating from a polymerization initiator, or allowing the termination of the polymerization, or a transfer agent,A is a compound derived from said polymerization initiator or a fragment derived from the polymerization method,a and r are identical or different and are 0 or 1,M is a metal,x is a number comprised from 0 to 6,p is the electrical charge of the metal complex ranging from −6 to +6,or comprising a homopolymer of 2,6-dicarboxyl-4-vinylpyridine represented by Formula VI,in which:the definitions of R4, A, a and r are as described for Formula V,or comprising a homopolymer of Formula XII,in which:the definitions of A, R4, r, and a are as described for Formula V,or comprising a homopolymer of 2,6-dicarboxyl-4-vinylpyridine complexed with uranium, represented by Formula XIII,in which:the definitions of A, R4, r, a are as described for Formula V,xU is a number comprised from 0 to 1. 5. The composition according to claim 1, comprising a polymer with at least one monomer unit derived from a co-monomer,or comprising a polymer with at least one monomer unit derived from a co-monomer, in which the level of monomer unit derived from a co-monomer varies from a value strictly greater than 0% to a value less than 80%. 6. The composition according to claim 1, comprising a polymer with at least one monomer unit derived from a co-monomer, said monomer units being derived from styrene or acrylic acid,or comprising a polymer with at least one monomer unit derived from a co-monomer, said monomer units being derived from styrene or acrylic acid, in which the level of monomer unit derived from a co-monomer varies from a value strictly greater than 0% to a value less than 80%. 7. The composition according to claim 1, comprising a block copolymer of Formula XVIII,in which:R4 is a compound allowing the propagation of the polymerization, whether or not originating from a polymerization initiator, or allowing the termination of the polymerization, or a transfer agent,A is a compound derived from said polymerization initiator or a fragment derived from the polymerization method,a and r are identical or different and are 0 or 1,R1 and R2 are substituents chosen from the following group: carboxylic acid, or ester of 1 to 20 carbons,B is a monomer unit derived from a non-forming co-monomer cross-linking bridge,d is the degree of polymerization of the block consisting of monomer units derived from 4-vinylpyridine and a is an integer,c is the degree of polymerization of the block consisting of monomer units derived from co-monomer and b is an integer,c+d=n,or comprising a two-block copolymer of Formula XIX ain which:the definitions of A, R4, a, r, c and d are as described for Formula XVIII. 8. The composition according to claim 1, comprising a cross-linked copolymer, said polymer being of Formula XX,in which:W is a strictly positive, indexed integer varying from 1 to the number of polymers cross-linked with the polymer of index 0, itself comprised from 1 to 1000,A0 and Aw are compounds derived from polymerization initiators,R4,0 and R4,w are compounds allowing the propagation of the polymerization, whether or not originating from said polymerization initiator, or allowing the termination of the polymerization, or a transfer agent,a0, r0, aw and rw are identical or different and are 0 or 1,i and j, w are integers which are strictly positive, indexed, varying respectively from 1 to ni and 1 to nj,w, ni and nj,w, being comprised from 1 to 9999,ni+n3,w=n, n being comprised from 3 to 10000,and for each i and each j, w:Ri,1, Ri,2, Rj,w,1 and Rj,w,2 are substituents chosen from the following group: carboxylic acid, or ester from 1 to 20 carbons,Bi,j,w is a monomer unit derived from a co-monomer, forming a cross-linking bridge between the polymer 0 in position i and the polymer of index w in position j,Ci and Cj,w are monomer units derived from a co-monomer which does not form a cross-linking bridge,ni, mi, oi, nj,w, mj,w, and oj,w, are integers equal to 0 or 1,ni+mi+oi=1,nj,w+mj,w+oj,w=1,the sum of the oj,w is non-zero and the sum of the oi is non-zero,or comprising a cross-linked copolymer of Formula XXI:in which:the definitions of w, A0, Aw, a0, aw, i, j, w, ni, nj,w, mi, mj,w, oi, oj,w, R4,0, R4,w, r0 and rw are as described for Formula XX. 9. The composition according to claim 1, comprising a polymer in which said polymer is complexed with a metal, or with a metal chosen from actinides, lanthanides or transition metals, or with uranium. 10. The composition according to claim 1, comprising a polymer soluble in aqueous solution or in sea water,or comprising a polymer complexed with a soluble or insoluble metal in aqueous solution,or comprising a polymer soluble in aqueous solution not complexed with a metal and insoluble in solution complexed with a metal, or with uranium. 11. The composition according to claim 1, comprising a polymer soluble in organic solvents, or in acetonitrile, or dimethylsulfoxide (DMSO). 12. The composition according to claim 1, comprising a polymer in which A, Ao and Aw are derived from a polymerization initiator chosen from 4-(chloromethyl)-benzoyl chloride, benzyl chloride, AIBN, methyl-2-bromo-2-methylpropanoate,said polymer being of Formula I,in which:R4 is a compound allowing the propagation of the polymerization, whether or not originating from a polymerization initiator, or allowing the termination of the polymerization, or a transfer agent,a and r are identical or different and are 0 or 1,i is a strictly positive, indexed integer, varying from 1 to n, itself comprised from 2 to 10000,and for each i:Ri,1 and Ri,2 are substituents chosen from the following group: carboxylic acid, or ester of 1 to 20 carbons,Bi is a monomer unit derived from a co-monomer, whether or not forming a cross-linking bridge,Mi is a metal,ni et mi, are integers equal to 0 or 1,ni+mi=1,xi is a number comprised from 0 to 6,pi is the electrical charge of the metal complex ranging from −6 to +6,said polymer being linear when there is no Bi forming a cross-linking bridge,said polymer being cross-linked when there is at least one Bi forming a cross-linking bridge between two linear polymers,or said polymer being of Formula IIin which:the definitions of Ri,1, Ri,2, R4, a, r, Mi, xi, ni, mi, pi and i are as described for Formula I,Bi is a monomer unit derived from a co-monomer which does not form a cross-linking bridge,or said polymer being of Formula IIIin which:R4 is a compound allowing the propagation of the polymerization, whether or not originating from a polymerization initiator, or allowing the termination of the polymerization, or a transfer agent,a and r are identical or different and are 0 or 1,i is a strictly positive, indexed integer, varying from 1 to n, itself comprised from 2 to 10000,and for each i:Ri,1 and Ri,2 are substituents chosen from the following group: carboxylic acid, or ester of 1 to 20 carbons,Mi is a metal,xi is a number comprised from 0 to 6,pi is the electrical charge of the metal complex ranging from −6 to +6,or said polymer being of Formula XVIII,in which:R4 is a compound allowing the propagation of the polymerization, whether or not originating from a polymerization initiator, or allowing the termination of the polymerization, or a transfer agent,a and r are identical or different and are 0 or 1,R1 and R2 are substituents chosen from the following group: carboxylic acid, or ester of 1 to 20 carbons,B is a monomer unit derived from a non-forming co-monomer cross-linking bridge,d is the degree of polymerization of the block consisting of monomer units derived from 4-vinylpyridine and a is an integer,c is the degree of polymerization of the block consisting of monomer units derived from co-monomer and b is an integer,c+d=n,or said polymer being of Formula XX,in which:W is a strictly positive, indexed integer varying from 1 to the number of polymers cross-linked with the polymer of index 0, itself comprised from 1 to 1000,R4,0 and R4,w are compounds allowing the propagation of the polymerization, whether or not originating from said polymerization initiator, or allowing the termination of the polymerization, or a transfer agent,a0, r0, aw and rw are identical or different and are 0 or 1,i and j, w are integers which are strictly positive, indexed, varying respectively from 1 to ni and 1 to nj,w, ni and nj,w, being comprised from 1 to 9999,ni+nj,w=n, n being comprised from 3 to 10000,and for each i and each j, w:Ri,1, Ri,2, Rj,w,1 and Rj,w,2 are substituents chosen from the following group: carboxylic acid, or ester from 1 to 20 carbons,Bi,j,w is a monomer unit derived from a co-monomer, forming a cross-linking bridge between the polymer 0 in position i and the polymer of index w in position j,Ci and Cj,w are monomer units derived from a co-monomer which does not form a cross-linking bridge,ni, mi, oi, nj,w, mj,w, and oj,w are integers equal to 0 or 1,ni+mi+oi=1,nj,w+mj,w+oj,w=1,the sum of the oj,w is non-zero and the sum of the oi is non-zero. 13. The composition according to claim 1, comprising a polymer in which R4, R4,0 et R4,w are chosen from a chlorine atom, or the compounds of Formula XXII,or comprising a polymer of Formula XXIII,said polymer being of Formula I,in which:A is a compound derived from said polymerization initiator or a fragment derived from the polymerization method,a and r are identical or different and are 0 or 1,i is a strictly positive, indexed integer, varying from 1 to n, itself comprised from 2 to 10000,and for each i:Ri,1 and Ri,2 are substituents chosen from the following group: carboxylic acid, or ester of 1 to 20 carbons,Bi is a monomer unit derived from a co-monomer, whether or not forming a cross-linking bridge,Mi is a metal,ni et mi, are integers equal to 0 or 1,ni+mi=1,xi is a number comprised from 0 to 6,pi is the electrical charge of the metal complex ranging from −6 to +6,said polymer being linear when there is no Bi forming a cross-linking bridge,said polymer being cross-linked when there is at least one Bi forming a cross-linking bridge between two linear polymers,or said polymer being of Formula IIin which:the definitions of Ri,1, Ri,2, A, a, r, Mi, xi, ni, mi, pi and i are as described for Formula I,Bi is a monomer unit derived from a co-monomer which does not form a cross-linking bridge,or said polymer being of Formula IIIin which:A is a compound derived from said polymerization initiator or a fragment derived from the polymerization method,a and r are identical or different and are 0 or 1,i is a strictly positive, indexed integer, varying from 1 to n, itself comprised from 2 to 10000,and for each i:Ri,1 and Ri,2 are substituents chosen from the following group: carboxylic acid, or ester of 1 to 20 carbons,Mi is a metal,xi is a number comprised from 0 to 6,pi is the electrical charge of the metal complex ranging from −6 to +6,or said polymer being of Formula XVIII,in which:A is a compound derived from said polymerization initiator or a fragment derived from the polymerization method,a and r are identical or different and are 0 or 1,R1 and R2 are substituents chosen from the following group: carboxylic acid, or ester of 1 to 20 carbons,B is a monomer unit derived from a non-forming co-monomer cross-linking bridge,d is the degree of polymerization of the block consisting of monomer units derived from 4-vinylpyridine and a is an integer,c is the degree of polymerization of the block consisting of monomer units derived from co-monomer and b is an integer,c+d=n,or said polymer being of Formula XX,in which:W is a strictly positive, indexed integer varying from 1 to the number of polymers cross-linked with the polymer of index 0, itself comprised from 1 to 1000,A0 and Aw are compounds derived from polymerization initiators,a0, r0, aw and rw are identical or different and are 0 or 1,i and j, w are integers which are strictly positive, indexed, varying respectively from 1 to ni and 1 to nj,w, ni and nj,w, being comprised from 1 to 9999,ni+nj,w=n, n being comprised from 3 to 10000,and for each i and each j, w:Ri,1, Ri,2, Rj,w,1 and Rj,w,2 are substituents chosen from the following group: carboxylic acid, or ester from 1 to 20 carbons,Bi,j,w is a monomer unit derived from a co-monomer, forming a cross-linking bridge between the polymer 0 in position i and the polymer of index w in position j,Ci and Cj,w are monomer units derived from a co-monomer which does not form a cross-linking bridge,ni, mi, oi, nj,w, mj,w, and oj,w, are integers equal to 0 or 1,ni+mi+oi=1,nj,w+mj,w+oj,w=1,the sum of the oj,w is non-zero and the sum of the oi is non-zero. 14. A method for the preparation of a polymer having a degree of polymerization n, ranging from 2 to 10000, and containing 2 to 10000 monomer units, said monomer units being:either monomer units derived from 4-vinylpyridine, in which the carbons in position 2 and 6 are substituted by one of the substituents of the following group: carboxylic acid, or ester of 1 to 20 carbons,or monomer units derived from a co-monomer,provided that said monomer units derived from 4-vinylpyridine represent at least 20% of the degree of polymerization n,said polymer being optionally complexed with a metal,said polymer being linear or cross-linked,said method comprising the following steps:A radical polymerization step starting with the initiation of a polymerization initiator to obtain an initiated polymerization initiator followed by the contacting of said initiated polymerization initiator with,at least one monomer derived from 4-vinylpyridine in which the carbons in position 2 and 6 are substituted by ester of 1 to 20 carbons,and optionally with at least a co-monomer,with or without cross-linking,to obtain a polymer,optionally a modification step by bringing said polymer into contact with a reagent to modify at least one of the above substituents in position 2 and/or 6, in order to obtain a optionally modified polymer,optionally a complexation step by bringing said optionally modified polymer into contact with a metal to obtain an optionally modified and optionally complexed polymer. 15. The method of preparation according to claim 14, wherein the polymer is a polymer of Formula I:in which:R4 is a compound allowing the propagation of the polymerization, whether or not originating from a polymerization initiator, or allowing the termination of the polymerization, or a transfer agent,A is a compound derived from said polymerization initiator or a fragment derived from the polymerization method,a and r are identical or different and are 0 or 1,i is a strictly positive, indexed integer, varying from 1 to n, itself comprised from 2 to 10000,and for each i:Ri,1 and Ri,2 are substituents chosen from the following group: carboxylic acid, or ester of 1 to 20 carbons,Bi is a monomer unit derived from a co-monomer, whether or not forming a cross-linking bridge,Mi is a metal,ni et mi, are integers equal to 0 or 1,ni+mi=1,xi is a number comprised from 0 to 6,pi is the electrical charge of the metal complex ranging from −6 to +6,said polymer being linear when there is no Bi forming a cross-linking bridge,said method comprising:a step of radical polymerization starting with the initiation of a polymerization initiator then continuing with the contacting of said initiated polymerization initiator with a 4-vinylpyridine derivative of Formula XXIV,in which:Zi,1 et Zi,2 are ester from 1 to 20 carbons,with optionally at least one co-monomer,with or without cross-linking bridge,to obtain the polymer of Formula XXV,optionally a modification step when at least one of the Zi,1 is different from Ri,1 or when at least one of the Zi,2 is different from Ri,2, of said polymer of Formula XXV with a reagent containing a hydroxide anion, carbonate or phosphate, to obtain the polymer of Formula XXVI,optionally a step of complexing said polymer of Formula XXVI with at least one metallic compound to obtain the polymer of Formula I. 16. The method of preparation according to claim 14, wherein the polymer is a polymer of Formula IV:in whichR4 is a compound allowing the propagation of the polymerization, whether or not originating from a polymerization initiator, or allowing the termination of the polymerization, or a transfer agent,A is a compound derived from said polymerization initiator or a fragment derived from the polymerization method,a and r are identical or different and are 0 or 1,i is a strictly positive, indexed integer, varying from 1 to n, itself comprised from 2 to 10000,and for each i:Mi is a metal,xi is a number comprised from 0 to 6,pi is the electrical charge of the metal complex ranging from −6 to +6,R1 et R2 are substituents chosen from the following group: carboxylic acid, or ester of 1 to 20 carbons,said method comprising:a radical polymerization step starting with the initiation of a polymerization initiator then continuing by bringing said initiated polymerization initiator into contact with a monomer derived from 4-vinylpyridine of Formula XXXI,in whichZ1 and Z2 are ester of 1 to 20 carbons,to obtain a polymer of Formula XXXII,in which:the definitions of A, R4, i, r, and a are as described for Formula IV,a modification step, when Z1 is different from R1 or when Z2 is different from R2, of said polymer of Formula XXXII with a reagent containing a hydroxide, carbonate or phosphate anion, to obtain the polymer of Formula XXXIII,optionally a complexing step of said polymer of Formula XXXIII with at least one metal to obtain the polymer of Formula IV,or wherein the polymer is a polymer of Formula V:in which:R4 is a compound allowing the propagation of the polymerization, whether or not originating from a polymerization initiator, or allowing the termination of the polymerization, or a transfer agent,A is a compound derived from said polymerization initiator or a fragment derived from the polymerization method,a and r are identical or different and are 0 or 1,M is a metal,x is a number comprised from 0 to 6,p is the electrical charge of the metal complex ranging from −6 to +6,said method comprising:a radical polymerization step starting with the initiation of a polymerization initiator and then continuing with the contacting of said initiated polymerization initiator with 4-vinylpyridine derivative of Formula XL,to obtain a polymer of Formula XII,a step of modifying said polymer of Formula XII with a base to obtain a polymer of Formula VI,a step of complexing said polymer of Formula VI with a metal to obtain a polymer of Formula V. 17. The method of preparation according to claim 14, wherein the polymer is a polymer of Formula XVIII:in which:R4 is a compound allowing the propagation of the polymerization, whether or not originating from a polymerization initiator, or allowing the termination of the polymerization, or a transfer agent,A is a compound derived from said polymerization initiator or a fragment derived from the polymerization method,a and r are identical or different and are 0 or 1,R1 and R2 are substituents chosen from the following group: carboxylic acid, or ester of 1 to 20 carbons,B is a monomer unit derived from a non-forming co-monomer cross-linking bridge,d is the degree of polymerization of the block consisting of monomer units derived from 4-vinylpyridine and a is an integer,c is the degree of polymerization of the block consisting of monomer units derived from co-monomer and b is an integer,c+d=n,said method comprising:a step of radical polymerization starting with 1 priming of a polymerization initiator then continuing by bringing said initiated polymerization initiator into contact with a monomer derived from 4-vinylpyridine of Formula XXXI,in which:Z1 and Z2 are ester of 1 to 20 carbons, with a polymer of Formula XXXIV,in which:the definitions of A, R4, a, and r are as described for Formula XVIII,B is a monomer unit derived from a co-monomer which does not form a cross-linking bridge,c is the degree of polymerization of the polymer and c is an integer strictly lower than 0.8 n,to obtain the polymer of Formula XXXV,in which:d is an integer,c+d=n,a modification step, when Z1 is different from R1 or when Z2 is different from R2, of said polymer of Formula XXXV with a reagent containing an anion hydroxide, carbonate or phosphate, to obtain the polymer of Formula XVIII,said method optionally comprising before the radical polymerization step by bringing a monomer derived from 4-vinylpyridine of Formula XXXI into contact with a polymer of Formula XXXIV,a radical polymerization step of a co-monomer of Formula XXXVI to obtain said polymer of Formula XXXIV,wherein the definition of B is a monomer unit derived from a co-monomer which does not form a cross-linking bridge. 18. The method of preparation according to claim 14, wherein the polymer is a copolymer of Formula XIX:in which:R4 is a compound allowing the propagation of the polymerization, whether or not originating from a polymerization initiator, or allowing the termination of the polymerization, or a transfer agent,A is a compound derived from said polymerization initiator or a fragment derived from the polymerization method,a and r are identical or different and are 0 or 1,d is the degree of polymerization of the block consisting of monomer units,c is the degree of polymerization of the block consisting of monomer units derived from co-monomer, said method comprising:a step of radical polymerization by bringing a polymer of Formula XLI into contact,with 4-vinylpyridine derivative of Formula XL, to obtain the polymer of Formula XIX,said method optionally comprising, before the radical polymerization step by bringing a monomer derived from 4-vinylpyridine of Formula XL into contact with a polymer of Formula XLI,a step of radical polymerization of styrene, to obtain said polymer of Formula XIX. 19. The method for the preparation of a polymer, according to claim 14, in which the said radical polymerization step is a polymerization of the MP, RAFT, ATRP, SARA ATRP type or a conventional radical polymerization. 20. The method of fighting counterfeiting, comprising a step of labeling of organic and/or inorganic surfaces or luxury products with a composition comprising a polymer having a degree of polymerization n, ranging from 2 to 10000, and containing 2 to 10000 monomer units, said monomer units being:either monomer units derived from 4-vinylpyridine, in which the carbons in position 2 and 6 are substituted by one of the substituents of the following group: carboxylic acid, or ester of 1 to 20 carbons,or monomer units derived from a co-monomer,provided that said monomer units derived from 4-vinylpyridine represent at least 20% of the degree of polymerization n,said polymer being optionally complexed with a metal,said polymer being linear or cross-linked.
	abstract	The invention relates to a process for separating, in an aqueous medium, at least one actinide element from one or more lanthanide elements by using at least one molecule which sequesters the said actinide element to be separated and membrane filtration, the said process successively comprising: a) a step of bringing at least one molecule which sequesters the said actinide element in contact with the aqueous medium, the said molecule not being retained in the non-complexed state by the said membrane and being capable of forming a complex with the actinide element to be separated, comprising the said element and at least two of the said sequestering molecules, which complex is capable of being retained by the membrane; b) a step of passing the aqueous medium over the membrane in order to form a permeate on one side, comprising an aqueous effluent depleted of the said actinide element, and a retentate comprising the said complex.
	abstract	A method of producing a localized concentration of energy comprises creating at least one shockwave (10) propagating through a non-gaseous medium (8) so as to be incident upon a pocket of gas (2) within the medium (8). The pocket of gas (2) is attached to a surface (6) comprising a depression (4) shaped so as partially to receive the gas pocket (2). An apparatus for producing a localized concentration of energy comprises a non-gaseous medium (8) having therein a pocket of gas (2). The pocket of gas (2) is attached to a surface (6) comprising a depression (4) shaped so as partially to receive the gas pocket (2). The apparatus further comprises means for creating at least one shockwave (10) propagating through the medium (8) so as to be incident upon the pocket of gas (2).
	abstract	The present invention relates to a particle therapy apparatus used for radiation therapy. More particularly, this invention relates to a gantry for delivering particle beams which comprises means to analyze the incoming beam. Means are integrated into the gantry to limit the momentum spread of the beam and/or the emittance of the beam.
051026180	description	Referring to FIG. 1, tritium is extracted from heavy or light water rich in tritium by a reactor comprising a number of stages, four stages being shown for simplicity. Each of the stages comprises essentially an evaporator 10, a superheater 11, a catalytic reactor 12 and a condenser 13. These components of the various stages are interconnected by piping as shown in FIG. 1, and their respective functions are as follows. In the description which follows, specific reference is made to the extraction of tritium from tritium-rich heavy water according to the reaction EQU DTO+D.sub.2 .fwdarw.D.sub.2 O+DT However, it is to be understood that the apparatus is useful in the extraction of tritium from tritium-rich heavy or light water according to any of the following reactions: EQU DTO+D.sub.2 .fwdarw.D.sub.2 O+DT EQU HTO+H.sub.2 .fwdarw.H.sub.2 O+HT, EQU HTO+HD.fwdarw.HDO+HT, or more generally EQU QTO+Q.sub.2 .fwdarw.Q.sub.2 O+QT where Q denotes either of the hydrogen isotopes H and D. The tritium-rich heavy water (DTO) from a nuclear reactor, or heavy water (D.sub.2 O) from a preceding stage, is delivered into the evaporator 10 by a metering pump (not shown). The evaporator converts the water to steam which is mixed with tritium-lean deuterium (D.sub.2) delivered from the subsequent stage, or in the case of the final stage with tritium-lean deuterium from a supply tank. This mixture is fed to the superheater 11 where it is superheated to 200.degree. C. for subsequent catalytic exchange. The superheated gas-steam mixture passes to the catalytic reactor 12 where in contact with a platinum catalyst the following reaction takes place: EQU DTO+D.sub.2 .fwdarw.D.sub.2 O+DT Isotopic equilibration takes place and part of the tritium is transferred from the tritium-rich heavy water to the tritium-lean gas. The equilibrated mixture passes to the condenser 13 where the tritium-lean water is condensed and separated from the tritium-enriched gas. The condensed water is fed to the evaporator of the succeeding stage, or in the case of the final stage to the lean water return. The tritium-rich gas is fed back to the evaporator 10 of the preceding stage, or in the case of the first stage is fed to a cryo-distillation unit. As will be apparent from the scheme shown in FIG. 1, the design of such a plant based on convention practice requires extensive piping interconnecting the components of the various reactors, with many joints which are potential sources of leakage of radioactive substances such as tritium gas and tritiated water. According to the present invention the components of each stage are integrated into a single casing structure, thus eliminating the piping between its components. This construction has the further advantage of compactness to facilitate enclosure of the system in a secondary pressure vessel, as may be required for safety reasons in high tritium applications. As shown in FIG. 2, the casing structure is made up of a number of casing sections 20, 21, 22 and 23. The casing section 21 consists of a vertical steel cylinder having upper and lower flanges 24, 25 to which the casing section 20 having a flanged opening 26 and the casing section 22 having a flanged opening 27 are respectively bolted. The casing section 23 having a flanged opening 28 is bolted to a flanged opening 29 of the casing section 22. The four casing sections when joined together as illustrated form a unitary pressure vessel housing the components of the reactor. The joints between sections must be leaktight to prevent leakage of steam and gases; this may be achieved by seal welding the flanged joints between the sections. The vertical cylinder 21 houses the superheater 31, the catalytic reactor 32, and the condenser 33. Cooling fluid for the condenser 33 is supplied from a header within the casing section 20, the latter having inlet and outlet connections 34. The casing section 22 houses the evaporator 35, which in the present example comprises a tubular heat exchanger. The tubes of the heat exchanger extend horizontally, steam being admitted to the tubes from the casing section 23 having inlet and outlet connections 36, 37 for the steam. Tritium-rich water to be treated is admitted to the casing section 22 from the condenser of the preceding stage, or in the case of the first stage from the nuclear reactor itself, through an inlet pipe 38. The casing section 22 of the evaporator has a neck portion 39 defining a mixing and entrainment separation chamber 40 adjacent to the superheater 31. The neck portion 39 has an inlet 41 through which tritium-lean gas is supplied to the mixing chamber 40 from the subsequent stage or, in case of the final stage, from a source of deuterium. In operation of the reactor, the mixture of tritium-lean gas and steam from the evaporator section is passed to the superheater 31. The superheater 31 in the present embodiment of the invention comprises a tubular heat exchanger the tubes of which are vertically oriented within the casing 21. Heating fluid is supplied to the tubes of the heat exchanger via inlet and outlet connections 42, 43. It will be appreciated that the heating of the evaporator section and/or the superheater section of the reactor may alternatively be accomplished by electrical heating elements instead of heat exchangers supplied with heating fluids from external sources as shown. The superheated mixture of tritium-lean gas and tritium-rich steam passes to the catalytic reactor bed 32 at a temperature of approximately 200.degree. C. where an exchange of isotopes takes place. The tritium-lean gas receives tritium from the steam while part of the tritium of the steam is replaced by deuterium. The resultant mixture passes to the condenser 33. The condenser 33 comprises a vertically oriented tubular heat exchanger mounted within the upper part of the cylindrical casing 21. As previously mentioned, cooling fluid is supplied to the tubes of the condenser from a header within the casing section 20. Cooling of the steam/gas mixture results in separation of the condensed steam. The depleted water passes to the evaporator section of the subsequent stage, or in the case of the final stage to the lean water return, via an on output pipe 44. The tritium-enriched gas passes via an outlet pipe 45 to the mixing chamber of the preceding stage, or in the case of the first stage to the cyro-distillation unit. All the stages of the tritium separation plant are constructed in the same manner, the stages being interconnected as described so as to effect countercurrent flow of the deuterium and the tritiated water from stage to stage. Thus, tritium-rich water is fed to the evaporator section of the first stage, the depleted water being taken from the condenser section of the final stage, while tritium-lean deuterium gas is fed to the evaporator section of the final stage, the enriched deuterium being taken from the condenser section of the first stage and passed to the cryo-distillation unit. In a modification of the tritium separation plant described with reference to FIG. 2, the evaporator section of the first stage is omitted, the water to be treated being passed directly from the nuclear reactor to the evaporator section where it is mixed with the tritium-lean deuterium gas. An alternative VPCE reactor constituting one stage of the tritium separation plant is illustrated in FIG. 3. In this reactor the evaporator casing 22' is a vertical cylinder aligned with the casing section 21, the tubes of the evaporator 35' being vertically oriented within the casing section 22'. Otherwise, the construction of the reactor is essentially as described with reference to FIG. 2 and corresponding parts are denoted by the same reference numerals. FIG. 4 shows schematically an arrangement of reactors of the kind shown in FIG. 3, the essential components of the four stages and their various connections being denoted by the same reference numerals as are shown in FIGS. 1 and 3 to identify the corresponding parts. In this arrangement the first, second, third and fourth stages are identified by the references A, B, C and D respectively. The reactors of the four stages are vertically oriented, the condensed water flowing from the respective condenser sections under gravity. FIG. 5 shows schematically an alternative arrangement of reactors of the kind shown in FIG. 3, wherein the reactors are inverted so that the evaporators are at the top of each stage and the condensers are at the bottom. In this arrangement pumps 47 are used to pump the condensed water from each stage to the evaporator of the subsequent stage. This provides convenient flexibility in the layout of the tritium separation plant and it will be particularly noted that the reactors of the various stages may be arranged horizontally instead of vertically as previously described. FIG. 6 shows schematically yet another alternative arrangement in which the reactors are of the kind shown in FIG. 2, but in which the evaporators 10 are omitted from the integrated casing structures of the various stages and replaced by independent evaporator sections 50 which are interconnected with the respective superheater sections 12 by piping 51. As shown in the fragmentary view in FIG. 7, the evaporator section of the first reactor stage may be omitted altogether, the water to be treated being received directly from the nuclear reactor as steam and fed directly to the superheater 11 of the first stage by piping 42. As previously mentioned, the reactors are useful in the extraction of tritium from heavy or light water by catalytic reaction with either or both of the hydrogen isostopes (H) and (D). In the following claims the term "water" means heavy or light water and the term "hydrogen" means light hydrogen (H) or deuterium (D).
	claims	1. A radiation image conversion panel comprising a substrate, a step-cured resin layer subjected to a step-by-step curing process and formed on a main face of the substrate, and a phosphor formed during a curing reaction of the resin layer by vapor deposition on a main face of the resin layer on a side opposite from the substrate and made of a columnar crystal;wherein at least the main face of the resin layer has a surface energy of at least 20[mN/m] but less than 35 [mN/m]. 2. A radiation image conversion panel according to claim 1, wherein the resin layer is formed by a silicone resin. 3. A radiation image conversion panel according to claim 1, wherein the surface energy of at least the main face of the resin layer is set according to the crystallinity of the phosphor. 4. A radiation image conversion panel according to claim 1, wherein the substrate is a radiation-transmitting substrate. 5. A radiation image conversion panel according to claim 1, wherein the substrate is a fiber optic plate. 6. A radiation image conversion panel according to claim 1, wherein the substrate is an image sensor. 7. A radiation image conversion panel comprising a substrate, a step-cured resin layer subjected to a step-by-step curing process and formed on a main face of the substrate, and a phosphor formed by during a curing reaction of the resin layer vapor deposition on a main face of the resin layer on a side opposite from the substrate and made of a columnar crystal;wherein the phosphor is constructed such that the columnar crystal has a pillar form on the substrate side. 8. A radiation image conversion panel comprising a substrate, a step-cured resin layer subjected to a step-by-step curing process and formed on a main face of the substrate, and a phosphor formed during a curing reaction of the resin layer by vapor deposition on a main face of the resin layer on a side opposite from the substrate and made of a columnar crystal;wherein the columnar crystal on the substrate side of the phosphor has a pillar form substantially equal to that of the columnar crystal on the opposite side. 9. A radiation image conversion panel comprising a fiber optic plate constituted by a bundle of a plurality of optical fibers and a phosphor made of a columnar crystal;the radiation image conversion panel further comprising a heat-resistant step-cured resin layer subjected to a step-by-step curing process and formed on a main face of the fiber optic plate and transparent to output light emitted from the phosphor;wherein the phosphor is formed during a curing reaction of the resin layer by vapor deposition on a main face of the heat-resistant resin layer on a side opposite from the fiber optic plate. 10. A radiation image conversion panel according to claim 9, wherein the resin layer is formed by a silicone resin. 11. A radiation image conversion panel according to claim 9, wherein the surface energy of at least the main face of the resin layer is set according to the crystallinity of the phosphor. 12. A radiation image conversion panel comprising:a fiber optic plate constituted by a bundle of a plurality of optical fibers;a heat-resistant step-cured resin layer subjected to a step-by-step curing process and formed on a main face of the fiber optic plate; anda phosphor formed during a curing reaction of the resin layer by vapor deposition on a main face of the heat-resistant layer on a side opposite from the fiber optic plate and made of a columnar crystal;wherein the phosphor is constructed such that the columnar crystal has a pillar form on the fiber optic plate side. 13. A radiation image conversion panel comprising:a fiber optic plate constituted by a bundle of a plurality of optical fibers;a heat-resistant step-cured resin layer subjected to a step-by-step curing process and formed on a main face of the fiber optic plate; anda phosphor formed during a curing reaction of the resin layer by vapor deposition on a main face of the heat-resistant layer on a side opposite from the fiber optic plate and made of a columnar crystal;wherein the columnar crystal on the fiber optic plate side of the phosphor has a pillar form substantially equal to that of the columnar crystal on the opposite side.
	claims	1. An inspection apparatus for inspecting a wafer through imaging, the inspection apparatus comprising:an electron optical system configured to radiate an electron beam on a certain area of a wafer corresponding to an inspection region of the wafer, the electron optical system including a time delay integration (TDI) sensor, an image processor, a controller, and a wafer stage configured to move the wafer at a set speed, the TDI sensor being configured to detect an image of the wafer upon radiation of the electron beam on the certain area of the wafer and transmit the image as image signals to the image processor, the image processor being configured to receive the image signals, perform an inspection of the inspection region based on the image signals, and output a result of the inspection, the controller being configured to control operations of the electron beam, the TDI sensor, the image processor, and the wafer stage according to a set of conditions; anda display device coupled to the electron optical system and configured to display an indication of the result of the inspection output by the image processor, andwherein the controller is configured to implement a graphical user interface (GUI) that includes a graph element displayed via the display device, the graph element displaying a respective relationship among values of S and D for each of a plurality of values of P on a graph with respect to a specified value of L by assuming D and S as coordinates axes, where each of value of S defines an area of the wafer to be inspected per unit time, each value of D defines a width in a moving direction of the wafer that corresponds to a unit pixel of the image, each value of P defines an image signal acquisition frequency for image signal acquisition cycles, and each value of L defines a length in a direction perpendicular to the moving direction of the wafer for which image data is obtained in each image signal acquisition cycle in a range of an inspection image in the region on which the electron beam is radiated, such that a formula S=D×L×P is satisfied for the respective relationships displayed on the graph,wherein the GUI includes a pointer element displayed on the display screen and accessible by a user such that the user can move the pointer within the graph and select a position on the graph,wherein the graph element is implemented to display the values of D and S for each position of the pointer within the graph according to the specified value of L;wherein the controller is configured to control the electron optical system according to the values of D and S for the position on the graph selected by the user, andwherein the controller is configured to disregard the values for S, L, D, and P that are selected upon the pointer being positioned on the graph when at least one of the values is out of a range in which the apparatus is configured to operate. 2. The apparatus, according to claim 1, wherein the electron beam is a planar electron beam. 3. The apparatus, according to claim 1, wherein the electron optical system is a projection optical system in which the TDI sensor is configured to obtain the image signals to transmit by projecting the inspection region of the image during detection. 4. The apparatus, according to claim 1, wherein the graph element is displayed with a plurality of lines in the graph that each correspond to a respective value of P with respect to a set value of L. 5. The apparatus, according to claim 4, wherein the GUI is implemented to permit the user to specify the value of L, and wherein the graph element is implemented such that the graph is changed corresponding to the value of L specified by the user in accordance with the formula of S=D×L×P. 6. The apparatus, according to claim 1, wherein the GUI is implemented such that a maximum value of L that can be specified is 200 microns. 7. The apparatus, according to claim 1, wherein the values of D corresponding to positions on the graph are within a range between 0 nm and 250 nm. 8. The apparatus, according to claim 1, wherein the controller holds a numerical table in which a plurality of sets of conditions for operating the electron optical system each corresponding to a respective pair of D and L values are set beforehand and determines a set of conditions for operating the electron lens according to values of D and L specified by the user via the GUI with reference to the numerical table. 9. An inspection apparatus for inspecting a wafer through imaging, the inspection apparatus comprising:an electron optical system configured to radiate an electron beam on a certain area of a wafer corresponding to an inspection region of the wafer, the electron optical system including a time delay integration (TDI) sensor, an image processor, a controller, and a wafer stage configured to move the wafer at a set speed, the TDI sensor being configured to detect an image of the wafer upon radiation of the electron beam on the certain area of the wafer and transmit the image as image signals to the image processor, the image processor being configured to receive the image signals, perform an inspection of the inspection region based on the image signals, and output a result of the inspection, the controller being configured to control operations of the electron beam, the TDI sensor, the image processor, and the wafer stage according to a set of conditions; anda display device coupled to the electron optical system and configured to display an indication of the result of the inspection output by the image processor, andwherein the controller is configured to implement a graphical user interface (GUI) that includes a graph element displayed via the display device, the graph element displaying a graph related to a time delay integration formula S=D×L×P with respect to a plurality of values of P and a specified value of L by assuming D and S as coordinates axes, where each of value of S defines an area of the wafer to be inspected per unit time, each value of D defines a width of the wafer in a direction image in the region on which the electron beam is irradiated in each image acquisition cycle that corresponds to a unit pixel of the image, each value of P defines an image signal acquisition frequency for image signal acquisition cycles, and each value of L defines a length in the direction image in the region on which the electron beam is radiated in each image acquisition cycle,wherein the coordinate axis is S or S′ defined as a number of wafers to be processed per hour, an inverse number to define a processing time of one wafer or a time required for a unit area inspection by calculating S, andwherein the GUI includes a pointer element displayed on the display screen and accessible by a user such that a user can move the pointer within the graph and select a position on the graph,wherein the graph element is implemented to display the values of D and S or S′ for each position of the pointer within the graph according to the specified value of L;wherein the controller is configured to control the electron optical system according to the values of D and S or S′ for the position on the graph selected by the user, andwherein the controller is configured to disregard the values for S, L, D, and P that are selected upon the pointer being positioned on the graph when at least one of the values is out of a range in which the apparatus is configured to operate. 10. The apparatus, according to claim 9, wherein the electron beam is a planar electron beam. 11. The apparatus, according to claim 9, wherein the electron optical system is a projection optical system in which the TDI sensor is configured to obtain the image signals to transmit by projecting the inspection region of the image during detection. 12. The apparatus, according to claim 9, wherein the graph element is displayed with a plurality of lines in the graph that each correspond to a respective value of P with respect to a set value of L. 13. The apparatus, according to claim 12, wherein the GUI is implemented to permit the user to specify the value of L, and wherein the graph element is implemented such that the graph is changed corresponding to the value of L specified by the user in accordance with the formula of S=D×L×P. 14. The apparatus, according to claim 9, wherein the GUI is implemented such that a maximum value of L that can be specified is 200 microns. 15. The apparatus, according to claim 9, wherein the values of D corresponding to positions on the graph are within a range between 0 nm and 250 nm. 16. The apparatus, according to claim 9, wherein the controller holds a numerical table in which a plurality of sets of conditions for operating the electron optical system each corresponding to a respective pair of D and L values are set beforehand and determines a set of conditions for operating the electron lens according to values of D and L specified by the user via the GUI with reference to the numerical table.
	abstract	A component cooling water system for a nuclear power plant. In one embodiment, the system includes an inner containment vessel housing a nuclear reactor and an outer containment enclosure structure. An annular water reservoir is formed between the containment vessel and containment enclosure structure which provides a heat sink for dissipating thermal energy. A shell-less heat exchanger is provided having an exposed tube bundle immersed in water held within the annular water reservoir. Component cooling water from the plant flows through the tube bundle and is cooled by transferring heat to the annular water reservoir. In one non-limiting embodiment, the tube bundle may be U-shaped.
040574679	summary	BACKGROUND OF THE INVENTION The present invention concerns a pressurized-water reactor whose primary cooling system gives off its heat within a pressure vessel to a steam generator within the vessel and which is inserted between the reactor core and the intake space of the circulating pumps, i.e., in the cold leg of the main coolant loop. This pressurized-water reactor, whose primary cooling system is normally operated with forced circulation, is particularly well suited for ship propulsion and stationary installations of smaller power rating. In order to avoid damage to the circulating pumps of the primary cooling system due to cavitation, particularly in the case of rapid load changes, these pumps should operate in a zone as cool as possible, of the primary cooling system. If these pumps operate at close to the boiling temperature of the pressurized-water coolant, a small brief lowering of the normal pressure is sufficient to cause the water in the pump housings to boil. The steam bubbles produced in this manner collapse during any subsequent pressure increase and cause destruction at the pump internals in a locally closely confined area. Particularly exposed in this respect are circulating pumps which, for reasons of better maintenance and inspection, are preferably located above the core and steam generator, because as compared to pumps located underneath, such pumps receive a considerably smaller inflow head or hydrostatic pressure and therefore have a greater tendency toward cavitation. These problems are often avoided by means of an external pressurizer for the coolant which, however, requires much space and is expensive. A typical example for a shipboard reactor proven in practice is represented by the propulsion plant of the nuclear vessel "OTTO HAHN". In this arrangement, with a pump located at the bottom of the pressure vessel, the desired flow can be realized with water as cold as possible ahead of the pump and with a considerable inflow head to the pump. If, with the design the same otherwise, the circulating pump or pumps are arranged at the upper end of the pressure vessel, e.g., in its closure head end, the pumps not only are approached by the hotter primary coolant, but the inflow head is at the same time also reduced considerably. In the disclosures of the German Offenlegungsschrift No. 22 27 895 (U.S. Ser. No. 153,304, filed June 15, 1971 now U.S. Pat. 3,888,734) the described problems with the circulating pumps are avoided by using a very large and, therefore, also expensive external pressurizer. In this external pressurizer, the pressure in the pressure vessel is adjusted to the desired level by controlled heating of the water in the pressurizer or injection of cold water. In the same disclosures, it is proposed to support all the internals in the pressure vessel at the vessel's upper closure head, so that for every refueling and every inspection or maintenance action on internals in the pressure vessel, practically all the internals must be removed together from the pressure vessel, with their considerable weight. Such refueling is necessary about every one to two years, while a routine inspection is necessary only every 8 years. It is therefore desirable that the fuel assemblies can be exchanged without the necessity to remove the steam generator. In order to obtain the desired flow path from the reactor through the heat exchanger to the circulating pump, it is proposed in the same disclosures to have the flow through the reactor core go in normal operation from top to bottom. This flow path results, in the case of load changes or at low load as well as in the event of a failure of the circulating pumps, in very confusing flow conditions in the reactor core which should be avoided. An object of the present invention is to provide a pressurized-water reactor whose primary cooling system gives off its heat within a pressure vessel, to a steam generator which within the pressure vessel is inserted between the reactor core and the suction space of the circulating pumps, and which can be disassembled without destruction for inspecting the pressure vessel. A specific object of the present invention is to provide an arrangement of a steam generator which can take up the decay heat of the reactor core through natural circulation in the event of a failure of the circulating pumps and lowering of the water level. A further object of the present invention is to provide a pressurized-water reactor whose internals in the pressure vessel, fabricated from austenitic metal, can expand freely relative to the pressure vessel, which is made of ferritic metal. SUMMARY OF THE INVENTION According to the present invention, the annular steam generator is arranged in a hollow cylindrical housing, which is closed on all sides excepting for openings for directing the coolant flow, and which is supported on the upper end of the pressure vessel via a hollow ring supported by the vessel; a core support structure with an assembled reactor core being likewise suspended by this housing; and this housing being radially braced at the lower end via several vertically sliding surfaces against the inner wall of the pressure vessel. With this arrangement, all important internals within vessel are suspended in the pressure vessel with the exception of the pumps and the feed water and steam nozzles, by the housing, the hollow ring being firmly clamped at the upper end of the pressure vessel between a step or internal flange formed by the pressure vessel and the inner periphery of closure head of the pressure vessel. This housing is made of austenitic metal but can freely expand downward within the ferritic metal pressure vessel, and it is supported at its lower end at several places, vertically sliding at the inner wall of the pressure vessel, so that forces in the horizontal direction can be taken up by the pressure vessel. The core housing is fastened on the one hand at the already described steam generator housing and can be centered with its lower end by a likewise vertically sliding post, so that is follows all expansions of the housing in the axial direction and transmits in the horizontal direction the occurring forces to the pressure vessel. In this manner, all welded steam for lugs or other fastening means, which cannot be checked, are avoided at the pressure vessel and the ultrasonic testing of the pressure vessel wall is facilitated. All the pressure vessel internals may be fabricated of the same material, so that stresses induced by unequal expansion are eliminated. The arrangement of the steam generator in a cylindrical housing closed on all sides not only has the advantage that the thin tubes of this steam generator can be protected better against vibrations, but it also protects this steam generator against damage during installation and disassembly. The hollow ring of box-like cross section serves, on the one hand, to stiffen the steam generator housing with all internals attached thereto, and, on the other hand, serves to guide the flow. The inclined positions, which are unavoidable in a shipboard reactor, do not cause trouble with the water circulation either in normal operation or in the event of a failure of the pumps, as they can have an effect only in the relatively narrow, ring-shaped steam generator housing. Thus, the structural height of the pressure vessel above the pumps can be decreased considerably. In a further embodiment of the invention, the hollow ring forms the pressure chamber of the primary circulation pumps and is penetrated by several inlet sections which communicate with the intake space of the circulating pumps. In this manner, each pump has its own inlet section favorably shaped from a flow point of view; all the pumps together, however, push into the hollow ring of box-shaped cross section which has several openings at its underside, so that the primary cooling water can flow to the reactor core, distributed uniformly over the circumference, also in the event of a failure of a pump.
	summary
055770819	claims	1. A method of forming a nuclear fuel assembly grid, which includes the steps of (a) preparing a plurality of formed alloy straps each having slits, (b) arranging the straps into a grid form by intersecting the straps with each other through the slits, and (c) brazing intersections of the associated straps, wherein said method further comprises the step of: (d) subjecting those portions to be brazed to a pretreatment prior to said arranging step (b), said pretreatment including applying a paste comprising a mixture of a filler metal and a vehicle, to the portions to be brazed to form a thin film thereon. 2. The method according to claim 1, wherein said pretreatment step (d) further includes drying the paste film, heating the straps in a vacuum furnace to melt the filler metal in the paste, and cooling the straps; and wherein said brazing step (c) includes placing a filler metal daub on top of the intersections of the straps, and heating the straps in a vacuum furnace to effect the brazing of the straps. 3. The method according to claim 1, further comprising the steps of drying the paste film, heating the straps in a vacuum furnace to melt the filler metal in the paste, and cooling the straps between said arranging step (b) and said brazing step (c); wherein said brazing step (c) includes placing a filler metal daub on top of the intersections of the straps, and heating the straps in a vacuum furnace to effect the brazing of the straps. 4. The method according to claim 1, wherein said brazing step (c) includes placing a filler metal daub on top of the intersections of the straps, and heating the straps in a vacuum furnace to effect the brazing of the straps; said method further comprising the steps of drying the paste film after said pretreatment step (d), subsequently effecting said brazing step (c), and heating the straps in a vacuum furnace to effect the brazing of the straps. 5. The method according to claim 1, wherein said brazing step (c) includes placing a filler metal daub on top of the intersections of the straps immediately after said arranging step (b), and heating the straps in a vacuum furnace to effect the brazing of the straps. 6. The method according to claim 1, wherein said paste further contains an additional substance selected from the group consisting of water, organic solvent, and liquid surface active agent. 7. The method according to claim 1, wherein said pretreatment step (d) comprises employing screen printing. 8. A grid for a nuclear fuel assembly, formed by the application of the method according to any one of claims 1 to 7.
	abstract	Application of axial seed magnetic fields in the range 20-100 T that compress to greater than 10,000 T (100 MG) under typical NIF implosion conditions may significantly relax the conditions required for ignition and propagating burn in NIF ignition targets that are degraded by hydrodynamic instabilities. Such magnetic fields can: (a) permit the recovery of ignition, or at least significant alpha particle heating, in submarginal NIF targets that would otherwise fail because of adverse hydrodynamic instability growth, (b) permit the attainment of ignition in conventional cryogenic layered solid-DT targets redesigned to operate under reduced drive conditions, (c) permit the attainment of volumetric ignition in simpler, room-temperature single-shell DT gas capsules, and (d) ameliorate adverse hohlraum plasma conditions during laser drive and capsule compression. In general, an applied magnetic field should always improve the ignition condition for any NIF ignition target design.
	abstract	A method of preventing separation of a feedwater sparger end bracket assembly comprises installing first and second clamp members of a clamp over a feedwater sparger end bracket assembly in opposition to one another, and leaving the clamp in place to constrain the feedwater sparger end bracket assembly against separation in first, second and third directions.
047117575	claims	1. Position indicating device for producing an indication of the position of a displaceable structure, said device comprising: a position representing member mounted for movement in response to displacement of the structure, said member being movable along a defined path from a starting point such that the distance to which said member extends from the starting point corresponds to the position of the structure; a plurality of sensing elements spaced apart along the defined path such that each said element is associated with a respective location along the defined path, each said element being operative to respond to the presence of said member when said member extends from the starting point to the respective location associated with that said element; means operatively coupling said elements into respective pairs of elements, said means having, for each pair of elements, an output producing a signal only when a single element of its respective pair is responding to the presence of said member, said elements of each said pair being spatially separated from one another by at least one said element of a different said pair; a plurality of signal producing members each operative for producing a signal representing a predetermined logic state in response to a predetermined input signal, the number of said signal producing members being smaller than the number of said sensing elements; and circuit means operatively connecting said outputs to said signal producing members for causing a signal at each said output to produce a predetermined input signal at a corresponding signal producing member and for causing a predetermined input signal to be produced at at least one said signal producing member whenever a signal is present at either one of at least two of said outputs. 2. A device as defined in claim 1 wherein said at least two outputs producing a predetermined input signal at said at least one signal producing member are associated with respective element pairs which are physically spaced apart along the defined path such that a signal is produced at only one of said at least two outputs at any time. 3. A device as defined in claim 2 wherein each said sensing element is an annular coil and said position representing member is of a material which varies the effective impedance of each said coil when said member extends to the location associated with that said coil. 4. A device as defined in claim 3 wherein said circuit means comprise a plurality of impedance elements each connected between one said output and one said signal producing member. 5. A device as defined in claim 4 wherein the number of said impedance elements is greater than the number of said signal producing devices. 6. A device as defined in claim 5 wherein each said impedance element is a resistor. 7. A device as defined in claim 1 further comprising at least one reference element corresponding in structure to each said sensing element and disposed at a location removed from the defined path, and wherein said means operatively coupling said elements couple said reference element to a respective one of said sensing elements. 8. A device as claimed in claim 1 in combination with a nuclear reactor having at least one operating element and a movably mounted drive rod supporting said operating element, wherein said movably mounted drive rod and said operating element constitute said displaceable structure. 9. A device as defined in claim 1 wherein said circuit means connect at least a first one of said signal producing members to at least two of said outputs and connect at least a second one of said signal producing members to only one of said outputs.
	description	1. Field of the Invention The present invention generally relates to apparatus and techniques for vaporizing metallic elements or salts and further temperature controlled handling and processing of the metal or metal salt vapors for use in material processing operations, for example, semiconductor processing. The further processing may include ionization for possible subsequent ion implantation, or participation in a deposition process. 2. Description of Related Art A variety of techniques and apparatuses are known for modifying or coating substrates to form new compositions. For example, ion implantation is used to implant ions into substrates to create new materials. Also, deposition processes, such chemical vapor deposition (CVD) and physical vapor deposition (PVD or sputtering) are well known to practitioners of materials science, for example, in the semiconductor processing arts, for depositing layers of different materials on substrates. Such techniques and the resulting compositions have proven useful in the semiconductor processing arts for a variety of purposes. For example, ion implantation has been used to dope semiconductor substrates and CVD and PVD processes have been used to deposit polysilicon or barrier layer materials. With continued miniaturization of integrated circuits, new materials and processing techniques with improved performance characteristics in smaller device size realms are constantly being sought. Accordingly, techniques and apparatuses for the fabrication of new materials with uses in semiconductor processing are desirable. Also, techniques and apparatuses with improved process control enabling the more precise tailoring of a material for particular purpose would be desirable. Such apparatuses and techniques would find use in the semiconductor processing context and beyond. The present invention provides techniques for vaporizing a metallic element or metallic element salt with a heated inert carrier gas and further temperature controlled handling and processing of the metal or metal salt vapors for use in material processing operations, for example, semiconductor processing. The vaporized metallic element or salt is then transported to a temperature controlled ionization chamber where the vaporized metal or salts are ionized to generate metal ions. The metal ions may be extracted from the ionization chamber as an ion beam for use in materials processing operations, such as ion implantation in a semiconductor device fabrication process. Apparatus for accomplishing these techniques, which include vaporizing chambers with carrier gas heating capability and heated ionization chambers are also provided. Suitable metallic elements include alkaline earth elements (Group II metals) or any metallic element or salt with relatively high vapor pressure (e.g., at least 5 mTorr) at relatively low temperature (e.g., less than about 1000° C.), for example, alkaline earth elements (Group II metals) such as Ca, Ba or Sr. Certain transition metals, such as transition metals with vapor pressures greater than 0.01 mTorr at temperatures below 1000° C., (e.g., Cd, Zn or Mn), or certain metal salts (e.g., CaCl2, CaBr2, NbCl5, or ZrCl4) may be used in various aspects of the invention. In one aspect, the invention pertains to a method of generating an metallic ion source. The method involves heating an inert carrier gas, vaporizing a metallic element or metallic element salt in the presence of the heated inert carrier gas, transporting the vaporized metallic element or salt in the heated inert carrier gas to a temperature-controlled ionization chamber, and ionizing the vaporized metallic element or salt in the chamber in the presence of the heated inert carrier gas to generate ions of the metal. In another aspect, the invention pertains to an apparatus for vaporizing and ionizing a metallic element or metallic element salt. The apparatus includes a carrier gas heating chamber configured to heat an inert carrier gas to a temperature in the range of 100 to 1000° C., and a vaporizer chamber, connected with the carrier gas heating chamber, and configured to vaporize a metallic element in the presence of the inert carrier gas heated in the carrier gas heating chamber. An ionization chamber is connected with the vaporizer chamber, the ionization chamber having surfaces heated to prevent deposition of the vaporized metallic element or salt thereon, and configured to ionize the vaporized metallic element in the presence of the inert carrier gas heated in the carrier gas. The ionization chamber surfaces may contain or contact resistive heating elements. A photo-ionization source may be appended to the ionization chamber. The interior surfaces of the ionization chamber have a mirror finish. These and other aspects and advantages of the present invention are described below where reference to the drawings is made. It is to be understood that, in the drawings, like reference numerals designate like structural elements. Also, it is to be understood that the depictions in the figures are not necessarily to scale. In the following description, numerous specific details are set forth to provide a thorough understanding of the present invention. It will be apparent, however, to one skilled in the art that the present invention may be practiced without some or all of these specific details. In other instances, well known process steps have not been described in detail in order to avoid unnecessarily obscuring the present invention. Introduction The present invention provides techniques for vaporizing a metallic element or metallic element salt with a heated inert (i.e., noble or rare) carrier gas and further temperature controlled handling and processing of the metal or metal salt vapors for use in material processing operations, for example, semiconductor processing. In one embodiment, the vaporized metallic element or salt can be applied to the substrate by deposition to form composites, for example with very thin metal or metal salt coatings. Apparatus for accomplishing these techniques, which include vaporizing chambers with carrier gas heating capability and heated deposition chambers are also provided. In another embodiment, the vaporized metallic element or salt is then transported to a temperature controlled ionization chamber where the vaporized metal or salts are ionized to generate metal ions. The metal ions may be extracted from the ionization chamber as an ion beam for use in materials processing operations, such as ion implantation or plasma immersion (PLAD) in a semiconductor device fabrication process. Apparatus for accomplishing these techniques, which include vaporizing chambers with carrier gas heating capability and heated ionization chambers are also provided. FIG. 1 is a flowchart of a generic method of vaporizing a composition by applying a vaporized metallic element to a substrate in accordance with one embodiment of the present invention. The method (100) involves heating an inert carrier gas (101), for example to a temperature sufficient to produce a vapor pressure of the metallic element or salt of at least 0.01 mTorr, such as between about 100 and 1000° C., and vaporizing the metallic element or metallic element salt with the heated inert carrier gas (103). The vaporization of the metal or salt could include heating the metallic element or metallic element salt to the same temperature as the heated inert carrier gas, such as in a crucible, for example. The result is a vaporized metallic element or salt together with the heated inert carrier gas. The vaporized metallic element or salt in the presence of the heated carrier gas may then be handled, for example transported, by the heated carrier gas which retains it in the vapor phase during handling. In this way, the vaporized metallic element or salt is then provided to a temperature-controlled chamber (105), which may be heated to about the same temperature as the carrier gas or may use the heated carrier gas to maintain a sufficiently high temperature that deposition on the walls of the chamber is suppressed. In a preferred embodiment, the chamber sidewalls are heated, preferably resistively by resistive heating elements embedded in or applied to the chamber sidewalls. Suitable metallic elements include alkaline earth elements (Group II metals) or any metallic element or salt with relatively high vapor pressure (e.g., at least 5 mTorr) at relatively low temperature (e.g., less than about 1000° C.), for example, alkaline earth elements (Group II metals) such as Ca, Ba or Sr. Certain transition metals, such as transition metals with vapor pressures greater than 0.01 mTorr at temperatures below 1000° C., (e.g., Cd, Zn or Mn), or certain metal salts (e.g., CaCl2, CaBr2, NbCl5, or ZrCl4) may be used in various aspects of the invention. The metals or metal salts are vaporized in accordance with the present invention in the presence of a heated inert carrier gas, such as Ar or Ne. The carrier gas is heated to a temperature sufficient to vaporize and retain in the vapor state enough of the metal or salt for the desired application purpose. One of the advantages of the present invention is that metal or metal salt vapor pressures sufficient for well controlled deposition of thin films (e.g., as thin as one atom thick) or relatively shallow concentrated implants may be achieved at relatively low temperatures, for examples temperatures below 1000° C. Apparatus for accomplishing these techniques include carrier gas heating chambers, vaporizers and heated processing chambers, including, in various embodiments, ionization and deposition chambers. In the case of ionization, the apparatus includes a component for ionizing the metallic vapors without causing shorting of the system due to byproduct metal deposition. In the case of deposition, the apparatus includes a component for cooling substrate in the substrate holder in order to induce deposition and solidification of the metal or metal salt vapors on the substrate. Ionization In one aspect of the preset invention, vaporized metals may further be ionized and the ions then used in an ion implantation process to create a new material. FIG. 2 of a method of vaporizing a metallic element or salt, and handling and ionizing a vaporized metallic element in accordance with the present invention. The method (200) involves heating an inert carrier gas (101), for example to a temperature at or above the vaporization temperature of a metal or metal salt, for example between about 100 and 1000° C., and vaporizing a metallic element or metallic element salt in the presence of the heated inert carrier gas (103). The vaporized metallic element or salt in the heated inert carrier gas is then provided to a temperature-controlled ionization chamber (205). Next, the vaporized metal is ionized (207). Ionization chambers normally ionize material by means of thermo-ionic emission sources and high voltages, e.g., a filament, cathode, and anode. However, these sources are susceptible to shorting due to metal deposition from the gaseous mixture onto electrical components. In accordance with the present invention, the ionization is conducted by a process that is not susceptible to shorting due to deposition of metal on its components. Also, since the metal or metal salt vapor is conveyed by a carrier gas, the technique should ionize the metal in the vapor without ionizing the carrier gas. In a specific embodiment, photo-ionization is used. Microwave and laser sources, which ionize gaseous material, are known and can be used. Also, electron cyclotron resonance sources used in the nuclear industry for very heavy ions (CERN) may be used. Both of these methods can be considered special cases of photo-ionization. By careful selection of wavelengths, selective ionization can occur. In a specific embodiment, the photo-ionization is conducted with a bright light source having a wavelength sufficient to ionize the metallic element but not the inert carrier gas. The ionization energy relates to photo-ionization according to the following relationship wavelength=12398/E(eV) in units of angstroms (Å). As an example, Ca will be selectively ionized and Ne will not be ionized by light having a wavelength(s) in the range of 600 to 2028 Å (e.g., about of about 600 to 2100 Å). The light source can be filtered to remove wavelengths low enough to ionize the inert carrier gas, for example, of less than about 619 Å. An ion beam can be extracted from the chamber using an extractive electrode assembly to generate an ion beam. The extraction electrode assembly should be very close to the chamber and extract only the metal ions. Unionized metal and the carrier gas are removed from the chamber. The ion beam may be subjected to a magnetic mass analysis and any required acceleration/deceleration for obtaining the desired energy for subsequent applications. When selective ionization is performed, such as in the case of Ca in a Ne carrier gas as described above, no mass analysis is required and the ion beam can go directly to the substrate. Mass analysis may still be performed if a change in energy from the extraction potential is desired. Following ionization, the ion beam generated may be applied to a substrate to implant metal ions into the substrate. The substrate may be composed of organic or inorganic materials or a combination. A particularly useful substrate in the context of semiconductor processing is silicon or a metal oxide dielectric material, such as SiO2, ZrO2 or HfO2. In specific embodiments, the ionized metallic element is Ca, Sr, Ba, Mn, Cd, or Zn, and the substrate is SiO2, ZnO or HfO2. In a preferred embodiment, the ionized metallic element is Ca and the substrate is SiO2. The metal ions are applied to a substrate in an ion implantation chamber via the ion beam extracted from the ionization chamber. Such chambers are well know in the art, for example the Axcelis GSD ion implantation system available from Axcelis Technologies, Beverley, Mass. Other suitable ion implanters are available from Varian, Inc. and Applied Materials, Inc. Pressures in the ionization chamber are less than 500 mTorr, and as low as 8E-7 Torr. Actual pressures may vary over a large range, depending on the application. Implant energies are a function of the extraction potential or substrate potential. Energies of ion beams can vary depending on the mechanism for extraction but could be as low as 10 eV and high as 20 MeV. The ion source could be attached as a source to any number of implantation configurations. Other than those parameters specifically noted, standard ion implantation techniques, known to those of skill in the art, may be used. Ion implantation in accordance with the present invention may be used to fabricate new materials with beneficial properties. For example, for a 2 KV extraction a composition with a Ca implant peak around 45 Å deep in SiO2 can be produced. For a low energy implantation system, an energy of 200 eV would produce a Ca peak around 13 Å deep in SiO2. A concentration of about 8–10 atomic percent can be achieved with doses in range the range of about 1E15/cm2 to 1E17/cm2. Also, by implanting Hf at 200 eV into a SiO2 substrate, an implanted layer of about 40 Å depth is created. The implant is characterized by a maximum concentration of the metal located about 22 Å into the oxide substrate with a straggle (analogous to one standard deviation for the metal distribution) of about 6 Å. Hence, for an implant dose of 1E12/cm2, a peak concentration of about 6.6E18 atoms/cm3 would be created within the oxide layer. Such materials may be useful in semiconductor processing, for example as a barrier layer (for example a barrier to B) or dielectric. FIG. 3 is a block diagram of an apparatus for conducting vaporization, handling and ionization in accordance with the present invention. The apparatus may, for example, be a rapid thermal anneal (RTA) reactor adapted to the purposes of the present invention. An apparatus 300 in accordance with the present invention includes a carrier gas heating chamber 302 configured to heat an inert carrier gas to a temperature at or above the vaporization temperature of a particular metallic element or salt, generally in the range of about 100 to 1000° C. An inert carrier gas is supplied to the heating chamber 302 via an inlet 301. For ionization implementations, a carrier gas that is not readily ionized by a photo-ionization process, for example neon (Ne), is used. The carrier gas heating chamber 302 is connected with a metal vaporizer chamber 304. The vaporizer 304 is configured to vaporize a metallic element in the presence of the inert carrier gas heated in the carrier gas heating chamber 302. The vaporizer 304 may include a resistive heater and crucible to facilitate vaporization of the metal or salt. The heated carrier gas in the heating chamber 302 is maintained at the same temperature as the vaporizer 304 by a control feedback loop 305. After vaporization, the vaporized metal or salt is held in the vapor state by the heated carrier gas for handling, for example, transport to a another chamber for use in materials processing operations. In one embodiment, depicted in FIG. 3, the vaporizer chamber 304 is connected with a temperature controlled (i.e., heated) ionization chamber 306. The ionization chamber 306, and its surfaces are heated to discourage deposition of the vaporized metallic element on them. The chamber surfaces may be heated by resistive heating elements applied to or embedded in the chamber's walls 307 and/or, at least in part, by the heated carrier gas itself. The chamber 306 is maintained at a suitable temperature to maintain the vapor state of the metal or salt and pressure for ionization, e.g., about 100 to 1000° C. and a moderate pressure of less that about 500 mTorr and as low as 8E-7 Torr. The vaporized metal or salt is conveyed from the vaporizer 304 to the ionization chamber 306 by the heated carrier gas via an inlet 308. The ionization chamber 306 ionizes the metal or metal salt vapor by a photo-ionization process. The ionization chamber has an appended a light source 308, as described above, separated from the carrier gas and vapor in the chamber by a transparent sheet 309, generally composed of glass. The light from the source 308 passes through the glass 309 and ionizes the metal vapor without ionizing the carrier gas. The light source may be filtered to remove wavelengths undesirable wavelengths, such as those low enough to ionize the carrier gas. As noted above, any surface inside the chamber 306 exposed to the gas and vapor where deposition is undesirable is heated to a temperature at or above the vaporization temperature of the metal or salt to discourage deposition. In this regard, the glass 309 separating the light source 308 from the gas and metal vapor is heated by the bright light source and by thermal conduction from its contact with the heated chamber walls 307. The chamber 306 may have mirrored interior surfaces 310 to reflect light and increase ionization efficiency. Further in this regard, the glass 309 separating the light source from the gas and metal vapor is preferably one way mirrored glass to reflect light back into the ionization chamber. An ion beam is extracted from the chamber using an extraction electrode assembly 318 to generate an ion beam. This is facilitated by biasing the ionization chamber sidewalls 307 with a positive voltage potential (e.g., +15 kV). An extraction electrode 312 at the chamber outlet 311 is biased at a negative potential (e.g., −15 kV) and paired with a ground potential electrode 314. The extraction assembly is very close to the chamber 306 and will extract only the metal, e.g., Ca, ions. Unionized metal vapor and carrier gas, e.g., Ne, and any other non-metal ion materials, are pumped from the chamber 306 by pump line P. In a specific embodiment, the pump line P is located in the upper portion of the chamber to further discourage deposition of metal/salt vapor onto the glass plate 309 between the light source 308 and the chamber 306. The vaporizer 304, bright light source 308 and extraction assembly 318 are electrically isolated from the biased chamber sidewalls by an electrical insulator material 313. The extracted ion beam may be subjected to a magnetic mass analysis and any required acceleration/deceleration for obtaining the desired energy for subsequent processing procedures, in particular, ion implantation. The remaining lenses and scanning system are those of any conventional ion implantation system. Deposition In another aspect of the preset invention, vaporized metals may be deposited on a substrate to create a composite, such as may be useful as a barrier layer (for example a barrier to B) in semiconductor processing. FIG. 4 is flowchart of a method of vaporizing a metallic element or salt, and handling and depositing a vaporized metallic element or salt on a substrate in accordance with the present invention. The method (400) involves heating an inert carrier gas (101), for example to a temperature at or above the vaporization temperature of a metal or metal salt, for example between about 100 and 1000° C., and vaporizing a metallic element or metallic element salt in the presence of the heated inert carrier gas (103). The vaporized metallic element or salt in the heated inert carrier gas is then provided to a temperature-controlled deposition chamber (405) for application to a substrate. A chamber pressure of about 50 to 50,000 mTorr and a chamber temperature sufficient to prevent substantial deposition of metal, about 100 to 1000° C. for example, may be used. The metal is then deposited onto a cooled substrate in the chamber (407). The method allows very controlled deposition on a surface of a very small number of atomic layers of a metal. The inert carrier and dilution gas can be the more economical (but more easily ionized) argon (Ar) since there is no ionization involved. The system, with respect to the carrier gas, is closed. The carrier gas and the metal whose vapors will be deposited are maintained at the same temperature through a suitable feed-back system. The heated carrier gas minimizes deposition on unwanted surfaces, and serves to control the deposition process by determining the flow rate and concentration of the metal vapor in the gas. For example, the vapor pressure of calcium (Ca) heated to 600° C. is about 10 mTorr. If the pressure of the argon is maintained at 10 Torr, then quantity of Ca would be 1 part Ca per 1000 Ar. The flow rate would determine the time required to achieved a desired deposition thickness. Increasing the argon pressure dilutes the calcium at a given flow rate. The combination of the carrier gas pressure and flow rate can be used to perform very controlled depositions. The deposition takes place in chamber where the substrate to have metal deposited on it is cooled to below the vaporization temperature of the vaporized metal or salt by being in contact with a cooled (e.g., water-cooled) platform (substrate holder). The vapor solidifies on the cooled substrate surface to form a thin film. For example, the substrate may be cooled to a temperature of about 400° C. (the vapor pressure of the calcium, for example, at 400° C. is about one-thousandth of its vapor pressure at 600° C.). In specific embodiments, the metallic element is Ca, Sr, Ba, Mn, Cd, or Zn. Alternatively, a metal salt, such as CaCl2, CaBr2, NbCl5, or ZrCl4 may be used. The substrate may be a metal oxide dielectric such as SiO2, ZnO or HfO2, or another suitable material such as silicon. In a preferred embodiment, the metallic element is Ca and the substrate is SiO2. FIG. 5 is a block diagram of an apparatus for conducting vaporization, handling and deposition in accordance with the present invention. The apparatus may, for example, be a rapid thermal anneal (RTA) reactor adapted to the purposes of the present invention. An apparatus 500 in accordance with the present invention includes a carrier gas heating chamber 502 configured to heat an inert carrier gas to a temperature at or above the vaporization temperature of a particular metallic element or salt, generally in the range of about 100 to 1000° C. An inert carrier gas, e.g., Ar, is supplied to the heating chamber 502 via an inlet 501. The carrier gas heating chamber 502 is connected with a metal vaporizer chamber 504. The vaporizer 504 is configured to vaporize a metallic element in the presence of the inert carrier gas heated in the carrier gas heating chamber 502. The vaporizer 504 may include a resistive heater and crucible to facilitate vaporization of the metal or salt. The heated carrier gas in the heating chamber 502 is maintained at the same temperature as the vaporizer 504 by a control feedback loop 505. After vaporization, the vaporized metal or salt is held in the vapor state by the heated carrier gas for handling, for example, transport to a another chamber for use in materials processing operations. In one embodiment, depicted in FIG. 5, the vaporizer chamber 504 is connected with a temperature controlled deposition chamber 506 configured for application of the vaporized metallic element or salt to a substrate 512. The substrate 512 is held in a substrate holder 510. The deposition rate of metal on surfaces is controlled by maintaining the partial pressure of the metal vapor in the carrier gas and the temperature of these surfaces. Partial pressure is controlled by the carrier gas temperature, pressure, and flow. The deposition chamber 506, and its surfaces are heated to prevent deposition of the vaporized metallic element on them. The chamber surfaces may be heated by resistive heating elements applied to or embedded in the chamber's walls 507 and/or, at least in part, by the heated carrier gas itself. The chamber 506 is maintained at a suitable temperature and pressure for deposition, e.g., about 100 to 1000° C. and a moderate pressure of less that about 50 Torr. The apparatus 500 also includes cooling apparatus to cool the substrate holder. For example, the substrate holder may be water cooled. Metal deposition on the substrate 512 occurs when the metal vapor cools and solidifies on the cooled substrate surface. Excess is metal vapor and carrier gas is pumped from the chamber 506 by a pump 516. Excess metal is removed in a condensation chamber 514 and the inert carrier gas (e.g., argon) is recycled back to the heating chamber 502. In one embodiment, the carrier gas heating and metal vaporizer chambers of deposition and ionization apparatuses in accordance with the present invention may be shared. In this configuration, a valve 518 at the outlet of the vaporizer may be used to shunt the metal vapors to one or the other chamber (e.g., 306 or 506). The following examples provide details concerning the vaporization, handling and ionization or deposition of metals in accordance with the present invention. It should be understood the following is representative only, and that the invention is not limited by the detail set forth in these examples: Ca is ionized in a hot ambient gas mixture of Ne, which is not ionized. Ca is heated in a vaporizer to a temperature of about to 590 to 600° C. A vapor pressure of about 7 to 8 mTorr associated with Ca source material is generated. Hot Ne is the carrier gas through the vaporizor. The Ne/Ca gas mixture is flowed to a heated, highly reflective chamber at a temperature of about 590 to 600° C. with a bright light source. Surfaces in the chamber are heated to prevent deposition of material and can be heated with resistive elements around or embedded in the chamber walls. The vapor pressure of Ca in the chamber is at least 5 mTorr. The chamber has a mirror finish on its inner walls to reflect light and increase ionization efficiency. The light source is filtered to remove wavelengths <619A. Ca will be ionized and Ne will not be ionized for light whose wavelength(s) is in the range of 2028 to 600 Å. An ion beam is extracted from the chamber using an extractive electrode/suppression electrode assembly to generate an ion beam. The extraction electrode assembly extracts only the Ca ions. The un-ionized Ca and Ne are pumped from the chamber. The beam goes through a magnetic mass analysis and any required acceleration/deceleration for obtaining the desired energy. Ca is vaporized in a hot ambient gas mixture of Ar. Ca is heated in a vaporizer to a temperature of about 580 to 590° C. A vapor pressure of about 5 mTorr associated with Ca source material is generated. Hot Ne is the carrier gas through the vaporizor. The Ne/Ca gas mixture is flowed to a heated chamber at a temperature of about 590° C. Surfaces in the chamber are heated to prevent deposition of material and can be heated with resistive elements around or embedded in the chamber walls. The vapor pressure of Ca in the chamber is at least 5 mTorr. The vaporized Ca in Ar carrier is deposited on a SiO2 dielectric layer on a semiconductor wafer substrate cooled to about 400° C. A very thin layer, as thin as one or a few atoms, of Ca is formed on the SiO2 substrate. Excess Ca and Ar are pumped from the chamber for collection and reuse. Although illustrative embodiments and applications of this invention are shown and described herein, many variations and modifications are possible which remain within the concept, scope, and spirit of the invention, and these variations would become clear to those of ordinary skill in the art. Accordingly, the present embodiments are to be considered as illustrative and not restrictive, and the invention is not to be limited to the details given herein, but may be modified within the scope and equivalents of the appended claims.
045445222	claims	1. A nuclear fuel assembly including a spacer for retaining fuel elements in lateral spaced position, said spacer comprising: a structure of divider members assembled together to form respective passages for said elements, each of said divider members having laterally extending upper and lower edges; double-sided springs of generally elliptical shape comprising a continuous, seamless loop of metal supported on said upper and lower edges of at least some of said divider members, said springs being slipped over said upper and lower edges of said divider members before said members are assembled together to form said passages, one side of each of said springs extending into one of said passages and the other side of each of said springs extending into an adjacent one of said passages for engagement with and resilient lateral support of the elongated elements extending through said passages, each said double-sided spring including upper and lower minor end portions for fitting said spring on said divider members, each side of each of said springs being formed with an outwardly bowed major middle portion for extension into one of said passages and including a substantially central apex portion for contact with the element extending through said one of said passages, each side of each of said springs being further formed with upper and lower intermediate portions between said middle and end portions, said intermediate portions being formed with bends arched in the same direction as said middle portion but extending outwardly to a lesser extent than said middle portion thereby to form stops for limiting extreme movement of said element in the direction of said spring; said springs being captured between upper and lower pairs of louvers formed in said some of said divider members near the upper and lower edges thereof after said springs are positioned thereon for retaining said springs in position on said divider members; at least some of said passages being bounded on four sides by divider members with springs mounted on divider members on two adjacent sides of said passage and the divider members on the two opposite sides of said passage each being formed with a pair of vertically spaced rigid projections formed by laterally oriented arched portions of the divider member whereby an element extending through said passage is laterally supported by two spring members and two pairs of rigid projections; a peripheral support band to which the ends of said divider members are secured, said support band forming the outer sides of the peripheral ones of said passages, said support band being formed with upper and lower rigid projections extending into said peripheral ones of said passages, said support band being formed with eight outwardly extending raised portions, two near each corner on each side, for providing spacing between said support band and a surrounding tubular flow channel, said support band being formed with eight inwardly curved upward projections, two near each corner on each side, to serve as lead-ins for installing a tubular coolant flow channel on said assembly. 2. The nuclear fuel assembly of claim 1 wherein at least one of said passages is adapted to receive therethrough an oversized element of greater outside diameter than the majority of said elements and a modified double-sided spring mounted on a divider member adjacent said one of said passages, said modified spring having a substantially straight side lying along the surface of said divider member adjacent said one of said passages whereby the space in said one of said passages for said oversized element is maximized, the other side of said modified spring being outwardly bowed and extending into a passage adjacent said one of said passages. 3. The nuclear fuel assembly spacer of claim 2 wherein two diagonally adjacent passages are adapted to receive oversized elements and wherein a laterally positioned ring-shaped spring member is mounted on the intersecting divider members between said adjacent passages, portions of said ring-shaped spring member extending into said adjacent passages and engaging said oversized elements for resilient lateral support thereof.
	description	1. Field of the Invention The present invention concerns an x-ray absorption grid system of the type produced by a lithography method for use in a system for x-ray phase contrast imaging, in particular a phase contrast CT system, wherein the grid system is composed of at least two individual grids arranged atop one another in the radiation propagation direction, with each individual grid having a grid area which exhibits a plurality of grid webs and grid gaps occurring in alternation. 2. Description of the Prior Art and Related Subject Matter X-ray phase contrast imaging (in particular phase contract CT) is generally known. For example, refer to the patent application EP 1 447 046 A1. A number of x-ray absorption grids are required for this x-ray phase contrast imaging. A focus-side grid is required for generation of quasi-coherent foci, as well as a phase grid for phase modulation of the x-ray radiation by half a wavelength and an amplitude grid are arranged at the detector. The amplitude grid transduces the phase information into measurable intensity signals by step-by-step shifting. Intensity signals can be acquired with the x-ray detector situated behind the amplitude grid. The design requirements are very demanding, in particular for the detector-side amplitude grid. Structure heights of approximately 50 μm-120 μm given structure widths of approximately 1 μm should be realized over as large an area as possible. A material with maximal absorption properties is necessary for the generation of such x-ray grids; the use of gold as a grid material is proposed for this purpose, for example. In principle such grids can be produced by nanotechnology such as, for example, the LIGA method, wherein the costs for this rise significantly with the ratio of structure heights to structure widths of the grid. In German patent application DE 10 2006 037 281.6 (not published before the priority date) it is proposed to stack multiple grids atop one another instead of a single grid, and to align said multiple grids such that the grid webs and grid gaps come to lie atop one another in the direction of the x-ray radiation such that, with the use of multiple x-ray grids that can be produced with an advantageous structure height and structure width ratio (for example 40 μm web height and 1-2 μm web and gap widths), a corresponding overall grid can be produced by a number of sub-grids. For alignment, in the last cited printed document it is proposed to move the grids in the beam path relative to one another and to minimize the Moiré patterns arising due to a possibly present misalignment. For a basic pre-alignment it is proposed to apply diverse markers to the sub-grids so that the basic alignment can orient on these markers. Although the proposed method can be implemented in principle, it is relatively complicated to execute such a method given an x-ray source located in operation, in particular if not just two grids should be aligned against one another, but rather a number of grids (for example three or four) arranged atop one another are to be aligned. An object of the invention is to provide an x-ray absorption grid produced from multiple individual grids such that the alignment of the individual grids can ensue in a simpler manner in the assembly of the entire x-ray grid. The invention is based on the recognition that it is possible, by appropriate application of toothed structures on the top and bottom of individual grids during the manufacturer process, to assemble these individual grids such that the toothed structures engage with one another and an automatic alignment of the grid structures relative to one another can hereby result in the predetermined manner. No elaborate alignment of the grid structures during an exposure is necessary. Furthermore, in contrast to the previously described method with the Moiré pattern, it is now also possible to already provide an arbitrary arrangement of the grid structures relative to one another in production. For example, not only can a congruent alignment of the sub-grids arranged atop one another be achieved, but there also exists the possibility to merge these sub-grids with one another with a defined offset so that grid structures are also achieved that are functional for an angled irradiation of x-ray radiation. Based on the above concepts, an x-ray absorption grid is produced by a lithography method (advantageously an x-ray lithography method) for use in an x-ray phase contrast system (advantageously a phase contrast CT system), wherein this x-ray absorption grid has at least two individual grids arranged atop one another in the radiation direction, and each individual grid has a grid area with a number of grid webs and grid gaps occurring in alternation. The improvement according to the invention is in that each individual grid have a region outside of the grid area (outer region) and the respective outer regions of the at least two individual grids have toothed structures corresponding to one another at least at two points. The toothed structures are generated as well in the production of the grid structure. The toothed structures have a position that is defined relative to the grid structure, such that individual grids lying on one another exhibit a defined alignment of the individual grids via engagement of the toothed structures with one another. The production of the toothed structures is thus integrated into the process of the grid production, such that it is possible to position these toothed structures precisely relative to the generated grids so that combining multiple sub-grids given simultaneous positive engagement of the toothed structures leads to an optimal alignment of the grids relative to one another. This alignment can thereby be arbitrarily selected. A congruent alignment of the grid structures may be desired, or it can also be required to generate a certain transverse offset of the grids relative to one another in the event that the offset of the x-ray grids ensues in a region in which the x-ray radiation no longer strikes the x-ray absorption grid perpendicularly. The toothed structure which leads to the relative alignment of the sub-grids relative to one another can be executed, for example, as a positive-negative structure. Examples of this are a cylindrical pin for a round cylindrical hole, an n-edged pin for an n-edged hole or the like. Moreover, the side walls of the toothed structure can also proceed conically for a better engagement. The outer region can also be a border surrounding the grid area on at least two sides. A particularly secure variant is thereby generated when the outer region is a border surrounding the grid area altogether, thus on four sides. Moreover, in another embodiment the outer region of the grid area is provided with multiple external tabs distributed around the grid area. The toothed structures can then be integrated into these external tabs. Furthermore, it is particularly advantageous when the individual grids are adhered to one another such that they form a structure that securely holds together. With regard to the manufacturing costs, it can additionally be particularly advantageous when, instead of at least two individual grids, at least three or even four individual grids are arranged atop one another. For embodiments in which a transverse offset of the grid is realized in order to also be able to be used for x-ray radiation radiated at an angle, it can be more advantageous to use a higher number of individual grids since the gradations of the individual grids can then turn out to be less severe. In addition to the arrangement variants of the sub-grids with grid structures arranged congruently atop one another or, grid structures arranged uniformly offset relative to one another, the possibility also exists to arrange a number of individual grids atop one another whose grid period rises in one direction from individual grid to individual grid. For example, a quasi-curved grid can be reproduced by a planar grid, in which quasi-curved grid the grid period is also less towards the radiation center (corresponding the expansion of the radiation) than away from the radiation center. Furthermore, it is concretely proposed that the individual grids are produced by a LIGA method (lithography, electroplating and molding). This method is, for example, described in L. Hahn et al., Microsystem Technologies 11 (2005) 240-245, “MODULIGA: The LIGA-process as a modular production method-current standardization status in Germany”. This LIGA method designates a method which is based on a combination of deep lithography, electroplating and micro-molding. Microstructures with dimensions up to 0.2 μm, structure heights up to 3 mm and aspect ratios up to 50, made of plastic, metal and ceramic materials can be produced with this method. In the present case metals (such as gold, for example) are used for production of the grids in order to achieve an optimally high absorption ratio between the grid webs and grid gaps. In such a LIGA method, initially an x-ray-or UV-sensitive plastic layer (PMMA) up to 1 mm thick is applied on a base plate with an electrically conductive cover layer. A lithographic deep structuring is subsequently generated by means of high-energy radiation, for example by means of parallel synchrotron radiation or (given lesser requirements) by means of mono-energetic UV radiation. The exposed regions are subsequently dissolved away with a suitable developer and metal is incorporated into the structure interstices via a galvanic deposition method. Gold appears to be particularly suitable here as a metal since it generates a high absorption ratio between the grid webs and grid gaps. A first version of an inventive x-ray absorption grid which consists of the three sub-grids 1.1, 1.2 and 1.3 is shown in FIG. 1. The sub-grids have a substrate 5 uniformly distributed over the entire area of the grid, on which substrate 5 the grid webs and the grid gaps are arranged that all together form the grid area 4. The substrate 5 runs on the side in an outer region 6 in which a grid structure is no longer located, wherein toothed elements 2 that can engage in opposite gaps 3 of the adjacent grid are arranged in this outer region, however. If the sub-grids 1.1, 1.2 and 1.3 are assembled, the x-ray absorption grid 1 shown in FIG. 2 in section A-A results with the individual grids 1.1, 1.2 and 1.3 that are aligned relative to one another in the desired manner due to the mutual engagement of the toothed structures, such that a further adjustment can be omitted. The x-ray absorption grid 1 is again shown in FIG. 3 in plan view. Here the centrally arranged grid area 4 with the grid webs and grid gaps is recognizable. The grid area 4 is circumferentially enclosed on four sides by an outer region 6 in which the toothed structure 2/3 is likewise circumferentially arranged. Three sub-grids 1.1, 1.2 and 1.3 that are different with regard to their toothed structure are used in the x-ray absorption grid shown here in FIGS. 1 through 3. As can be seen from FIGS. 1 and 2, the upper individual grid 1.1 has a toothed structure 2 only on the underside while the lower individual grid 1.3 has a toothed gap 3 only on its top side. The middle grid 1.2 possesses a toothed gap 3 on the top side and a toothed structure 2 on the underside of the outer region, such that after an assembly of these three sub-grids (as shown in FIG. 2) an entire x-ray absorption grid appears which no longer possesses toothed structures facing outwards. FIGS. 4, 5 and 6 show a further version of an x-ray absorption grid that is formed only of two individual grids 1.1 and 1.2. In this embodiment of the x-ray absorption grid the outer region 6 runs only on two opposing sides. Moreover, no continuous toothed structure is shown in the outer region; rather, individual gaps 3 and projections 2 engaging in them are apparent that (as is clear from FIG. 6) are arranged offset relative to one another, for example. The toothed structures additionally possess a slightly conical shape, such that the assembly of the individual grids 1.1 and 1.2 is made easier. The same effect as given the circumferential toothed structure is achieved in the two outer regions 6 due to the repeatedly divided toothed structures, namely that the grids can gradient echo securely assembled both in the grid line longitudinal direction and in the transverse direction and a slippage is prevented. A further embodiment of an inventive x-ray absorption grid is shown in FIGS. 7, 8 and 9. This has four individual grids 1.1 through 1.4. In this case all grids 1.1 through 1.4 are designed identically (at least with regard to their outer region, which here is formed from individual external tabs), and both the positive and the negative toothed structures 2 and 3 that engage in one another and therefore ensure the relative positioning of the individual grids 1.1 through 1.4 relative to one another are located in each external tab 7. FIG. 10 shows a portion of a further embodiment of the x-ray absorption grid in a side view, which has the same individual grid 1.1 as shown in the previous figures, but modified versions of the other grids 1.5 and 1.6. In the embodiment of FIG. 10, the toothed structures are offset from each other, so that the grid structures in the respective individual grids 1.1 and 1.6 are also offset relative to each other. As also schematically indicated in FIG. 10, the grid structure in the individual grid 1.5 has a different periodicity from the grid structures of the individual grids 1.1 and 1.6, as schematically illustrated by a larger spacing between the lines indicating the grid walls. As also shown in FIG. 10, the respective grids can be held together by an adhesive (shown exaggerated in FIG. 10) between adjacent individual grids. Furthermore, by means of the toothed structures on the outsides after the assembly of the four individual grids, the possibility exists to likewise use these toothed structures as a positioning aid upon assembly with the detector, for example, in that corresponding counter-structures exist in the region of the detector structure. Although modifications and changes may be suggested by those skilled in the art, it is the intention of the inventor to embody within the patent warranted hereon all changes and modifications as reasonably and properly come within the scope of his contribution to the art.
	description	This application is a continuation of U.S. application Ser. No. 12/410,558, filed Mar. 25, 2009, which is a continuation of U.S. application Ser. No. 11/181,809, filed Jul. 15, 2005, which claims priority under 35 U.S.C. §119 to Japanese Patent Application No. 2004-214541, filed Jul. 22, 2004, the entire disclosures of which are herein expressly incorporated by reference. This application is also copending with U.S. application Ser. No. 11/797,344, filed May 2, 2007, and U.S. application Ser. No. 12/165,781, filed Jul. 1, 2008, the entire disclosures of which are herein expressly incorporated by reference. 1. Field of the Invention The present invention relates to a method of suppressing deposition of radionuclides on components of a power plant such as a nuclear power plant, and a ferrite film formation apparatus for performing the radionuclide deposition suppression method. 2. Background Arts For example, in a nuclear power plant of boiling water type reactor (hereunder abbreviated as BWR), the heat generated by fuel is effectively transferred to cooling water by forcibly circulating the cooling water by a primary loop recirculation pump and internal pumps inside the nuclear reactor comprising a reactor pressure vessel and control rods contained therein. Most of the steam generated from the cooling water in this way is used for driving a steam turbine generator, the steam exhausted from the steam turbine is condensed in a condenser, and the condensate condensed in the condenser is almost fully deaerated and supplied again as cooling water of the nuclear reactor. In this case, in the condenser, oxygen and hydrogen generated by decomposition of water by radiation are also almost completely removed. Further, the condensate to be returned to the nuclear reactor is supplied to the nuclear reactor after removing mainly metal impurities by an ion-exchange resin filter and being heated to nearly 200° C., in order to suppress occurrence of activated corrosive products in the nuclear reactor. Further, since the activated corrosive products are generated inside the rector pressure vessel and at water-contact portions of a recirculation system, stainless steel such as nickel-base metal which is less in corrosion is used for components of the main primary system. Further, for a reactor pressure vessel made of low alloy steel, inside building up of stainless steel is performed, thereby to prevent the low alloy steel from contacting directly with the reactor water. In addition to taking into a consideration the above-mentioned matters, a part of the reactor water is purified by a reactor-water cleanup facility and metal impurities produced slightly in the reactor water are positively removed. However, even if the above-mentioned measures for corrosion prevention are taken, very slight metal impurities exist inevitably in the reactor water, so that a part of the metal impurities deposits on the surface of a fuel rod as metal oxide. Metal elements deposited on the surface of the fuel rod are irradiated by neutrons radiated from the fuel to cause nuclear reaction, whereby radionuclides such as cobalt 60, cobalt 58, chromium 51, manganese 54, etc are produced. Most of those radionuclides are, as they are, deposited on the fuel rod surface in a state of oxide, however, a part of the radionuclides is dissolved out into the cooling water according to the solubility of oxides taken in and released again into the reactor water as insoluble solid called cruds. Radioactive substances in the reactor water are removed by a reactor-water purifying system, however, radioactive substances that can not be removed are accumulated on surfaces of water-contact portions of the components while the substances are recirculating in a recirculation system and so on together with the reactor water. As a result, radiations are radiated from the surface of the components, which becomes a cause of radiation exposure on person engaged with regular inspection during the inspection. A dose of radiation exposure under working is managed not to exceed a specified value for each person, however, recently the specified value is raised and there occurs the necessity that the exposure dose for each person is reduced as low as economically possible. Therefore, various methods such as a method of reducing deposition of radionuclides on piping, a method of reducing the concentration of radionuclides in the reactor water, etc. are studied. For example, there is proposed a method of injecting metal ions such as zinc into the reactor water, forming a dense oxide film including zinc on the water-contact surface of piping of a recirculation system and suppressing to take radionuclides such as cobalt 60, cobalt 58, etc. into the oxide film (Patent literature 1). Further, there is proposed a method of forming a pre-oxidation film under a certain condition on inner surfaces of recirculation system piping and a reactor-water purifying system in which the reactor water flows during operation, before reaching to the condition that the radionuclides are dissolved out or released (Patent literature 2). Patent literature 1: JP 58-79196 A Patent literature 2: JP 62-95498 A The method of injecting metal ions such as zinc into the reactor water as described in Patent literature 1 has a problem that depleted zinc must be used to avoid zinc itself being radioactivated, in addition to the problem that injection of zinc ions must be always continued during operation. Further, since the method of forming an oxide film described in Patent literature 2 forms the oxide film in an operation-temperature range (250˜300° C.) of BWR, for instance, it is found that there are following problems: Namely, according to the study of the inventors, it is found that in the case where material of components to be formed of oxide film is stainless steel, inner oxide film having much chromium component is formed first on the surface of component, and an outer oxide film having less chromium component is formed on the inner oxide film surface. Particularly, in the case of such a double layer structure oxide film, it has such a property that radioactive Co-60 and Co-58 are easily taken in the inner oxide film, and it is found that a suppression effect of radioactive nuclide deposition is not so large. It is an object of the present invention to effectively suppress deposition of radionuclides on components of a nuclear power plant. In order to solve the above subject or achieve the above object, various studies are conducted, as a result, the following is found. First, when an oxide film is formed in a high temperature atmosphere such as a reactor water temperature range (for example 250˜300° C.) during service operation, the diffusion velocity of oxygen generated by radiolysis of water into metal parent material of a component that the oxide film is to be formed on, and oxidation velocity of parent material is fast, so that the oxide film (inner layer) formed initially becomes much in chromium component contained in stainless steel. Secondly, a main component of the oxide film (outer layer) formed on the inner oxide film is iron because iron ions dissolved out from the parent material precipitate in high temperature water beyond the solubility. Thirdly, when an oxide film is formed in a high-temperature atmosphere such as in a temperature range during service operation, a film structure such as crystal of the outer oxide film formed on the inner oxide film is not always densely formed. Therefore, radionuclides such as cobalt in the reactor water transmit the outer oxide film, and the radionuclides are taken in the inner oxide film in the form that they are replaced by ion components in the inner oxide film. The experiments conducted by the inventors to get such knowledge are explained, referring to FIGS. 2 to 4. FIG. 2 shows an experimental result about stainless steel in which oxide film is not formed in advance, the experimental result being a deposit amount of Co-58 deposited on a sample, obtained with time passage, by soaking the sample in cooling water under the condition of service operation of BWR. In FIG. 2, the abscissa shows soaking time and the ordinate shows relative values of Co-58 deposit. As shown in FIG. 2, such a tendency is shown that Co-58 deposit increases with passage of soaking time and finally the deposit saturates. This change in Co-58 deposition has a tendency similar to the tendency of oxide film growth. That is, a corrosion rate is fast at an initial time of soaking and a deposition rate of Co-58 also becomes fast because the film growth rate is fast. However, after passage of about 500 hours, the Co-58 deposition rate becomes slow as the film growth rate becomes slow by diffusion resistance of an oxidizing agent in high temperature water due to the grown oxide film. FIG. 3 shows an experimental result that a sample A that a pre-oxidation film is not formed on stainless steel and a sample B that a pre-oxidation film is formed on stainless steel are soaked in the cooling water under the condition of BWR service operation and relative values of deposit of Co-58 deposited on the samples A and B are compared. In those samples A and B, A1 and B1 each are an inner oxide film that chromium is a main component, and A2 and B2 are each are an outer oxide film that iron is a main component. From this experimental result, it is found that radioactive cobalt is taken in the inner oxide film B of chromium of a main component in the case of pre-oxidation film formed in the high temperature water under the condition of BWR service operation as in the sample B. That is, it is found that a deposition suppression effect of radionuclides is small even if it is tried to suppress taking in of cobalt by forming the pre-oxidation film in the high temperature water under the condition of BWR service operation and making the growth rate of the oxide film formed during the service operation slow. On the contrary, in the case of the sample A that the pre-oxidation film is not formed, it appears that taking in of cobalt is easily occurs in the course of growing of the outer oxide film A2 that iron is a main component. However, by providing the pre-oxidation film as in the sample B, it is found that it becomes difficult to take cobalt into the outer oxide film B2 becoming slow in growth. It is found that the outer oxide films A2 and B2 that iron is a main component are ferrite films (hereunder, referred to as magnetite films) that magnetite is a main component from analysis of laser Raman spectral. From the above-mentioned matters, if it is possible to produce a magnetite film as an oxide film, it is expected to be able to suppress deposition of cobalt. However, when forming an oxide film in a water of high temperature close to a temperature in a BWR service operation, since it is impossible to suppress having formed an inner oxide film B1 that chromium is a main component through diffusion of dissolved oxygen in the cooling water into the metal parent material, such a problem is left that the inner oxide film B1 becomes a source of taking in radioactive cobalt and it is impossible to suppress deposition of radionuclides. If it is possible to form only a magnetite film such as the outer oxide film A2, B2 under the temperature condition (for example, 100° C. or lower) that the diffusion rate of dissolved oxygen into metal parent material is slow, it is considered to be able to suppress taking in of cobalt of radionuclides. After forming a magnetite film on a stainless steel surface on the basis of this consideration, it is soaked in the high-temperature water under the condition of BWR service operation and a deposit amount of Co-60 is examined. As a result, it is found possible to suppress greatly the deposit amount of Co as shown in FIG. 4. In FIG. 4, the ordinate shows relative values of deposit of Co-58 of samples C, D, E. The sample C is a sample that a surface of stainless steel is mechanically grinded, the sample D is a sample that a pre-oxidation film is formed on the surface of stainless steel under the condition of BWR service operation, and the sample E is a sample that a magnetite film is formed on the surface of stainless steel at a temperature of 100° C. or less. As it is apparent from FIG. 4, the deposit amount of Co in the sample E in which the magnetite film is formed is greatly suppressed as compared with the samples C and D. Further, as a method of forming the magnetite film, it is not technique of suppressing corrosion of components of a nuclear power plant, but it is technique of forming a ferrite film of magnetic recording medium. However, a method described in JP 63-15990 B can be applied for the method, for instance. Although the method of JP 63-15990 B uses chlorine, however, it is impossible for the components of nuclear power plant to use chlorine from a viewpoint of securing of soundness of the components, so that it is necessary to take a method different from the method of JP 63-15990 B. The present invention is made to solve the above-mentioned subject and characterized by adsorbing iron (II) ions (ferrous ions) on surfaces of metal members composing a nuclear power plant, oxidizing the adsorbed iron (II) ions to form ferrite films under the temperature condition from a normal temperature to 200° C., preferably from a normal temperature to 100° C., more preferably from 60° C. to 100° C., and suppressing deposition of radionuclides on the metal members by the ferrite films, on the basis of the acknowledge. According to the above, the ferrite films formed on the surfaces of members composing a nuclear power plant are ferrite films that magnetite is a main component, corresponding to the sample E in FIG. 4, because the temperature at time of forming the films is lower than the temperature in the core service operation. That is, in the case where the temperature at time of ferrite film formation is in a temperature range at time of service operation of the core, since a diffusion velocity of an oxidizing agent (oxidizer) is fast, a conventional inner oxide film that chromium is a main component is formed, and cobalt is taken in there. On the contrary, in the case where the temperature at time of formation of the ferrite film is 60° C. to 100° C. which is low, since the diffusion velocity of the oxidizer is slow, the conventional inner oxide film that chromium is a main component is not formed, but a magnetite film that taking in of radionuclides is less and iron is a main component, corresponding to the sample E in FIG. 4 is formed, whereby it is possible to suppress deposition of radionuclides onto components. Further, after acting a solution including bivalent iron ion (iron (II) ion) obtained by solving iron in organic acid or carbonic acid onto the surfaces of components, an oxidizer such as oxygen is injected into the solution, whereby it is possible to form the oxide film. When the radionuclide deposition suppression method according to the present invention is applied, it is preferable for a water-contact surface of a component to be treated to be bare (to have nothing thereon). Particularly, on the surfaces of components having operated for service, oxide films for corrosion prevention are formed by oxygen of an oxidizer produced by radiolysis of water in the cooling water. The oxide film has been formed in a high-temperature range under service operation, so that the oxide film that chromium is a main component has been formed as in the inner oxide film B1 of the sample B in FIG. 3, and radionuclides have been already taken in. Therefore, even if a magnetite film according to the present invention is formed on the oxide film, it is impossible to obtain a sufficient exposure-dose reducing effect thereby. Therefore, the radionuclide deposition suppression method according to the present invention is preferable to be practiced after termination of a decontamination process such as chemical decontamination and so on conducted commonly in the nuclear power plant. Concretely, it is preferable to practice the method after a termination stage of the decontamination process and before starting the core. That is, the decontamination such as chemical decontamination or the like is a treatment for removing radionuclides deposited on the surfaces of nuclear power plant components in contact with reactor water together with oxide films. Therefore, the radionuclide deposition suppression method according to the present invention is practiced under the condition that surfaces of parent material of components to be treated are exposed by the decontamination, and any natural oxide film does almost not exist. As a result, a magnetite film which is a kind of ferrite film according to the present invention is formed directly on the surface of component, so that it is possible to effectively suppress deposition of radionuclides. Thereby, it is possible to reduce exposure dose at time of regular inspection working. As components or members that the radionuclide deposition suppression method according to the present invention is applied, components or members composing a reactor-water recirculation system or reactor-water purifying system of a BWR power plant are preferable, but the component or members are not limited thereto. Further, the present invention is not limited to the BWR plant, but it can be applied for suppression technique of radionuclides deposition on components in contact with reactor water in a pressurized water type (PWR) nuclear power plant. A ferrite film forming apparatus for practicing the radionuclide deposition suppression method according to the present invention comprises a surge tank for storing a processing solution, a recirculation pump for sucking the processing solution in the surge tank, processing solution supply piping for supplying the processing solution sucked by the recirculation pump to piping for film formation, a first chemical tank for storing iron (II) ions to be injected into the processing solution in the processing solution supply piping, a second chemical tank for storing an oxidizer to be injected into the processing solution in the processing solution supply pipe, a third chemical tank for storing a pH adjuster to adjusting the pH of the processing solution in the processing solution supply piping to be 5.5 to 9.0, a processing solution return piping for returning the processing solution to be returned from the piping for film formation to the surge tank, and a heater for heating the processing solution to a temperature of 60-100° C. Further, when the apparatus is used both as film formation and as chemical decontamination, it is possible to construct the apparatus so as to fluidly communicate a chemical tank for oxidizer and reducer for chemical decontamination of the piping that a film is to be formed on the above-mentioned processing solution supply piping. Effect of the Invention According to the present invention, it is possible to effectively suppress deposition of radionuclides on components of a nuclear power plant. Hereunder, embodiments concerning a method of suppressing deposition of radionuclides on components of a nuclear power plant according to the present invention will be explained, referring to the drawings. First Embodiment FIG. 1 shows a flowchart of an embodiment of a radionuclide deposition suppression method of the present invention, FIG. 5 shows a schematic diagram of a whole system construction of an embodiment in which the present invention is applied to recirculation piping of a nuclear power plant, and FIG. 6 shows a schematic diagram of a detailed system construction of a film forming apparatus for practicing the radionuclide deposition suppression method according to the present invention. As shown in FIG. 5, the nuclear power plant comprises a nuclear reactor 1 having fuel rods contained in a pressure vessel, main steam piping 2 connected to the top of the nuclear reactor 1, a steam turbine generator 3 connected to the main steam piping 2, and a condenser 4 connected to a steam outlet of the steam turbine generator 3. The condensate condensed in the condenser 4 is taken out by a condensate pump 5, and returned as feed water for the nuclear reactor 1 through a feed water piping system which comprises a condensate purifying apparatus 6, a feed water pump 7, a low pressure feed water heater 8, and a high pressure feed water heater 9. Heat sources of the low pressure feed water heater 8 and the high pressure feed water heater 9 are extraction steam from the steam turbine generator 3. Further, a plurality of reactor-water recirculation systems are provided for recirculating cooling water in the nuclear reactor 1, and constructed so that the reactor water extracted by a plurality of recirculation pumps 21 connected to a bottom portion of the nuclear reactor 1 is returned to an upper portion of the nuclear reactor 1 through reactor-water recirculation piping connected to the respective recirculation pumps, thereby to be recirculated. Further, a reactor-water purifying system for purifying the reactor water in the nuclear reactor 1 is constructed so that the reactor water extracted by a purifying system pump 24 connected to the bottom portion of the nuclear reactor 1 is cooled through a regenerative heat exchanger 25 and a non-regenerative heat exchanger 26, and the cooled reactor water is purified by a reactor-water purifying apparatus 27 and returned into the nuclear reactor 1 from a downstream side of the high pressure feed water heater of the feed water piping system after rising the temperature of the purified reactor water by the regenerative heat exchanger 25. Further, FIG. 5 shows the case where the present invention is applied to the reactor-water recirculation system, and the state that a ferrite film formation apparatus 30 for practicing the radionuclide deposition suppression method according to the present invention is connected to the reactor-water recirculation system by temporary piping. That is, when a service operation of the nuclear reactor 1 is stopped, connection between the nuclear reactor 1 and the reactor-water recirculation piping 22 is cut off, and the ferrite film formation apparatus 30 is connected to the cut-off reactor-water recirculation system though the temporary piping so that a processing solution can be recirculated. An embodiment of the ferrite film formation apparatus 30 is shown in FIG. 6. The film formation apparatus 30 of the present embodiment is constructed so that it can be jointly used with chemical decontamination processing. Namely, the apparatus 30 comprises a surge tank 31 filled with water used for processing, and a recirculation pump 32 for extracting the water in the surge tank 31 and feeding it to an end of the reactor-water recirculation piping 22 through valves 33, 34. Chemical tanks 42, 43 and 44 each used for chemical decontamination are connected to a processing solution piping 35 having the valves 33 and 34 connected thereto through valves 36, 37 and 38 and injection pumps 39, 40 and 41. In the chemical tank 42, permanganic acid ions for oxidative dissolution of contaminated substances inside the piping are stored. In the chemical tank 43, oxalic acid for reductive dissolution of contaminated substances in the piping is stored. In the chemical tank 44, hydrazine for pH adjustment is stored. Further chemical tanks 49 and 50 each used for ferrite film formation are connected to the processing solution piping 35 through valves 45 and 46 and injection pumps 47 and 48. In the chemical tank 49, chemicals including iron (II) ions prepared by dissolving iron with formic acid are stored. Further, chemicals for dissolving iron are not limited to the formic acid, but organic acid or carbonic acid, which becomes counter anion to the iron (II) ion, can be used therefor. In the chemical tank 50, water in which oxygen as oxidizing agent is dissolved is stored. Further, hydrazine stored in the chemical tank 44 is used for pH adjustment at time of ferrite film formation. Meanwhile, the processing solution fed to the one end of the reactor-water recirculation piping 22 by the recirculation pump 32 is passed through an inside of the reactor-water recirculation piping 22 and returned to the valve 51 from the other end. The processing solution returned through the valve 51 is returned to the surge tank 31 through a recirculation pump 52, a heater 53 and valves 55, 56 and 57. A cooler 58 and a valve 59 are connected to the heater 53 and the valve 55 in parallel. The valve 56, a cation exchange resin tower 60 connected to a valve 61 in series and a mixed bed ion exchange resin 62 connected to a valve 63 in series are connected to the processing solution piping 35 in parallel, respectively. Decomposition equipment 64 is connected in parallel with a valve 57 via a valve 65. A chemical tank 68 is connected to the decomposition equipment 64 via a valve and an injection pump 67, and constructed so that the hydrogen peroxide solution stored in the chemical tank 68 can be injected into the decomposition equipment 64. It is preferable that the position of the valve 46 for oxidizer injection is at a downstream side of the valve 45 for iron (II) ion injection and the valve 38 for pH adjuster injection, and possibly close to an object to be treated. Further, it is preferable to set a filter at an exit side of the surge tank 31. Further, it is preferable for removing oxygen in solution to bubble inert gas such as nitrogen, argon gas, etc. in the chemical tank storing chemicals including iron (II) ions, the chemical tank 44 storing pH adjusting chemicals, and the surge tank 31. Further, the decomposition equipment 64 is constructed so that organic acid used as counter anion of iron (II) ion and hydrazine of pH adjuster can be decomposed. That is, as the counter anion of iron (II) ion, organic acid which can be decomposed into water and carbon dioxide taking into consideration of reduction in waste amount, or carbonic acid which can be released as gas and does not increase an amount of wastes is used. Further, it is preferable for suppressing a use amount of chemicals to recover unreacted chemicals by separating and removing excessive reaction products and reuse them. Using the film formation apparatus 30 constructed as mentioned above, a processing procedure for practicing the radionuclide deposition suppression method according to the present invention in the reactor-water recirculation system will be explained along the flowchart of FIG. 1. First of all, in practice of the method of the present invention, it starts from connection of the film formation apparatus 30 to the piping system including a component that a film is to be formed (S1). For example, as in FIG. 5, in the case where the reactor-water recirculation system is to be treated, connection of the nuclear reactor 1 and the reactor-water recirculation piping 22 is cut off, and the film formation apparatus 30 is connected to the cut-off reactor-water recirculation system through temporary piping. Next, in the present embodiment, contaminated substances such as oxide films taken in radionuclides formed on the metal member surfaces in contact with the reactor water are decomposed by chemical processing, using the film formation apparatus 30 (S2). When the radionuclide deposition suppression method of the present invention is practiced, it is preferable to conduct the chemical processing, however, it is not limited to the chemical processing. It is necessary that the metal member surface or surfaces are exposed before practicing the radionuclide deposition suppression method, and it is possible to employ mechanical decontamination such as grinding. The chemical decontamination in step S2 is a conventional method, however, it will be explained briefly. First, the valves 33, 34, 51, 55, 56 and 57 are opened, under the condition that the other valves are closed, the recirculation pumps 32 and 51 are started, and the processing solution in the surge tank 31 is recirculated into the reactor-water recirculation system 22 that is to be chemically decontaminated. Further, the temperature of the processing solution is raised to a temperature of about 90° C. by the heater 53. Next, the valve 36 is opened and the injection pump 39 is started, permanganic acid ions in the chemical tank 42 is injected into the processing solution. Thereby, radioactive contaminant such as oxide films formed on the object that is to be decontaminated is oxidatively dissolved. After the oxidative dissolution of radioactive contaminant is finished in this manner, in order to decompose permanganic acid remained in the processing solution, the valve 37 is opened and the injection pump 40 is started, and oxalic acid in the chemical tank 43 is injected into the processing solution. Successively, in order to conduct reductive dissolution of the contaminant, the oxalic acid in the chemical tank 43 is further injected into the processing solution, and in order to adjust the pH of the processing solution, the valve 38 is opened and the injection pump 41 is started, and hydrazine is injected from the chemical tank 44 into the processing solution. After oxalic acid and hydrazine are injected in this manner, the valve 61 is opened and the valve 56 is closed, the processing solution is passed to the cation exchange resin tower 60, the metal cation dissolved into the processing solution is adsorbed on the cation exchange resin and removed from the processing solution. After, the reductive dissolution is finished, in order to decompose the oxalic acid in the processing solution, the opening of the valve 57 for bypassing the valve 65 at the exit side of the decomposition equipment 64 and the decomposition equipment 64 is adjusted, whereby a part of the processing solution is flowed into the decomposition equipment 64. At this time, the valve 66 is opened, the injection pump 67 is started, and hydrogen peroxide in the chemical tank 68 is injected into the processing solution flowing into the decomposition equipment 64, whereby the oxalic acid is decomposed in the decomposition equipment 64. After the oxalic acid is decomposed, in order to remove the impurities in the processing solution, the heater is turned off and the valve 55 is closed. At the same time, the valve 59 of the cooler 58 is opened and the processing solution is passed into the cooler 58 to be lowered in temperature. After the temperature of the processing solution is lowered to a temperature (for example 60° C.) at which it can be passed though the mixed bed resin tower 62 in this manner, the valve 61 of the cation exchange resin tower 60 is closed and the valve 63 at the side of the mixed bed resin exchange tower 62 is opened, and the processing solution is flowed into the mixed bed ion exchange tower 62, thereby to remove impurities in the processing solution. Those series of operations from temperature elevation, oxidizer dissolution, oxidizer decontamination, reductive dissolution, reductive decomposition and purifying operation are repeated twice or thrice, whereby contaminants including oxide films of the metal member that is to be decontaminated are dissolved and removed. In this manner, after the contaminants including oxide films of the metal members are removed, the processing is changed to a ferrite film forming processing according to the present invention. First, after finishing of the final purification operation, the processing solution is adjusted to be a prescribed temperature by the heater 53 (S3). The prescribed temperature at this time is preferable to be around 100° C., however, it is not limited to the temperature. The point is that the formed ferrite film is sufficiently dense in film structure such as crystal to the extent that radionuclides in the reactor water during service operation of the nuclear reactor cannot be taken in. Therefore, at least, 200° C. or lower is preferable and although the lower limit is sufficient to be a normal temperature, 60° C. or higher at which a forming rate of the film becomes in a practical range is preferable. Further, from a view point of suppression of a use amount of chemicals also, the temperature of the processing solution for forming a ferrite film is preferable to be 60-100° C. The reason is explained. FIG. 7 shows relationship between voltage and pH in an iron-water system, calculated by saturated solubility 0.065 g/100 g in atmospheric saturated water at 25° C. of iron(II) carbonate-iron concentration of water 5.6×10-3 mol/L. From FIG. 7, it is found that a stabilized region of magnetite (Fe3O4) which is a kind of ferrite is expanding to a low pH side as the temperature rises from 25° C. to 60° C., and from 60° C. to 100° C. Therefore, it is possible to reduce an amount of chemical for adjusting the pH of the processing solution by raising the temperature. In order to form a ferrite film, it is necessary that iron (II) ion is adsorbed on the surface of an object that a film is to be formed. However, iron (II) ion in the solution is oxidized to iron (III) ion by dissolved oxygen according to the following chemical formula 1, the iron (III) ion is precipitated as iron hydroxide according to the following chemical formula 2 because the iron (III) ion is lower in solubility than iron (II) ion, so that the iron (III) ion comes not to contribute to ferrite film formation. Therefore, it is preferable to conduct bubbling of inert gas or vacuum deaeration in order to remove oxygen in the processing solution.4Fe2++O2+2H2O→4Fe3++4OH— (1)Fe3++3OH—→Fe(OH)3 (2) When the temperature of the processing solution recirculated in this manner reaches a prescribed temperature, the valve 45 is opened and the injection pump 47 is started, chemicals containing iron (II) ion prepared by dissolving and adjusting iron by formic acid are injected into the processing solution from the chemical tank 49 (S4). Thereby, the iron (II) ion is adsorbed on the metal member surface that is to be treated. Successively, in order to oxidize the iron (II) ion adsorbed on the metal member surface to be iron (III) ion and to make it into ferrite, the valve 46 is opened and the injection pump 48 is started, whereby water having oxygen of oxidizer dissolved therein and stored in the chemical tank 50 is injected into the processing solution (S5). Successively, in order to adjust the pH of the processing solution to 5.5-9.0, the valve 38 is opened and the injection pump 41 is started, and hydrazine is injected from the chemical tank 44 into the processing solution (S6). Thereby, an oxide film of ferrite film (hereunder, referred to as magnetite film) that magnetite is a main component is formed on the position of an object to be treated. In step S7, when formation of the magnetite film is finished in this manner, the processing goes to a waste solution processing step S8. However, when the magnetite film formation is not finished, the processing goes to step S9 where chemicals are filled, and then a magnetite film of necessary thickness is formed. Since formic acid and hydrazine remain in the processing solution after forming of the magnetite film, it is necessary to have removed these impurities by practicing the waste liquid treatment in step S8 when the waste solution is discharged. Meanwhile, disposal of those impurities in the ion exchange resin tower 60 increases waste of the ion exchange resin. Therefore, in the waste liquid treatment in step S8, it is preferable to decompose and dispose the formic acid to carbon dioxide and water, and the hydrazine into nitrogen and water, using the decomposition equipment 64 in the decontamination system. Thereby, it is possible to reduce a load of the ion exchange resin tower 60 and to reduce a waste amount of the ion exchange resin. Further, since the decomposing disposal causes a part of the processing solution to flow in the decomposition equipment as in the decomposition of oxalic acid, formic acid and hydrazine are decomposed by adjusting the openings of the valve 57 bypassing the decomposition equipment 64 and the valve 65 of the decomposition equipment 64, and injecting hydrogen peroxide into the processing solution flowing in the decomposition equipment 64. In this manner, by forming the magnetite film on the object to be treated while suppressing the waste of ion exchange resin and a generation amount of radioactive waste, it is possible to suppress deposition of radionuclides or radioactive cobalt ions on the object to be treated during normal service operation of the nuclear reactor. As a result, it is possible to reduce exposure at time of regular inspection by suppressing a dose rate of the piping of the reactor-water recirculation system. Further, chemicals such as chlorine are not used for the film formation processing, so that the soundness of components of the nuclear power plant is not spoiled. Second Embodiment FIG. 8 shows a system of a concrete embodiment of a portion forming iron (II) ions stored in the chemical tank 49 shown in FIG. 6. In FIG. 8, formic acid solution is stored in a chemical tank 70, and nitrogen from nitrogen bubbling equipment 71 is bubbled in the chemical tank 70, whereby dissolved oxygen is removed. The formic acid from which the dissolved oxygen is removed is transferred to a metal iron dissolution tank 73 by an injection pump 72. Metal iron 74 is set in the metal iron dissolution tank 73, and iron (II) ion is dissolved by the formic acid transferred thereto. At this time, nitrogen from the nitrogen bubbling equipment 71 is bubbled in the metal iron dissolution tank 73. The formic acid solution in the metal iron dissolution tank 73 does not substantially include dissolved oxygen, so that the iron (II) ion is almost not oxidized to iron (III) ion. The formic acid solution which dissolved the iron (II) ion is transferred to the chemical tank 49 by an injection pump 75 and stored therein. Further, nitrogen from the nitrogen bubbling equipment 71 is bubbled in the chemical tank 49. Nitrogen from the nitrogen bubbling equipment 71 is bubbled also during storage, and prevents the iron (II) ion from being oxidized to Iron (III) ion by oxygen in the air. The iron (II) ion solution formed and stored in this manner is used for the film formation processing of the first embodiment shown in FIG. 6. Therefore, the iron (II) ion solution prepared according to the embodiment of FIG. 8 is not in contact with oxidizer until the solution is mixed with the oxidizer such as oxygen to be injected from the chemical tank 50 in the film formation apparatus 30, so that an amount of the iron (II) ion as it is becomes much. Therefore, the amount of iron (II) ion adsorbed on the object portion to be treated increases and the formation reaction of magnetite film becomes easy to occur. Third Embodiment FIG. 9 shows a system of another embodiment of iron (II) ion formation shown in FIG. 8. In the present embodiment, carbonic water instead of formic acid in the embodiment of FIG. 8 is used as an iron ion solution. Carbonic acid is stored in a chemical tank 76, carbon dioxide supplied form a carbon dioxide bubbling equipment 77 is bubbled into the chemical tank 76, and saturated carbonic water of one atmosphere is formed. The formed one atmosphere saturated carbonic water is transferred to the metal iron dissolution tank 73 by the injection pump 72. Iron (II) carbonate 78 is contained in the metal iron dissolution tank 73, and the one atmosphere saturated carbonic water is dissolved. Further, carbon dioxide from the carbon dioxide bubbling equipment 71 is bubbled into the metal ion dissolution tank 73. By the way, according to the literature (Chemical Unabridged Dictionary 5, page 729 (1997), Reduced-size edition), usually, the solubility of iron (II) carbonate is 0.065 g/100 g-water at 20° C., however, in the case of one atmosphere saturated carbonic water, the solubility rises to 0.1 g/100 g-water. Further, it can exhaust dissolved oxygen and does not substantially contain dissolved oxygen, so that the dissolve iron (II) ion is almost not oxidized to iron (III) ion. The saturated carbonic water solution in which iron (II) ion is dissolved is transferred to and stored in the chemical tank 49 by the injection pump 75 via a filter 79 removing undissolved iron (II) ion. During storage, also, carbon dioxide from carbon dioxide bubbling equipment 77 is bubbled into the chemical tank 49, whereby the iron (II) ion is prevented from being oxidized to be iron (III) ion by oxygen in the air. The iron (II) ion solution produced and stored in this manner is used for the film formation processing shown in FIG. 6. Since the iron (II) ion solution prepared in the embodiment shown in FIG. 9 does not contact with oxidizer until it is mixed with the oxidizer such as oxygen injected from the chemical tank 50 in the film formation apparatus 30, an amount of iron (II) ion as it is and reached to an object portion to be treated becomes much. Therefore, an amount of iron (II) ion adsorbed on the object portion to be treated increases, and the reaction of magnetite film production becomes easy to occur. Further, since carbonic acid ion is used as counter anion of iron (II) ion, it is possible to reduce an amount of solid waste. Fourth Embodiment FIG. 10 shows a system of another embodiment of the film formation apparatus 30 according to the present invention. A difference of the present embodiment from the embodiment shown in FIG. 6 is that the nitrogen bubbling equipment 71 is connected to each of the surge tank 31, the chemical tank 44 for hydrazine and the chemical tank for iron (II) ion. Thereby, it is possible to make each solution contained in each tank into the solution substantially not containing oxygen by exhausting dissolved oxygen in each tank. As a result, it is possible to reduce formation of iron (III) ion contained in the solution and not distributed to magnetite film formation and to suppress decrease in magnetite film formation reaction, as well as the above-mentioned second and third embodiments. Fifth Embodiment FIG. 11 shows a system of another embodiment of the film formation apparatus 30 according to the present invention. In the present invention, a filter 80 for removing side reaction products is provided between the recirculation pump 32 at the exit side of the surge tank 31 and the valve 33, in addition to the embodiment shown in FIG. 10. In the magnetite film formation, iron (II) ion is adsorbed on an object surface to be treated, a part thereof is oxidized to iron (III) ion by oxidizer injected from the chemical tank 50 to form ferrite, and it grows as a film by repeating the adsorption and oxidization. However, in the solution being transferred, also, iron (II) ion reacts with oxidizer injected from the chemical tank 50, and a part thereof becomes iron (III) ion and turns into ferrite particles and iron (III) (ferric) hydroxide particles in the solution. Since those particles grow as time passes, they are easy to grow inside the surge tank 31 in which they stay for a long time. When the grown those particles are supplied again to the object portion that a film is to be formed, there is the possibility that they are taken in the magnetite film which is growing and uniform growth of the magnetite film is spoiled. Particularly, when ion (III) hydroxide is taken in, in some cases, radioactive cobalt is taken in when hematite is produced by dehydration reaction. In order to make the magnetite film grow uniformly on the object portion surface, it is preferable to remove particle-shaped products grown by side reaction, and it is effective to install a filter 80 for removing side reaction products at the exit side of the surge tank 31 in which the particle-shaped products are easy to grow. Thereby, an amount of particle-shaped products grown by side reaction to the object portion to be treated decreases and the magnetite film easily grows uniformly. Sixth Embodiment FIG. 12 shows a system of another embodiment of the film formation apparatus 30 according to the present invention. The present embodiment is a concrete example in which oxygen saturated water is formed as oxidizer which is stored in the chemical tank 50, in addition to the embodiment shown in FIG. 11. As shown in FIG. 12, water is filled in the chemical tank 50, and while recirculating the water by a recirculation pump 81, oxygen-containing gas is injected into the recirculating water from oxygen-containing gas injection equipment 82, whereby oxygen saturated water at the oxygen concentration of the oxygen-containing gas is formed. The oxygen saturated water is used as oxidizer for magnetite film formation. Thereby, waste does not occur from the oxidizer, and it is possible to reduce an amount of waste by the amount corresponding to the waste that does not occur from the oxidizer. Seventh Embodiment FIG. 13 shows a system of further another embodiment of the film formation apparatus 30 according to the present invention. In the present embodiment, oxygen-containing gas is supplied to micro-bubble generation equipment 83 from the oxygen-containing gas injection equipment 82 and oxygen-containing gas micro-bubble is injected into the processing solution piping 35 from the micro-bubble generation equipment 83, instead of the chemical tank for oxidizer in FIG. 6. The micro-bubble generation equipment 83 is provided on branch piping branched from the processing solution piping 35 to supply a part of processing solution supplied to the object portion to generate micro-bubble, and the branched piping is formed so that the part of processing solution is mixed with the oxygen-containing gas supplied from the oxygen-containing gas injection equipment 82 at the branched piping, generates oxygen-containing gas bubble, and is returned to the processing solution piping 35. The air bubble micro-bubbled is very slow in rising velocity in liquid phase, so that it is not separated soon into gas and liquid phases as in usual air bubble, and the air bubble as it is flows along the liquid phase in the field that flow exists. Therefore, according to the present embodiment, the oxidizer reaches the object portion to be treated as micro-bubble of oxygen-containing gas. Further, according to the present embodiment, since the oxygen contained in the micro-bubble does not react directly with iron (II) ion in liquid phase because the oxidizer of micro-bubble is used, it is possible to reduce an amount of iron (II) ion oxidized to iron (III) ion in liquid phase before reaching to the object portion to be treated, as compared with the sixth embodiment. Eighth Embodiment FIG. 14 shows a system of further another embodiment of the film formation apparatus 30 according to the present invention. In the present embodiment, in addition to the embodiment shown in FIG. 13, ascorbic acid injection equipment is provided. The ascorbic acid injection equipment is constructed so that ascorbic acid solution is injected into the processing solution piping 35 from a chemical tank 84 storing ascorbic acid solution via a pump 85 and the valve 86. The ascorbic acid solution in the chemical tank 84 is bubbled by nitrogen from the nitrogen bubbling equipment 71, and dissolved oxygen is removed therefrom. The ascorbic acid injected in the processing solution piping 35 is mixed with processing solution containing bubble of oxygen-containing gas reaches to the object surface to be treated together with iron (II) ion and pH adjuster. Meanwhile, a part of iron (II) ion has oxidized to iron (III) ion by a part of dissolved oxygen dissolved in liquid from micro-bubble of oxygen-containing gas. However, according to the present embodiment, ascorbic acid exists, so that iron (III) ion is reduced to the iron (II) ion, and it is possible to reduce a consumption amount of iron (II) ion. In this manner, most of the injected iron (II) ion as it is reaches to the object surface to be treated and adsorbed. A part of the adsorbed iron (II) ion is oxidized by oxygen contained in micro-bubble, and causes ferrite forming reaction together with iron (II) ion adsorbed without being oxidized to form magnetite film. Therefore, it is possible to form a magnetite film by using a chemical containing at least one of iron (II) ion and iron (III) ion instead of iron (II) ion in the present embodiment, and contacting the processing solution formed by mixing ascorbic acid as a reducer reducing iron (III) ion and a chemical adjusting the pH of the processing solution to 5.5 to 9.0 with a metal member surface. Ninth Embodiment FIG. 15 shows a system of further another embodiment of the film formation apparatus 30 according to the present invention. In the present embodiment, ultraviolet irradiation equipment 87 is used instead of the dissolution equipment 64 in FIG. 6. As in the first embodiment, in order to decompose formic acid and hydrazine after finishing of magnetite film formation, openings of the valve 57 and the valve 65 of the ultraviolet irradiation equipment 87 are adjusted, and a part of the processing solution is flowed into the ultraviolet irradiation equipment 87. At the same time, the valve is opened and the injection pump 67 is started, and the hydrogen peroxide in the chemical tank 68 is injected into the ultraviolet irradiation equipment 87. Thereby, a part of the hydrogen peroxide received ultraviolet irradiation produces OH radical, and the OH radical conducts oxidative decomposition of formic acid and hydrazine into carbon dioxide and water, and into nitrogen and water, respectively. As a result, it is possible to decompose the chemical contained in a waste solution after finishing magnetite film formation and to reduce a waste amount. Tenth Embodiment FIG. 16 shows a system of further another embodiment of the film formation apparatus 30 according to the present invention. The present embodiment is characterized in that micro-bubble generation equipment 89 is used instead of the decomposition equipment 64 in the embodiment shown in FIG. 6. As shown in FIG. 16, the present embodiment is constructed so that ozone gas is blown in the micro-bubble generation equipment 89 from ozone gas supply equipment 90. As in the decomposition equipment 64, openings of the valve 57 and the valve 65 at the exit side of the micro-bubble generation equipment 89 are adjusted, whereby a part of the processing solution in the micro-bubble generation equipment 89 is passed through. Thereby, the ozone gas blown in the micro-bubble generation equipment 89 becomes ozone micro-bubble in the processing solution containing formic acid and hydrazine. The ozone dissolved in the processing solution directly decomposes oxidatively formic acid and hydrazine or formic acid and hydrazine are oxidatively decomposed by OH radical generated by a reaction of ozone and water to becomes carbon dioxide, nitrogen and water. Thereby, a waste amount is reduced by decomposing chemicals contained in the waste after finishing of magnetite film formation to gas and water. Eleventh Embodiment Further another embodiment according to the present invention will be described, referring to FIG. 17. FIG. 17 shows a flowchart showing a procedure for performing noble metal deposition treatment which is usually not conducted immediately after chemical decontamination. The noble metal deposition treatment is conducted for suppressing stress corrosion cracking of components of a nuclear reactor. Usually it is conducted using about 48 hours when the temperature of reactor water reaches to 150° C. from a starting time of stopping operation in time when the nuclear reactor is going to be stopped or intermediately stopped to a time when the nuclear reactor is stopped. The reason that the treatment is not conducted immediately after chemical decontamination is that noble metal is difficult to deposit on the nuclear reactor component surface from which an oxide film is removed by the chemical decontamination. The noble metal deposition treatment is conducted after an oxide film grows to some extent. The reason that it is conducted at 150° C. is that the efficiency to deposition is best as a result of studying of a noble metal deposition amount under various temperature conditions. Namely, as shown in FIG. 17, magnetite film formation processing is conducted (S12) immediately after chemical decontamination (S11). Next, temperature elevation of the nuclear reactor is started (S13, S14) at the same time as an regular inspection terminates. After the temperature of reactor water is kept at 150° C. for 48 hours (S15), noble metal is injected (S16), and then temperature elevation of the nuclear reactor is restarted (S17) and the operation is shifted to the rated operation (S18). In this manner, according to the present embodiment, the magnetite film of metal oxide is formed on the object surface to be treated by the magnetite film processing immediately after the chemical decontamination, so that a noble metal element is easy to be deposited by the noble metal deposition treatment. When the temperature of reactor water reaches to 150° C. from starting of the nuclear reactor, injection of a noble metal compound including platinum and rhodium is started* and the condition is kept for about 48 hours, whereby it is possible to conduct the noble metal deposition treatment. Twelfth Embodiment FIG. 18 shows a flowchart showing another procedure for performing noble metal deposition treatment. In the present embodiment, after magnetite film processing is finished, immediately, a noble metal deposition treatment is conducted by using the film formation apparatus 30 as it is. As noble metal injection equipment, the oxalic acid injection equipment as shown in FIG. 6, for instance can be used. That is, a solution of a noble metal compound containing platinum and rhodium is contained into the chemical tank 43, and the solution is injected into the processing solution by using the injection pump 40. The temperature rises only to about 100° C., however, since no nuclear fuel is contained in the processing solution, it is unnecessary to worry about influence on zircalloy, so that it is possible to raise the concentration higher than the usual noble metal deposition treatment and extend the treatment time, whereby it is possible to deposit noble metal of a prescribed amount. Namely, as shown in FIG. 18, after chemical decontamination is conducted (S21), immediately, film formation processing of magnetite film is conducted (S22), successively, noble metal deposition treatment is conducted (S23), then elevation of the temperature of the nuclear reactor is started at the same time as the regular inspection finishes (S24, S25) and the nuclear reactor operation is shifted to the rated operation (S18).
	description	The present invention relates to a core-structure storing method. More specifically, the present invention relates to a core-structure storing method capable of further reducing irradiation on an operator. Generally, a core structure is stored in a storage container and then the storage container is carried into a nuclear-reactor storing housing. A typical storage container includes a storage-container upper cover, a storage-container body, and a storage-container bottom cover. Conventional core-structure storing methods are disclosed in Patent Documents 1 and 2. The conventional core-structure storing method disclosed in Patent Document 1 includes a step of craning down a storage-container upper cover above a core structure, which is inserted into a nuclear-reactor vessel in a cavity, while water in the cavity is discharged; a step of attaching the storage-container upper cover to the core structure; a step of craning down the storage-container body so that the bottom of the storage-container body is closely in contact with the top surface of the nuclear-reactor vessel; a step of craning up the core structure with the storage-container upper cover so that the core structure with the storage-container upper cover is positioned inside the storage-container body; a step of attaching the storage-container upper cover to the storage-container body; a step of craning up the storage-container body to be positioned at a certain level; a step of positioning a storage-container bottom cover under the storage-container body; and a step of attaching the storage-container bottom cover to the storage-container body by craning up the storage-container bottom cover. In the above conventional core-structure storing method, a moving device is set in the cavity and the storage-container bottom cover is placed on the moving device. The moving device is moved such that the storage container is positioned under the storage-container body. Because the core structure, which is stored in the storage container is highly radioactive, there is a demand for means and measures capable of further reducing irradiation on an operator. Patent Document 1: Japanese Patent Application Laid-open No. 2004-69354 Patent Document 2: Japanese Patent Application Laid-open No. 2003-215294 An object of the present invention is to provide a core-structure storing method capable of further reducing irradiation on an operator. To achieve the object, according to an aspect of the present invention, there is provided a method of storing a core structure in a storage container after the storage container is carried into a reactor containment, including a step of carrying, by using a moving device which is movable, a bottom cover of the storage container to a body of the storage container; a step of positioning the body and the bottom cover by adjusting a position of the bottom cover on the moving device while the bottom cover is supported by a free bearing; and a step of attaching the position adjusted bottom cover to the body. In the core-structure storing method, the storage-container body and the storage-container bottom cover are positioned by adjusting the position of the storage-container bottom cover on the moving device while the storage-container bottom cover is supported by the free bearing. In this method, because the storage-container bottom cover is supported by the free bearing, the storage-container bottom cover moves smoothly. Thus, the positioning of the storage-container bottom cover is performed smoothly. This advantageously facilitates the operation of the attachment of the storage-container bottom cover. Accordingly, the operation period is shortened, and this advantageously reduces irradiation on an operator. Furthermore, the camera monitoring of the positioning allows remote operations, and this advantageously reduces irradiation on an operator. According to another aspect of the present invention, there is provided a method in which the step of positioning includes positioning the body and the bottom cover, by using a guide pin and a guide bush, in a way that the guide pin is engaged with and guided by the guide bush, the guide pin being attached to any one of the bottom cover and the body, and the guide bush being attached to any one of the bottom cover and the body, to which the guide pin is not attached. According to the above core-structure storing method, the guide pins and the guide bushes are used for the positioning, and the storage-container body and the bottom cover are positioned in a way that the guide pins are engaged with and guided by the guide bushes, respectively. This advantageously facilitates the operation of attachment of the bottom cover. Specifically, the position of the bottom cover is adjusted smoothly, because the guide pins are engaged with the guide bushes while the bottom cover is supported by the free bearings. In this manner, the positioning is performed smoothly. This advantageously facilitates the operation of attachment of the bottom cover, and advantageously shortens the operation period, thus reducing irradiation of an operator. According to another aspect of the present invention, there is provided a method in which the step of attaching includes attaching the bottom cover to the body tightly by pushing up the bottom cover on the moving device. According to the above core-structure storing method, the bottom cover on the moving device is pushed up to be attached tightly to the storage-container body. This method advantageously facilitates the repositioning (reattachment) in the case of an operational error, or the like, compared to the case where the storage-container body that is heavier than the bottom cover is craned down and attached to the bottom cover. Furthermore, the clearance between the bottom cover and the storage-container body can be made smaller than that in a case where the bottom cover is craned up to be tightly attached to the storage-container body. This advantageously reduces the leakage of the radioactive substance from the storage container, thus reducing the irradiation of an operator. The bottom cover is pushed up by the jacks or an alternative lifting-up unit. According to another aspect of the present invention, there is provided a method in which the step of carrying includes fixing the bottom cover onto the moving device, and the step of positioning includes releasing the fixing of the bottom cover, so that the bottom cover is supported by the free bearing. According to the above core-structure storing method, the bottom cover is prevented from moving on the free beatings while being carried because the bottom cover is fixed on the moving device. This advantageously prevents an accident during the transfer of the bottom cover. In the core-structure storing method according to the present invention, a storage-container body and a storage-container bottom cover are positioned by adjusting the position of the storage-container bottom cover on a moving device while the storage-container bottom cover is supported by free bearings. This method facilitates the operation of the attachment of the storage-container bottom cover, so that the operation period is shortened. This advantageously inhibits an operator from being irradiated. 1 Nuclear reactor 2 Operation floor 21 Cavity 3 Nuclear-reactor vessel 31 Target 32 Protection plate for nuclear-reactor vessel 4 Core structure 5 Storage container 51 Upper cover of storage-container 52 Storage-container body 521 Guide bush 524 Compensating shield 525 Chain block 53 Bottom cover of storage-container 531 Guide pin 532 Guide hole for positioning bottom-cover 536 Bolts for attaching bottom-cover 61 Moving device 611 Guide pin for positioning bottom-cover 612 Jack 6121 Hydraulic-pump control panel 6122 Hydraulic pump 6123 Hydraulic hose 6124 Cable 613 Free bearing 62 Moving operation device 621 Winch 622 Wire 623 Emergency winch 624 Cable 625 Remote monitoring camera 631 Detecting unit 632 Cable 633 Wire 64 Auxiliary device 641 First monitoring camera 642 Second monitoring camera 643 Camera protection plate 65 Carriage rail 66 Greenhouse 67 Local discharging device 7 Shielding plate for nuclear-reactor vessel Exemplary embodiments of the present invention are described below in detail with reference to the accompanying drawings. The embodiments should not be interpreted so as to limit the present invention. The components of the embodiments include components that can be replaced easily by those skilled in the art and components substantially same as those of the embodiments. The modifications disclosed in the embodiment can be arbitrarily combined within the scope of those skilled in the art. Embodiment FIGS. 1 to 6 are schematic diagrams of a nuclear reactor of a nuclear power plant. FIGS. 7 to 24 are explanatory diagrams for explaining a core-structure storing method according to an embodiment of the present invention. [Configuration of Nuclear Reactor and Core-Structure Storing Method] A nuclear reactor 1 includes a nuclear-reactor containment whose outer surface is covered with an exterior shielding wall. An operation floor 2 is formed inside the nuclear-reactor containment (see FIGS. 1 and 2). A cavity 21 is formed inside the operation floor 2. A nuclear-reactor vessel 3 is positioned at the bottom of the cavity 21. A core structure (used core structure) 4 is placed inside the nuclear-reactor vessel 3. The used core structure 4 needs to be replaced from time to time. The core structure 4 includes, for example, an upper core structure and a lower core structure. The core structure 4 is lifted up with a crane (not shown) from the upper side of the cavity 21, and then the lifted core structure 4 is put inside a storage container 5. The storage container 5 includes an upper cover 51, a storage-container body 52, and a bottom cover 53. The upper cover 51 and the bottom cover 53 are substantially circular, and the storage-container body 52 is substantially cylindrical. The upper cover 51 is attached to an upper portion of the storage-container body 52 while the bottom cover 53 is attached to a bottom portion of the storage-container body 52. Water in the cavity 21 is drained out before putting the core structure 4 in the cavity 21. Next, the upper cover 51 is craned down and put on the core structure 4 in the nuclear-reactor vessel 3, and then the upper cover 51 is fixed to the core structure 4. The storage-container body 52 positioned above the nuclear-reactor vessel 3 is then craned down, so that the bottom of the storage-container body 52 is closely in contact with the upper surface of the nuclear-reactor vessel 3. The core structure 4 and the upper cover 51 are craned up to the inside of the storage-container body 52, and then, the upper cover 51 is attached to the storage-container body 52. The storage-container body 52 is craned up to a certain height, and then, the bottom cover 53 is positioned under the storage-container body 52. Then, the bottom cover 53 is attached to the storage-container body 52. In this manner, the core structure 4 is stored in the storage container 5. Thereafter, the storage container 5, containing the core structure 4, is carried outside the nuclear reactor 1. [Attachment Structure of Bottom Cover] A plurality of (for example, two) guide pins 531 is provided to the bottom cover 53 of the storage container 5 (see FIGS. 3 and 6). The guide pins 531 are provided to the outer surface of the outer periphery of the bottom cover 53, and each have a tip facing upward (toward the storage-container body 52 during attachment of the bottom cover 53). The guide pins 531 have a tapered shape, i.e., the diameter thereof gradually decreases toward the tip, and the body of the guide pin 531 is cylindrical. On the other hand, the storage-container body 52 is provided with guide bushes 521 such that the guide bushes 521 correspond to the guide pins 531, respectively (see FIGS. 4 and 5). The guide bushes 521 protrude from the side surface of the storage-container body 52. The guide bushes 521 each have a hole into which a corresponding guide pin 531 is inserted, so that the guide pins 531 fit respectively into the guide bushes 521. During the attachment of the bottom cover 53 to the storage-container body 52, the bottom cover 53 and the storage-container body 52 are positioned by inserting the guide pins 531 into the guide bushes 521. Once the bottom cover 53 and the storage-container body 52 are positioned, the bottom cover 53 and the storage-container body 52 are fixed with bolts 536 (see FIG. 21). The tips of the guide pins 531 are painted yellow. This makes it easier to see the tips of the guide pins 531 during the positioning of the bottom cover 53 with respect to the storage-container body 52. Marking (gap-checking scale) including a plurality of grooves and numerical marks is formed on the body of each of the guide pins 531. The marking helps checking how much the guide pins 531 are inserted into the guide bushes 521 during the positioning. On the other hand, the guide bushes 521 are painted yellow and red. Yellow and red paint helps determination on the allowable range of positional change of the guide bushes 521 with respect to the guide pins 531 during the remote monitoring. [Unit for Attachment of Bottom Cover] A bottom-cover attaching device is used in the process of attaching the bottom cover 53 to the storage-container body 52 (see FIGS. 1 and 2). The bottom-cover attaching device includes the moving device 61, a moving operation device 62, a position checking device, and a position-determination auxiliary device 64. The moving device 61 is a carrier (trolley) for carrying the bottom cover 53 (see FIGS. 1 and 2). The moving device 61 is craned down by a polar crane (not shown) into the cavity 21, and the moving device 61 is then placed on a carriage rail set in the cavity 21. The moving device 61 moves on a carriage rail 65 to be positioned under the storage container 5 that is hanging. A plurality of jacks 612 is provided on the top of the moving device 61 (see FIGS. 2 and 11). The jacks 612 supports the storage-container bottom cover horizontally during movement of the moving device 61 or attachment of the bottom cover 53, and is elevated or lowered so that the bottom cover 53 can move vertically. The jacks 612 are, for example, hydraulic jacks, and are connected to hydraulic pumps 6122 and a hydraulic-pump control panel 6121 via a hydraulic hose 6123. An operation pendant of the jacks 612 (not shown) is set in a remote operation floor, and is connected to the hydraulic-pump control panel 6121 via a cable 6124. Such a configuration allows driving operation of the jacks 612 from the remote operation floor. The hydraulic hose 6123 is supported by a cable bearing (energy chain) 6125 near the moving device 61. This prevents the hydraulic hose 6123 from tangling when the moving device 61 moves. If it is necessary to arrange the hydraulic-pump control panel 6121 and the hydraulic pumps 6122 are in a highly-radioactive area due to arrangement of the components of the nuclear-power plant, the above components are protected with, for example, a lead plate. A plurality of free bearings 613 is provided on the top of the moving device 61 (see FIGS. 2 and 11). The free bearings 613 and the jacks 612 are arranged on the same surface of the moving device 61. The free bearings 613 support the bottom cover 53 horizontally instead of the jacks 612 after the jacks 612 are lowered. The bottom cover 53 can move horizontally slightly on the moving device 61 (free bearings 613) in a state where the free bearings 613 support the bottom cover 53. The moving operation device 62 includes a winch 621 (see FIGS. 1 and 2). The winch 621 is provided in the cavity 21. The winch 621 winds up a wire 622, which is stretched across the moving device 61, to pull the moving device 61. The operation pendant (not shown) of the winch 621 is provided in the remote operation floor, and is connected to the winch 621 via a cable 624. The winch 621 of the moving operation device 62 is driven to pull the moving device 61 with the wire 622, so that the moving device 61 moves on the carriage rail 65. The moving device 61 is remotely monitored in the remote operation floor, using a plurality of remote monitoring cameras 625 arranged in the cavity 21. Such a configuration allows remotely operating the moving device 61. In addition to the winch 621, an emergency winch 623 is set in the cavity 21. The position checking device includes, for example, a detecting unit 631, an operation unit (not shown), and a display unit (not shown) (see FIGS. 1 and 2). The detecting unit 631 is, for example, a linear encoder and is disposed on the carriage rail 65. The operation unit and the display unit include, for example, a personal computer (PC) and a monitor thereof. The operation unit and the display unit are arranged in the remote operation floor and connected to the detecting unit 631 via a cable 632. The detecting unit 631 of the position checking device detects the amount of the positional change of the wire 633. Based on the detected amount, the operation unit calculates the position of the moving device 61, and the calculated position is displayed on the monitor. Hence, an operator can remotely check the position of the moving device 61 in the cavity 21 (on the monitor at the operation floor). The position-determination auxiliary device 64 includes first monitoring cameras 641 and second monitoring cameras 642 (see FIGS. 4 and 5). The first monitoring camera 641 monitors the positional relationship between the guide bush 521 and the guide pin 531 from a vertical direction, and is provided to the side surface of the storage-container body 52 above the guide bush 521. The second monitoring camera 642 monitor the positional relationship between the guide bush 521 and the guide pin 531 from a horizontal direction, and is provided to the side surface of the storage-container body 52 beside the guide bush 521. The first monitoring camera 641 and the second monitoring camera 642 of the position-determination auxiliary device 64 record images which represent the positional relationship between the guide bush 521 and the guide pin 531 during attachment of the bottom cover 53. The information about the recorded images is transmitted to the operation floor. Thus, the storage-container body 52 and the bottom cover 53 can be positioned while an operator at the remote operation floor checks the positional relationship between the guide bush 521 and the guide pin 531. Above each of the first monitoring cameras 641, a camera protection plate 643 is provided. Above each of the second monitoring cameras 642, a camera protection plate 644 is provided. The camera protection plates 643 and the camera protection plates 644 each have a function of preventing a plastic sheet of a greenhouse 66 from directly covering lenses of the monitoring cameras 641 and 642. In addition, the camera protection plates 643 and 644 each have a function of inhibiting the monitoring cameras 641 and 642 from changing the direction for monitoring due to the plastic sheet and protecting the monitoring cameras 641 and 642 from being damaged due to the weight of the plastic sheet. A framework is constructed in the cavity 21 to form the greenhouse 66. The greenhouse 66 includes the plastic sheet, and covers over the storage container 5 and the carriage rail 65 (therefore, radioactive dust is prevented from scattering from the cavity 21 to the operation floor). The air in the greenhouse 66 is discharged by a local discharging device 67 provided in the cavity 21. The local discharging device 67 includes a removing device (not shown) for removing a radioactive substance in the air. The local discharging device 67 prevents the radioactive substance from scattering and leaking during attachment of the bottom cover 53. [Method of Attaching Storage-Container Bottom Cover] Prior to the process of attaching the bottom cover 53, the moving device 61 is placed on the carriage rail 65 in the cavity 21 (see FIG. 7). A shielding plate 7 for shielding the nuclear-reactor vessel 3 is previously hung from the moving device 61 outside the nuclear reactor. The nuclear-reactor vessel shielding plate 7 is used to close the opening of the nuclear-reactor vessel 3 while an operator is working inside the cavity 21, thereby reducing the leakage of the radioactive substance from the nuclear-reactor vessel 3. The bottom cover 53 is craned down into the cavity 21 and placed on the moving device 61 (see FIG. 7). In the bottom surface of the bottom cover 53, a plurality of guide holes 532 for positioning the bottom-cover 53 is provided along the circumference of the bottom cover 53 (see FIG. 3). On the top surface of the moving device 61, a plurality of guide pins 611 for positioning the bottom-cover 53 is provided at positions corresponding to the guide holes 532, respectively (see FIG. 6). The guide pins 611 are inserted respectively into the guide holes 532, so that the bottom cover 53 is placed at a certain position on the moving device 61. The storage-container body 52 is craned down toward the nuclear-reactor vessel 3, and the bottom surface of the storage-container body 52 (the surface of the bottom opening portion to be attached) comes closely in contact with the top surface (the flange surface) of the nuclear-reactor vessel 3 (see FIGS. 8 and 9). On the top surface of the nuclear-reactor vessel 3, targets 31 for attachment of the storage-container body 52 are provided (see FIG. 9). On the inner periphery of the flange of the nuclear-reactor vessel 3, protection plates 32 for protecting the nuclear-reactor vessel 3 are provided. In the process of attachment of the bottom cover 53, first, a chain jack (not shown) winds up the chains to crane up the storage-container body 52 by a few meters (see FIGS. 7 and 8). Accordingly, a space is formed below the storage-container body 52, in which the bottom cover 53 (and the moving device 61) is placed. The moving device 61 is moved horizontally on the carriage rail 65, and thus positioned below the storage-container body 52 (above the nuclear-reactor vessel 3) (see FIGS. 10 and 11). As illustrated in FIG. 11, the storage-container body 52 includes a vertical protruding portion 1101 that protrudes from its bottom. The moving device 61 is moved by using the moving operation device 62 and the position checking device 63 while an operator at the remote operation floor checks the position of the moving device 61. On the moving device 61 that is moving, the bottom cover 53 is supported by the jacks 612 (see FIG. 11). This prevents the bottom cover 53 from sliding on the free bearings 613 while the moving device 61 is being moved. The bottom cover 53 includes a horizontal protruding portion 1102 that protrudes from the circumference portion of the bottom cover 53. Subsequently, the storage-container body 52 and the bottom cover 53 are positioned horizontally (the position of the guide bushes 521 with respect to the guide pins 531 are adjusted) (see FIG. 12). Specifically, while the first monitoring camera 641 monitors the horizontal positional relationship between the guide pin 531 and the guide bush 521 from the vertical direction, the tip of the guide pin 531 is inserted into the guide bush 521. When the positions of the guide pin 531 and the guide bush 521 are not correctly positioned, a jack at a temporary craning equipment (craning structure) 22 is operated to horizontally moves and rotate the storage-container body 52. In this manner, the positions are adjusted. After the positioning, the chain jack (not shown) unwinds the chains to crane down the storage-container body 52. After the tip of the guide bush 521 protrudes from the top edge of the guide pin 531 (penetrates through the hole), the craning down of the storage-container body 52 is stopped (see FIG. 13). At this stage, clearance is present between the tip of the guide pin 531 and the inner periphery of the guide bush 521 having the hole, because the tip of the guide pin 531 is tapered. In a range where the clearance allows, the position of the bottom cover 53 can be changed horizontally slightly with respect to the storage-container body 52. The jacks 612 of the moving device 61 are lowered to the lowest position to place the bottom cover 53 on the free bearings 613 (see FIGS. 14 and 15). Accordingly, the entire load of the bottom cover 53 is put on the free bearings 613, and thus, the position of the bottom cover 53 can be changed horizontally slightly on the moving device 61 by rolling the free bearings 613. Because the tips of the guide pins 531 penetrate through the guide bushes 521, the storage-container body 52 and the bottom cover 53 are positioned horizontally within a predetermined range. The craning down of the storage-container body 52 is restarted and the storage-container body 52 is lowered, so that the guide bushes 521 comes in contact with the bodies of the guide pins 531. Accordingly, an external force corresponding to the shape of the tip of the guide pin 531 (tapered shape) is put on the guide bush 521. Because the free bearings 613 support the bottom cover 53, the external force smoothly changes the position of the bottom cover 53 horizontally slightly along the shapes of the tip portion of the guide pins 531. In this manner, the storage-container body 52 and the bottom cover 53 can be positioned horizontally smoothly. When the clearance (vertical distance) between the connecting portion of the storage-container body 52 and the connecting portion of the bottom cover 53 (flange) reaches a certain amount, the craning down of the storage-container body 52 is stopped (see FIGS. 16 and 17). The second monitoring camera 642 monitors the vertical positional relationship between the guide pin 531 and the guide bush 521 from the horizontal direction. Specifically, the position of the guide bush 521 with respect to the marking of the body of the guide pin 53 is monitored. Based on the monitoring result, the remote operation is performed for the positioning of the storage-container body 52 in the vertical direction. The jacks 612 are elevated to elevate the bottom cover 53 (see FIGS. 18 and 19). After the bottom cover 53 is elevated to the position distant from the storage-container body 52 by a certain distance (for example, 5 millimeter), the jacking up is suspended. In this case, the bottom cover 53 is being elevated while the second monitoring camera 642 monitors the vertical positional relationship between the marking of the guide pin 531 and the guide bush 521 from the horizontal direction. Therefore, depending on the monitoring, the jacks 612 of the moving device 61 are remotely operated, and whereby the positional relationship between the bottom cover 53 and the storage-container body 52 are adjusted. The jacks 612 are inched up while the marking of the guide pins 531 is checked until the storage-container body 52 and the bottom cover 53 are closely in contact with each other. When the bottom cover 53 is closely in contact with the storage-container body 52, the jacking up is stopped. The pendant switch of the jacks 612 is provided with, in addition to a button for elevati ng and lowering the jacks 612, a button for stopping driving the jacks 612 when a pressure equal to or higher than a set pressure is detected. The buttons are used to operate the jacks 612 so that the storage-container body 52 and the bottom cover 53 are closely in contact with each other. The radiation dose in the cavity 21 is measured and it is judged whether an operator can enter the cavity 21. When it is judged that the operator can enter the cavity 21, the shielding plate 7 is craned down from the bottom of the moving device 61 to close the opening of the nuclear-reactor vessel 3 (see FIG. 20). The operator operates a manual chain block, so that the shielding plate 7 is craned down. The shielding plate 7 reduces the leakage of the radioactive substance from the nuclear-reactor vessel 3, and thus, the operator who works in the cavity 21 can be further inhibited from being irradiated. The operator manually fixes the bottom cover 53 and the storage-container body 52 with the bottom-cover attaching bolts 536 (see FIG. 21). The moving device 61 is moved on the carriage rail 65 toward the canal, so that the moving device 61 can be positioned at the original position (see FIG. 22). The chain jack unwound the chains, so that the storage container 5 (the storage-container body 52 and the bottom cover 53) is lowered and stopped at a certain position. Thereafter, the first monitoring cameras 641 above the guide pins 531 are detached, and an eye bolt and a chain block 525 are attached to a compensating shield 524 (see FIG. 23). The guide pins 531 are craned up by the chain blocks 525, so that the guide pins 531 are detached from the bottom cover 53. Similarly, the guide bushes 521 are detached from the storage-container body 52. The detachment operations are performed in the greenhouse 66. Hence, the dust is prevented from leaking outside. Thereafter, the chain jack winds up the chains, and thus, the storage container 5 (see FIG. 24) is craned up. A temporary floor (not shown) is reconstructed and the bottom-cover attaching bolts 536 are firmly fastened, so that the storage-container body 52 and the bottom cover 53 are firmly fixed. [Effect] In the core-structure storing method, the storage-container body 52 and the bottom cover 53 are positioned in a way that the position of the bottom cover 53 is adjusted while being supported by the free beatings 613 on the moving device 61 (see FIGS. 14 and 15). Tn this method, the storage-container body 52 and the bottom cover 53 are positioned smoothly because the bottom cover 53 can be moved (moved horizontally) while the bottom cover 53 is supported by the free bearings 613. This advantageously facilitates the operation of the attachment of the bottom cover 53. Accordingly, the operation period is shortened, and this advantageously reduces irradiation on an operator. Furthermore, in the core-structure storing method, the guide pins 531 and the guide bushes 521 are used for the positioning, and the storage-container body 52 and the bottom cover 53 are positioned in a way that the guide pins 531 are engaged with (inserted into) and guided by the guide bushes 521, respectively (see FIGS. 14 and 15). This advantageously facilitates the operation of attachment of the bottom cover 53. Specifically, the position of the bottom cover 53 is adjusted smoothly, because the guide pins 531 are engaged with the guide bushes 521 while the bottom cover 53 is supported by the free bearings 613. In this manner, the positioning is performed smoothly. This advantageously facilitates the operation of attachment of the bottom cover 53, and advantageously shortens the operation period, thus reducing irradiation of an operator. In the above method, the guide pins 531 are attached to the bottom cover 53 and the guide bushes 521 are attached to the storage-container body 52. However, the structure is not limited to this, and alternatively, the guide bushes 521 can be attached to the bottom cover 53 and the guide pins 531 can be attached to the storage-container body 52. In the core-structure storing method, the bottom cover 53 on the moving device 61 is pushed up to be attached tightly to the storage-container body 52. This method advantageously facilitates the repositioning (reattachment) in the case of an operational error, or the like, compared to the case where the storage-container body 52 that is heavier than the bottom cover 53 is craned down and attached to the bottom cover 53. Furthermore, the clearance between the bottom cover 53 and the storage-container body 52 can be made smaller than that in a case where the bottom cover 53 is craned up to be tightly attached to the storage-container body 52 (see Patent Document 1). This advantageously reduces the leakage of the radioactive substance from the storage container 5, thus reducing the irradiation of an operator. The bottom cover 53 is pushed up by the jacks 612 or an alternative lifting-up unit. In the core-structure storing method, the bottom cover 53 is prevented from moving on the free beatings 613 while being carried because the bottom cover 53 is fixed on the moving device 61. This advantageously prevents an accident during the transfer of the bottom cover 53. In the above method, it is preferable that as an elevating unit for elevating and lowering the bottom cover 53, for example, the jacks 612 be adopted. In the method, the jacks 612 are elevated during the carriage of the bottom cover 53, and are lowered (jacked down) during the positioning so that the bottom cover 53 is placed on the free bearings 613. During the attachment of the bottom cover, the jacks 612 are elevated (jacked up), so that the bottom cover 53 is tightly in contact with the storage-container body 52 (see FIGS. 10 to 19). This leads to an advantage that the bottom-cover attaching device can be configured with a simple configuration including the elevating unit (the jacks 612) and the free bearings 613. It is preferable that a winch (not shown) be additionally provided as a backup winch for emergency. This leads to an advantage that, in case the normally used winch malfunctions, the backup winch can be used, thus improving the operational environment. In the core-structure storing method, because the moving operation device 62 for moving the moving device 61 is provided, an operator at the operation floor can move the moving device 61 while checking the position of the moving device 61. This advantageously further reduces irradiation on an operator. In the core-structure storing method, because the position checking device for checking the position of the moving device 61 is provided, an operator at the operation floor can operate the moving device 61 while checking the position of the moving device 61. This advantageously reduces irradiation on an operator. In the core-structure storing method, the position-determination auxiliary device 64 for checking the positions of the storage-container body 52 and the bottom cover 53 is provided. Hence, an operator at the operation floor can position the storage-container body 52 and the bottom cover 53 while checking the positional relationship between the guide bushes 521 and the guide pins 531. This advantageously reduces irradiation on an operator. In the core-structure storing method, the jacks 612 that support the bottom cover 53 on the moving device 61 are provided, and an operator at the remote operation floor can remotely operate the jacks 612. This advantageously further reduces irradiation on an operator. Industrial Applicability As described above, the core-structure storing method according to the present invention is useful because this method can further inhibit an operator from being irradiated.
	summary
053435044	description	DETAILED DESCRIPTION The spring constant gauge of the present invention, generally indicated at 10 in FIG. 1, is illustrated in position to measure the spring constant of a spring 12 assembled with a pair of ferrules 14 and 16 of a nuclear fuel bundle spacer, generally indicated at 18. For details of the spacer construction, reference may be had to the above-cited Matzner et al. patent, the disclosure of which is expressly incorporated herein by reference. As mentioned above in connection with this patent, spring 12 is a double-acting loop spring of generally elliptical shape having one resilient side 12a acting in ferrule 14 and a second resilient side 12b acting in ferrule 16. Thus, spring side 12a exerts a force on a fuel rod (not shown) inserted through ferrule 14 to bias it against inwardly formed stops 20, best seen in FIG. 2, thereby maintaining the fuel rod centered within the ferrule bore. Spring side 12b performs the same function with respect to a fuel rod inserted through ferrule 16. Gauge 10 is uniquely structured to accurately measure the spring constant of the individual spring sides 12a, 12b to determine if spring 12 meets quality assurance standards. Thus, as seen in FIGS. 1 and 2, gauge 10 includes a gauge body, generally indicated at 22, having a depending cylindrical probe 24 for insertion into a ferrule, ferrule 14 in the drawing. The upper end of the gauge body is joined with a handle 25 to accommodate manual or robotic manipulation of the gauge into spring constant gauging position. An integral flange 26, extending laterally from the gauge body, serves to mount a cylindrical alignment rod 28 via a shouldered bolt 30 extending through a clearance hole 32 in the flange and threaded into a counter-sunk and tapped axial bore 34 in the alignment rod. Thus, as probe 22 is inserted into ferrule 14, alignment rod 28 is inserted into ferrule 16. The diameters of the alignment rod and the probe are each equal to the nominal diameter of a fuel rod, and thus their insertions into the ferrule bores simulate the presence of fuel rods. The shoulder of bolt 30 bottoms out on the shoulder of bore 34 before the bolt head can clamp down on flange 26 to provide for limited floating motion of the alignment rod relative to the probe body. This feature accommodates acceptably minor nonparallelism between the axes of the alignment rod and the probe as spring 12 forces them against stops 20 and into centered portions in their respective ferrule bores. A plunger 36 is received in a bore 38 formed in the probe, which is oriented transversely to probe axis 39. The plunger is loosely captured in this bore by a roll pin 40 passing through a transversely elongated hole 42 in the plunger. Thus the plunger is free for limited reciprocation in its bore. The axial location of the plunger is such that its face 43, which is of a curvature corresponding to that of a fuel rod peripheral surface, confronts and is acted upon by side 12a of the spring, while spring side 12b is being loaded by the presence of the alignment rod in ferrule 16. The plunger is then subjected to the fuel rod-centering force exerted by spring side 12a in ferrule 14. The gauge body 22, including probe 24, is formed with an axially elongated slot 44 opening at its lower end into transverse bore 38 for accommodating an elongated arm 46 pivotally mounted to the gauge body at a mid-length point by a roll pin 48. The lower end of the arm extends into a slot 50 formed in the plunger to present a contact surface in engagement with the plunger at the bottom surface of the slot. The upper end of the arm is positioned to engage the tip 51 of a miniature load cell 52 slidingly received in a transverse bore 54 formed in the gauge body. The load cell may be of a conventional button strain gauge type, such as an Omega model LCK-25 available from Omega Engineering of Stamford, Conn. A micrometer, generally indicated at 56, is mounted to gauge body 22 by means of a bushing 58 having external threads engaging a tapped hole 60 in the gauge body and internal threads engaging a threaded collar 62 of the micrometer. The spindle 64 of the micrometer extends coaxially with bore 54 into abutting engagement with the back end of load cell 52. Rotation of the micrometer thimble 66 adjusts the extension of spindle 64 in convention fashion, which is seen to be effective in linearly varying the position of the load cell in bore 54. By virtue of the mechanical coupling provided by pivotal arm 46, variation of the linear position of the load cell in its bore 54 correspondingly varies the linear position of plunger 36 in its bore 38, resulting in deflection variation of side 12a of spring 12. The micrometer can be easily calibrated to a zero deflection reading by varying the extension position of spindle 64 until the output of load cell 52 is reduced to a zero spring force reading. Then thimble 66 is rotated to produce a measured deflection of side 12a of spring 12 and a spring force reading is taken from the load cell. Preferably spring force readings are taken at several spring deflection values and analyzed to determine an appropriate spring constant value i.e., the ratio of spring force or load to spring deflection, for each of the sides 12a, 12b of the springs. As seen in FIG. 1, spring deflection readings may be taken from a micrometer display 68. Preferably however, particularly in the case of spring constant gauging in a "hot cell", spring force measurements by the load cell and spring deflection measurements by the micrometer are read out over leads 70 and 72, respectively, to a remote spring constant indicating meter 74. Moving gauge 10 between gauging positions and rotation of the micrometer thimble are effected by manually controlled manipulators or automatically controlled "pick and place" robotic apparatus. A micrometer particularly suited for application in the present invention is a Series 350-712 digimatic micrometer head available from Mitutoyo of Paramus, N.J. It will be appreciated that, rather than a simple spring constant meter, the load cell and micrometer readouts may be fed to a data acquisition system where they are processed and recorded for subsequent printout of the spring constants of the individual springs identified by their locations in spacer 18. Completing the description of the gauge construction, an L-shaped spacer cover includes a vertical portion 76 affixed to gauge body 22 by screws 78 and a lateral portion 80 having holes 82 through which alignment rod 28 and probe 24 extend. The lateral portion serves a spacing function by engaging the upper edges of the ferrules to control the depth of alignment rod-probe insertion and thus ensure that the plunger face is properly aligned with the spring side whose spring constant is to be measured. The present invention thus provides a compact gauge which is conveniently inserted into the multiple ferrules of a nuclear fuel bundle spacer in succession to accurately measure the spring constants of the multiplicity of springs in the spacer. This quality assurance test can be performed expeditiously to qualify spacers for service in a reactor. Other embodiments of the invention will be apparent to those skilled in the art from consideration of the specification and practice of the invention disclosed herein. It is intended that the specification be considered as exemplary only with the true scope and spirit of the invention being indicated by the following claims.
	summary
043895719	abstract	A means is provided for compensating for third and higher order aberration in charged particle beam devices. The means includes two sextupoles with an intermediate focusing lens, all positioned between two focusing lenses.
054835720	description	DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS FIG. 1 shows a schematic diagram of an x-ray apparatus according to the invention. A carrier 1 in the form of a C-shaped arm carrier supports an x-ray source unit 2 and an x-ray detector 3 in the form of an x-ray image intensifier facing the x-ray source unit. The carrier is moveable through a sleeve 4 and by means of a bearing 5. The carrier is attached by way of the bearing 5 to a predominantly vertical stand 6 which is rotatably mounted to a ceiling of a room in which the x-ray apparatus is set up. Thus, the x-ray source unit and the x-ray detector are moveable so as to orient a beam-path 7 of x-radiation emitted by the x-ray source unit so that various projections can be applied for making an x-ray image. A falter-arrangement 10 comprising absorption filters 11 and filter-drive-means 12 is incorporated in the x-ray source unit 2, the filter-arrangement 10 being positioned between the x-ray source 2a and the x-ray detector 3. When an x-ray image is being made, overexposure in certain areas of the x-ray image is to be avoided. Overexposure is liable to occur when regions of very low x-ray absorption and adjacent regions of high x-ray absorption are imaged in the same x-ray image. For instance, when imaging a patient's heart having a high x-radiation absorption, the surrounding lung-tissue having a low x-ray absorption will be overexposed. By covering regions of high x-ray transmittance by positioning absorption filters 11 in the x-ray beam path 7 overexposure by regions of high transmittance are avoided. Correct positioning of the absorption filters depends on the orientation of the x-ray beam-path with respect to an object that is being examined, because the object is imaged according to a projection corresponding to the orientation of the beam-path. Filter-drive-means 12 are provided for positioning the absorption filters 11. The filter-drive-means are connected to drive-control-means 13 to which also carrier-drive-means 14 are coupled. For orientating the beam-path 7 and adjusting the absorption filters accordingly, a selection-means 15 is provided for supplying a positioning signal to a geometry controller 19. Furthermore, a posture sensor 18 is further provided to produce a posture signal pertaining to the position of the carrier. The posture signal is supplied to the geometry-controller. The geometry-controller 19 supplies a positioning signal to the drive-control-means 13, for positioning the carrier 1 in correspondence with a required beam-path orientation. The carrier-drive means 14 is controlled by the drive-control-means 13 for displacing the carrier 1. The geometry-controller 19 also applies said positioning signal to a memory-device 16 for selecting a filter-position. To this end, one or a plurality of filter-position curves is stored in the memory-device 16; each of the filter-position curves represents filter-adjustments as a function of the position of the carrier 1. When a positioning signal is supplied to the memory-device 16, a filter-adjustment signal is supplied by the memory-means to the drive-control-means 13. In this manner, filter-control-means 17 is constituted by the drive-control-means in combination with the memory device 16. Subsequently, the drive-control-means controls the filter-drive-means 12 so as to adequately position the absorption filters 11. The positioning of the absorption filters will be described in more detail with reference to FIGS. 2 and 3. FIG. 2 shows a side elevation of a filter-arrangement for use in an x-ray apparatus according to the invention. The absorption filters have a wedge-shaped extremity so that a gradual transition between a covered region and an uncovered region is obtained and consequently steep gradients in the x-ray image are avoided. The absorption filters are mounted to filter-drive-means 12 in such a way that the filters are moveable into and out of the x-my beam path 7 as indicated by arrows 20. Furthermore the absorption filters are rotatable about axes 22 as indicated by arrows 21. FIG. 3 shows a top view of an absorption filter incorporated in a filter-arrangement shown in FIG. 2. The absorption filter has an x-ray absorbing part 23, e.g. containing lead, and an x-ray transmitting part 24 which has the shape of a section of a substantially circular disk. Thus, when rotating either one of the absorption filters 11 about its axis 22 the part of the x-ray beam that impinges on the absorption filter is varied and consequently, the part of the x-ray image being covered by the absorption filter is varied. FIG. 4a shows a graph of a first adjustment curve, representing a rotation angle of the absorption filter as a function of the beam-path orientation. The rotation angle .phi. according to which the absorption filter is positioned is given as a function of the angle .theta., which determines the orientation of the beam path with respect to a patient to be examined. For further illustration, a few particular orientations of the absorption filter have been presented as small diagrams, along the abscissa. When the carrier is positioned in accordance with .theta.=-80.degree., then the absorption filter is placed in an orientation of .phi.=-50.degree.. When the carrier is moved so as to change the beam-path orientation to .theta.=10.degree., is increased to .phi.=86.degree.. Finally, when the beam-path orientation angle is increased further, the angle of orientation of the absorption filter is slightly decreased. FIG. 4b shows a graph of a second adjustment curve, representing translation of the absorption filter with respect to the beam-path as a function of beam-path orientation. As the curve shows, the absorption filter is moved further into the beam-path as the beam path is inclined from perpendicular incidence, viz. .DELTA. has a minimum when .theta.=0.degree. and .DELTA. increases as .vertline..theta..vertline. increases. The features of the adjustment curves are determined by anatomical properties, notably the location of a heart of a patient with respect to surrounding lungtissue. In practice, these curves can be determined empirically, viz. by manually positioning absorption filters, for a multitude of carder orientations, so as to obtain optimum image quality. It appears that a single set of adjustment curves is already quite satisfactory for a multitude of patients. Further improvement can, however, be achieved by determining sets of adjustment curves, each set pertaining to a class of patients, such as corpulent or slender patients, adults or infants, etc.
	description	Removal of tritium (3H, T) from all the materials that are contaminated thereby is a fundamental process for all the plants that operate with tritium. Said process has two important functions: the first regards the limitation of the release of tritium outside plants; the second is linked to the possibility of storing materials already completely de-categorized (with much lower levels of contamination and hence with much lower storing costs). So far the processes used for recovering tritium from said materials envisage as by-product tritiated water, with a low concentration of tritium, and, at times, further radioactive gaseous flows. The process according to the present invention proposes a solution to this problem. In said innovative process, in fact, tritium is recovered from the waste thanks to a thermal treatment (T<120° C.), in a slightly oxidizing atmosphere. The invention envisages the use of a reactor in which the reaction for removal of tritium from the waste takes place, said waste being recovered by means of a flow of moist inert gas in which an extremely low percentage of humidity is used. The heated waste releases a current of tritiated gases, this current of gases being removed from the reactor via the moist inert gas, which conveys it into a membrane reactor for decontamination. The membrane reactor, in fact, is able to remove selectively the tritium present in the mixture of gases: there is thus the dual advantage of purifying the mixture of gases and of recovering the tritium contained therein. A better understanding of the invention will be gleaned from the ensuing detailed description and with reference to the annexed figures, which illustrate, purely by way of non-limiting example, a preferred embodiment. Handling of radioactive waste is a critical problem both for plants that use tritium and for fusion machines that envisage tests with tritium. The so-called “soft housekeeping waste” is produced throughout the life cycle and also during disposal of the aforesaid plants and machines (JET, ITER, DEMO); their treatment is hence an issue of fundamental importance. Within a nuclear plant, it is estimated that the amount of housekeeping waste produced is approximately 0.2 kg/h per worker. The so-called “soft housekeeping waste” comprises gloves, overshoes, overalls, gas filters, paper, etc. In order to arrive at the most promising treatment processes, in the last few years there have been studied on a laboratory scale various techniques aimed at removal of tritium from this kind of waste [2]. The major difficulty consists in reaching a compromise between an adequate decontamination factor, which results in a potential de-categorization of the final waste, and an acceptable volume of product resulting from the detritiation process. Amongst all the techniques illustrated in the references [3], a process already studied with good characteristics of industrial effectiveness and feasibility regards continuous combustion with pure oxygen at atmospheric pressure. In this process, the gases produced during combustion need to be treated before being stored. FIG. 1 illustrates the complete scheme of said known process, in which there is a gas-treatment unit that envisages the following steps: separation of the solids and of the flow of the gases through a cyclone; condensation of the tritiated water (4° C.); neutralization of the non-condensable gases through a solution of KOH (50 wt %); and, as final operation, absorption on a molecular-sieve bed of Q2O, CO2 and NOx (the letter Q indicates generically a hydrogen isotope and hence also tritium). As regards the process described in the reference [4] and illustrated in FIG. 2, it is based upon injection of vapour into a container (1) where the (dried) waste is previously loaded. The water vapour produced in the vapour generator (7) is set in (1) in strict contact with the waste and promotes extraction of tritium therefrom. The contaminated water vapour is sent into a two-stage condenser through the piping (9). The first stage (11) operates at 15° C., whilst the second stage (13) operates at the temperature of liquid nitrogen: in this way, all the tritium can be gathered in the form of tritiated water in the tank (15). The main disadvantage of this known method is represented by the fact that in order to separate tritium from the tritiated water a further detritiation process is necessary. The process according to the present invention advantageously enables simultaneous removal of tritium from laboratory waste (de-categorization) and its recovery in the gaseous phase (valorization). In particular, according to the invention, use of two devices in series is envisaged: a reactor RT for removal of tritium, which carries out detritiation of the waste, and a membrane reactor RM in which tritium is recovered in the gaseous form. The membrane used is preferably, but not exclusively, made of palladium alloy. In the ensuing sections, using as guide a block diagram and a flowchart, the operating modalities of the process and the main functions of the two reactors RT and RM are described. 2.1 Block Diagram The ensuing description of the plant for implementing the process according to the invention refers to the diagram appearing in FIG. 3. Inert Gas The inert gas comes, for example, from commercial cylinders: said gas can be helium or argon, or some other gas suited for the purpose, the flow of which is controlled and recorded. Demineralized Water Demineralized water, without tritium, is used as decontaminating agent; the amount of water is controlled, optimised, and recorded. This water is stored at ambient temperature in an appropriated container. Evaporating Zone In this area evaporation of water and mixing thereof with the inert gas take place in a device (gas mixer) in which both of the flows (liquid and gas) are mixed and heated. The heat supplied for heating and evaporation is controlled and recorded. The internal volume of said evaporation device must be optimised in order to avoid dead volumes. Consequently, at outlet from this mixing area there is a moist gas that feeds the reactor RT in which the detritiation reaction takes place. The presence of water is important since it favours transfer of tritium from the waste to the inert gas. Detritiation Reactor The detritiation reactor RT is a hermetically sealed container in which the reaction of removal of tritium from the waste takes place. Detritiation can be considered as decontamination from tritium. This operation is favoured at temperatures higher than atmospheric temperature; for this reason, according to the type of the waste (mainly plastic) the reactor is kept at a temperature of 120° C. It should also be emphasized that, in addition to the temperature, another important parameter that regards sizing of the reactor is the time of stay of the moist gas. This stay time must be long enough to guarantee transfer of tritium and of the (hydrogen) isotopes without, however, high values of concentration of tritium in the moist inert gas leaving the reactor RT being reached. For this purpose it is necessary to guarantee a high flow rate of the moist inert gas (approximately 30 or 50 times the internal volume of the reactor per hour). Another parameter to be considered is the time of stay of the waste to be detritiated: this must be long enough to guarantee that the required decontamination values are reached. Membrane Reactor A membrane reactor is a device that combines in a single device the properties of separation of a membrane with the characteristics of a catalytic-bed reactor. It enables removal of one (or more) products from the site of the reaction itself, thus enabling reaction conversions that are higher than those of a traditional reactor. Said device has been developed at the ENEA laboratories of Frascati. Swamping Gas The swamping gas is supplied by commercial cylinders: to obtain the maximum isotope-exchange effect in the detritiation reactions pure hydrogen is preferably used. The flow of the swamping gas is controlled and optimised according to the characteristics of the waste and to the process requirements (amount of waste, level of tritium content, decontamination factor, etc.). In the plant (FIG. 3) there may be distinguished various sections: the utilities (upstream and downstream); the detritiation reactor RT; and the gas-treatment unit RM. The upstream utilities basically regard supply of inert gas (He or Ar) and swamping gas (H2), which is ensured by commercial cylinders. The flow of the gases is controlled through flow controllers and recorded in order to monitor the flows of the incoming gases. A water injection system and an evaporator are connected to the inert-gas line in order to obtain moist gas. The degree of humidity is determined knowing and controlling the exact amount of water and gas. The temperature of the evaporator is kept at 120° C. in order to prevent condensation in the connectors between the evaporator and the reactor RT. The material used for the connectors, the connections, and the valves is preferably stainless steel in order to guarantee an optimal seal and reduce corrosion phenomena. 2.2 Flowchart The waste to be treated is placed within a detritiation reactor RT, the scheme of which is illustrated in FIG. 5A. In order to homogenize better the waste a preliminary shredding step should be envisaged. Within the reactor RT the waste is positioned on a bottom perforated plate P (fixed-bed reactor) so as to enable circulation of the moist inert gas through the waste itself. Provided in the top flange of the reactor RT is a metal inlet tube T that traverses the area for containing the waste and terminates on the bottom of the reactor RT itself, underneath said perforated plate P, with the function of transporting the moist inert gas underneath the waste. Also provided in said top plate is a duct F for enabling outflow of the current of the gases containing tritium. In order to facilitate thermal desorption, the reactor RT is introduced into an oven (see FIG. 5A), which controls and regulates the temperature of the reactor RT around the pre-defined set point (120° C.) The control of the temperature is performed, for example, via a thermocouple positioned on the outer wall of the reactor. The reactor RT operates at atmospheric pressure or at a pressure slightly higher than atmospheric pressure. Gas-Treatment Unit The unit for treating the gases containing tritium basically consists of a membrane reactor RM that preferably uses membranes made of Pd—Ag alloy. It should be noted that the alloys commonly used for the permeator tubes are palladium-based alloys, such as, for example, PdCu, but also employed are metal alloys with a base of Ni, Nb, V, Ta, Ti. The thicknesses of practical interest for said dense metal tubular membranes substantially fall in the range 50-200 μm. The membrane reactor used in the process described herein is provided with thin-walled permeator tubes (as has been already said, the wall thickness falls within the range 50-150 μm) made of a commercially available palladium-silver alloy (23-25 wt % of Ag). The permeator tube (see FIG. 6) is housed within the module preferably according to a configuration of the “finger-like” type. Heating of the membrane reactor RM, the working temperature of which falls within the range 300-400° C., is obtained through the passage of an electric current through the tube itself to obtain heating of an ohmic type. The gas containing tritium to be treated is sent into the shell of the reactor RM, whilst the swamping gas, which in the example described herein is a current of pure hydrogen, is sent into the lumen of the membrane (as represented in FIG. 6). Alternatively, the current of gas to be treated and the current of pure swamping hydrogen can be reversed. The operating scheme is substantially that of a reactor of a PERMCAT type [5, 6, 7], illustrated in FIG. 5B. The gas to be treated (the letter Q indicates generically a hydrogen isotope and hence also tritium) is sent on a catalytic bed set—in this case—in the shell of the reactor RM, whilst sent in countercurrent into the lumen of the membrane is pure hydrogen. The membrane reactor RM performs through the membrane itself (selectively permeable to just the hydrogen isotopes) the isotope-exchange that carries out the required process. Described by way of example are two possible isotope-exchange reactions corresponding to detritiation of methane and water:2H2+CQ4CH4+2Q2 (1)H2+Q2OH2O+Q2 (2) It may be readily understood that in the reactions (1) and (2) the tritium atoms contained respectively in the methane and in the water are exchanged with H2 (protium, i.e., hydrogen of atomic mass 1). The innovative content of the membrane reactor RM used in the present invention is represented by the use of a particular device applied to the closed end of the permeator tube. Said device consists of a bimetallic spring, which has two separate functions: applying a tensile force on the permeator tube so as to prevent contact between the tube itself and the internal wall of the module, and hence prevent deformations of the membrane linked to the thermal cycles and hydrogenation cycles; and ensuring a thermal continuity between the closed end of the permeator tube and the outside of the membrane module, and hence enabling heating of the tube by the Joule effect. A solution proposed is that of providing said bimetallic spring using: a wire made of Inconel®, able to guarantee, even at the operating temperatures, the required mechanical performance (i.e., capacity of applying on the permeator tube a tensile force sufficient to keep the tube itself in a linear position, even during its expansion); and a wire made of silver with a low resistance to ensure passage of electric current and prevent heating of the spring itself. Downstream Utilities The main task of the downstream section is: monitoring of the level of contamination of the current of retentate gas leaving the membrane reactor (whether this be current leaving the lumen in the case of FIG. 3 or current coming from the shell of the reactor in the case of FIG. 6) so as to ensure release in conditions of safety into the environment (by means of a flue), and supply of the value of negative pressure necessary for circulation of the gas; storage of the extracted isotopes (current of swamping hydrogen enriched with tritium released from the contaminated gases), and supply also in this case of the value of negative pressure necessary for circulation of the gas. At this point, the isotopes extracted during the process can be valorized. The negative pressure for the two circuits (decontaminated-gas line and tritiated-hydrogen line) is obtained by two vacuum pumps, which are connected to the circuit via appropriate regulating valves. The hydrogen pressure on the permeate side is approximately 900 mbar, whilst the hydrogen pressure on the retentate side is approximately 100 mbar. The pressure sensors and the thermocouples also form part of the utilities and are used for regulation of the process parameters. All the information is recorded through a data-acquisition system. For what has been said so far, the process described herein basically comprises the following steps: A) shredding and mixing uniformly the waste to be detritiated; B) placing said material to be treated in a detritiation reactor RT; C) sending the inert gas and the demineralized water to an evaporation/mixing device; D) feeding said moist gaseous mixture, constituted by inert gas and vapour, to said reactor RT so that said mixture traverses all the material to be detritiated, giving rise to formation of a moist gaseous current containing tritium; E) sending said gaseous current containing tritium to a purposely provided catalytic membrane reactor RM; and F) feeding said membrane reactor RM with a swamping gas such as, for example, pure hydrogen, thus obtaining that from the reactor RM itself there come out, as end products, a gaseous current of hydrogen isotopes containing tritium extracted from the treated waste and a gaseous current of detritiated gases. The process forming the subject of the present patent has been specifically designed for decontaminating (detritiating) waste of the so-called “soft housekeeping” type (for example gloves, paper, etc.) coming from JET laboratories. Said waste is first treated within a detritiation reactor designed and developed by CEA. The process proposed, in addition to decontamination of the waste, also enables recovery and valorization of the tritium extracted (1 gT˜30 000 ). These Research and Development activities were conducted in the framework of the Tasks JW9-FT-2.34 (Preliminary Design and Tests for the detritiation of JET Soft House Keeping Waste) and JW10-FT-2.35 (Implementation of a Pd-membrane reactor into a detritiation facility treating JET Soft House Keeping Waste) that refer to the Research and Development programme “EFDA JET Fusion Technology Workprogramme”. The purpose of this activity is to design and assemble a pilot plant to test this new detritiation process. More in general, this device can be applied to the treatment of material coming from the machines tokamak (for example, JET, ITER and DEMO) or else to all the structures in which hydrogen isotopes such as H (protium), D (deuterium), and T (tritium) are used. According to the application it may be necessary to change one of the components (type of detritiation reactor, dimensional ratio, materials used, position of inlet and outlet, type of valves, type of pump), or else the dimensions of the objects described (length, diameter, type and volume of the catalyst, etc.). According to the amount of waste to be treated or the degree of decontamination that is to be obtained it is likewise possible to use membrane devices that contain a number of permeator tubes or provide connections in series or in parallel of the membrane modules. It is known that a person skilled in the sector will be able, by modifying the type of detritiation device, the type of membrane, etc., to design a similar process having the same functions. [1]. EFDA, European Fusion Development Agreement. ERB 5035 CT 99 0001. ANNEX VI, Information & Intellectual Property. [2]. Rist-Lambert, A., Detritiation of soft housekeeping materials. CEA Internal report, DTN/STPA/LPC/2005/024. [3]. Liger, K., Detritiation process for JET waste. CEA Internal report, DTN/STPA/LPC/2007/013. [4]. P. Giroux, D. O., J C. Durand, FR2620262A1, Patent for solid organic waste treatment. [5]. M. Glugla, A. Perevezentsev, D. Niyongabo, R. D. Penzhorn, A. Bell, P. Hermann, A PERMCAT Reactor for Impurity Processing in the JET Active Gas Handling System, Fusion Engineering and Design 49-50 (2000) 817-823 [6]. B. Bornschein, M. Glugla, K. Gunther, R. Lasser, T. L. Le, K. H. Simon, S. Welte, Tritium tests with a technical Permcat for final clean-up of ITER exhaust gases, Fusion Engineering and Design 69 (2003) 51-56 [7]. S. Tosti, L. Bettinali, F. Marini, Dispositivo per la rimozione di trizio da correnti gassose, Italian Patent n. RM2005U000165 (14 Dec. 2005).
	summary
	description	1. Field of the Invention Embodiments of the invention relate to the field of semiconductor device fabrication. More particularly, the present invention relates to a system for controlling ribbon beam uniformity in an ion implanter by adjusting the beam current profile along the beam path. 2. Discussion of Related Art Ion implantation is a process used to dope impurity ions into a semiconductor substrate to obtain desired device characteristics. An ion-beam is directed from an ion source chamber toward a substrate. The depth of implantation into the substrate is based on the ion implant energy and the mass of the ions generated in the source chamber. One or more ion species may be implanted at different energy and dose levels to obtain desired device structures. FIG. 1 is a block diagram of an ion implanter 100 including an ion source chamber 102. A power supply 101 supplies the required energy to source 102 which is configured to generate ions of a particular species. The generated ions are extracted from the source through a series of electrodes 104 and formed into a beam 10 which passes through a mass analyzer magnet 106. The mass analyzer is configured with a particular magnetic field such that only the ions with a desired mass-to-charge ratio are able to travel through the analyzer for maximum transmission through the mass resolving slit 107. Ions of the desired species pass from mass slit 107 through a deceleration stage 108 comprising multiple electrodes with defined apertures that allow the ion beam to pass through. By applying different combinations of voltage potentials to the multiple electrodes, the deceleration stage 108 manipulates the ion energies. A corrector magnet 110 shapes or collimates the ion beam generated from the deceleration stage 108 into the correct form for deposition onto a wafer or substrate. In particular, the corrector magnet receives a divergent ion beam and collimates the beam. Corrector magnet 110 is energized to deflect ion beamlets in accordance with the strength and direction of the applied magnetic field to provide a ribbon beam targeted toward a work piece or substrate positioned on support (e.g. platen) 114. In addition, the corrector magnet 110 filters out any ions from the beam that may have been neutralized while traveling through the beamline. In some embodiments, a second deceleration stage 112 may be disposed between corrector magnet 110 and support 114. The ions lose energy when they collide with electrons and nuclei in the substrate and come to rest at a desired depth within the substrate based on the acceleration energy. Generally, beam current, energy contamination and beam uniformity (current density and angle of implantation) are the parameters that jeopardize device throughput during semiconductor manufacturing processes. For example, if the beam current is too low, this will reduce the throughput of the implanter for a given total ion dose. Energy contamination occurs when there is a small fraction of the ion beam that is at a higher energy than desired which rapidly increases the depth of the desired junction that is formed in the substrate when creating an integrated circuit. This leads to degraded performance of the desired circuit profile. Beam uniformity is determined, in part, by the profile of the ion beam used for ion implantation. The beam profile is a map of ion beam intensity in a plane orthogonal to the direction of beam transport. The beam current may vary over the cross-sectional area of the ion beam, particularly in the case of large area beams such as ribbon ion beams. Furthermore, the beam profile may vary with implant conditions, such as dopant species, energy and current, and with time. These variations can compromise the desired device characteristics which produce lower manufacturing yields and lead to higher processing costs. Accordingly, it is desirable to measure and, if necessary, adjust the beam profile in order to enhance ion implanter performance before the ribbon beam is incident on the surface of a wafer or substrate. Exemplary embodiments of the present invention are directed to a system for adjusting ion ribbon beam profiles in an ion implanter. In an exemplary embodiment, a beam blocking array is positioned across a path of an ion ribbon beam in an ion implanter. The ion ribbon beam having a corresponding beam profile. The beam blocking array including a drive motor, an axle connected to the drive motor and a plurality of profile wheels disposed along the axle where each of said profile wheels is configured to rotate when the axle rotates. In addition, the array is disposed across a width of the ribbon beam such that each of the profile wheels has a position corresponding to a location across the ion beam. The present invention will now be described more fully hereinafter with reference to the accompanying drawings, in which preferred embodiments of the invention are shown. This invention, however, may be embodied in many different forms and should not be construed as limited to the embodiments set forth herein. Rather, these embodiments are provided so that this disclosure will be thorough and complete, and will fully convey the scope of the invention to those skilled in the art. In the drawings, like numbers refer to like elements throughout. FIG. 2A is a top block diagram view of a portion of an ion implanter 100 including an ion source chamber 102, extraction electrode assembly 120 and beam blocking array 130. Different feed gases are supplied to the source chamber 102 to generate ions having particular dopant characteristics. The ions are extracted from source chamber 102 via a standard three (3) electrode, extraction assembly 120 comprising a plasma or arc slit electrode 125, suppression electrode 126 downstream from plasma electrode 125 and ground electrode 127 downstream from suppression electrode 126. These electrodes are used to create a desired electric field to focus ion beam 10 extracted from source chamber 102. Plasma electrode 125 may be biased at the same large potential as the source chamber 102. Plasma electrode 125 includes a slot through which ions extracted from ion source chamber 102 pass. The slot has a width in the X direction that is significantly greater than its height in the Y direction to provide a high aspect ratio to form a ribbon ion beam traveling in the Z direction. Similarly, suppression electrode 126 includes a slot which is aligned with the slot from the plasma electrode which also has a width (X direction) significantly greater than its height (Y direction). Suppression electrode 126 is connected to a power supply and is typically biased at a moderate negative value to prevent electrons from entering back into source chamber 102 and to assist in focusing ion beam 10. Ground electrode 127 is positioned downstream from suppression electrode 126 and is maintained at ground potential. Ground electrode 127 also includes a slot aligned with the slot of the plasma electrode 125 and the slot of the suppression electrode 126 which also has a width significantly greater than its height. The strength of the electric field generated by the electrodes can be tuned to a desired beam current to extract a particular type of ion ribbon beam 10 from ion source chamber 102. Beam blocking array 130 may be disposed downstream of extraction electrode assembly 120 as illustrated in FIG. 2A or may be disposed upstream from corrector or collimator magnet 110 of ion implanter 100 as illustrated in FIG. 2B, or may be located just upstream from platen 114. Beam blocking array 130 is positioned across the width of beam 10 in the X direction and comprises a plurality of profile wheels 1311 . . . 131N. Each profile wheel 1311 . . . 131N is a non-circular, shaped element mounted on a centrally disposed stationary axle 132. Blocking array 130 may be fabricated from graphite or similar suitable conducting material. Generally, beam blocking array is configured to mechanically block, to varying degrees, selected portions of ribbon beam 10 extracted from ion source chamber 102 across the width of the beam to provide a desired uniform beam profile as illustrated in the exemplary profile of beam current vs. position included in FIG. 2A. Prior attempts to mechanically block portions of a ribbon beam employed a plurality of drive systems to individually position each of the profile wheels. However, this arrangement required a full set of drive motors for each profile wheel and associated complex sealing arrangements between the air-vacuum interface from the drive mechanics to the vacuum environment of the implanter and the ribbon beam. This resulted in vacuum leaks and consequently poor reliability. Alternatively, vacuum-compatible drive motors such as, for example, piezo-electric actuators may be employed which only require electric connections across the air-vacuum interface. However, even with such piezo-electric actuators there remains a requirement to have as many drive systems as beam profile wheels which increases cost and complexity which may compromise reliability. As described below, the beam blocking array 130 of the present disclosure utilizes a single drive assembly and position feedback system to arrange the plurality of profile wheels 1311 . . . 131N to mechanically block portions of ribbon beam 10 to varying degrees across the width of the beam. If the blocking array 130 is disposed upstream of the collimator magnet 110 as shown in FIG. 2B, the array acts upon the full width of the beam 10 and the number and width of each of the profile wheels will be greater than the number and width of the profile wheels of array 130 disposed downstream of extraction electrode assembly 120 shown in FIG. 2A. For example, array 130 shown in FIG. 2A with respect to the extraction electrode assembly 120 may be approximately 90 mm wide comprising twelve (12) separate profile wheels each having a width of about 7.5 mm. Alternatively, array 130 shown in FIG. 2B with respect to collimator magnet 110 may be approximately 500 mm wide comprising twenty (20) separate profile wheels each having a width of about 25 mm. Of course, these values are only provided as an example and may vary depending on the base uniformity profile to be corrected. In addition, the number of profile wheels or elements utilized will correspond to the minimum length scale in the desired profile and the degree of uniformity desired. FIG. 3 is a perspective view of beam blocking array 130 having a plurality of profile wheels 1311 . . . 131N rotatably disposed about axis 132. Each profile wheel has an elliptical shape and is ideally in contact with an adjacent profile wheel. The profile wheels may be prevented from lateral movement in direction X by the use of washers and/or C clamps. Blocking array 130 is located below (in direction Y) ribbon beam 10 such that the upper portions 131A of each of the profile wheels extends into the lower part of the beam 10 traveling above array 130 in direction Z. Alternatively, one end of each profile wheel for example, 131A can be used for blocking beams of p type dopants and the opposite end of each profile wheel can be used to block beams of n type dopants. Each profile wheel 1311 . . . 131N has a major axis M along the longitudinal length of the profile wheel and a minor axis N along the width of the profile wheel transverse to the major axis M. As each profile wheel 1311 . . . 131N rotates about axis 132, the amount or degree that the respective element blocks a portion of the ribbon beam 10 varies depending on what portion of the wheel along its respective major axis M blocks the beam. For example, as profile wheel 1311 is rotated such that its major axis M is parallel to the beam line of beam 10 traveling in direction Z, then the portion 131A of the profile wheel 1311 is below the beam and will not block the corresponding beam location. If profile wheel 1311 is rotated such that its major axis M is 90° to the beam line of beam 10, then the portion 131A of the profile wheel 1311 will block a portion of the beam corresponding to the position of profile wheel 1311 across the width of the beam 10. FIG. 4 illustrates a beam blocking array assembly 200 in accordance with the present disclosure including beam blocking array 130, drive wheel 210 disposed within housing 215 and drive motor 220 disposed outside of housing 215. Ball race 240 is disposed between drive motor 220 and housing 215. Axle 132 is supported by drive wheel 210 and to the housing 215 at 133 about which each of the profile wheels rotate. In other words, each of the profile wheels 1311 . . . 131N rotate about a stationary axle 132. The drive motor 220 rotates drive wheel 210 via drive shaft 221 and is located on the atmosphere side of the vacuum environment of the implanter along with encoder 240. The rotary motion from the drive motor 220 on the atmosphere side is transferred to the drive wheel 210 via drive shaft 221 on the vacuum side via a ferrofluid seal 235. When the drive wheel 210 rotates, a drive pin 250 engages profile wheel 1311 upon rotation. In particular, when the drive wheel 210 is rotated in one direction (either clockwise or counterclockwise), the drive pin 250 engages a first profile wheel 1311. Each of the profile wheels 1311 . . . 131N also includes a drive pin 258 which protrudes from the respective profile wheels parallel to axle 132 in direction X away from drive motor 220 as well as a pick-up pin 268 which extends parallel to axle 132 in direction X toward drive motor 220. This can be more easily seen turning briefly to FIG. 5 which is a perspective view of an exemplary profile wheel 131N having a pair of opposed sides 280 and 281. The profile wheel has an elliptical shape (although other configurations may be utilized) with major M and minor N axes. Axle 132 is centrally disposed through profile wheel 131N. Drive pin 258 extends away from drive motor 220 from surface 281 and pick-up pin 268 extends toward drive motor 220 from surface 280. The pick-up pin 268 and drive pin 258 are spaced the same distance from axle 132 on respective surfaces 280 and 281. An additional pick-up pin 268A shown in shadow may also be used which extends toward drive motor 220 from surface 280. If the dual pick-up pin configuration is employed, the pins are separated by at least 120°. Each of the pick-up pin 268 (268A) and drive pin 258 are located away from the major M and minor N axes by more than an angle subtended at the axle 132 by the width of the respective drive and/or pick-up pin to allow for consecutive locations along the profile of beam 10 to have the same blocking position. In other words, this allows two adjacent profile wheels to be in the same position relative to the beam. As the first profile wheel 1311 is rotated, its associated drive pin 258 engages adjacent profile wheel 1312 forcing it to rotate about axle 132. The drive pin associated with profile wheel 1312 engages adjacent profile wheel 1313 forcing it to rotate. This process continues until all of the profile wheels are engaged by rotation of the drive wheel 210. Returning to FIG. 4, once the last profile wheel 131N is in the desired position with respect to blocking a portion of ribbon beam 10 corresponding to the location of the profile wheel 131N as confirmed by the profiler described below with reference to FIG. 7, the drive wheel 210 stops and rotates in the opposite direction where pick-up pin 168 engages an adjacent profile wheel. For example, drive wheel 210 is rotated clockwise until each of the profile wheels 1311 . . . 131N are engaged by the respective drive pins 258 and likewise forced to rotate in a clockwise direction. When profile wheel 131N which is farthest from drive motor 220 is in the desired position with respect to its location along ribbon beam 10, drive wheel 210 stops rotating in the clockwise direction. Drive wheel 210 then rotates in the opposite, in this case counterclockwise direction, again engaging each profile wheel 131N-1 . . . 1311 via respective pick up pins 268 of each profile wheel except for profile wheel 131N which remains in the desired position. This counterclockwise rotation continues until each of the remaining profile wheels 1311 . . . 131N-1 is in the desired position to block a portion of the ribbon beam 10 to obtain the desired beam profile uniformity. Rotation of the profile wheels may be done at relatively high speeds. The average number of rotations (clockwise and counterclockwise) of drive wheel 210 needed to engage and set each of the profile wheels 1311 . . . 131N corresponds to ½ N(N+2) where N is the number of profile wheels along array 130. Thus, for an array having N=10 profile wheels, the number of turns of drive wheel 210 is approximately 60 (½(10)(10+2)). Rotation of the profile wheels can generate unwanted heat within the vacuum environment and can be cooled through axle 132. In one embodiment, axle 132 may be defined by a pair or concentric tubes having an inner portion and an outer portion. A coolant (e.g. water) may be pumped into the outer portion of the tube and returned via the inner portion of the tube. Once the profile wheels 1311 . . . 131N are in the desired position with respect to a location along beam 10, drive wheel 210 stops and the profile wheels 1311 . . . 131N do not rotate freely about axis 132. Each of the profile wheels 1311 . . . 131N is affixed to axle 132 by adjusting the slip torque. Although the slip torque of each of the profile wheels 1311 . . . 131N is sufficient to prevent the wheels from rotating once a respective position has been obtained, the slip torque must be low enough to allow drive motor 220 to rotate all of the profile wheels 1311 . . . 131N simultaneously. For example, a slip torque of approximately a few oz-inches is adequate for the beam blocking array shown in FIGS. 2A and 2B. FIG. 6 is a perspective view of an alternative configuration of any one of the profile wheels 1311 . . . 131N shown in FIG. 5 having a collar portion 290 and recessed opposed first and second internal wall surfaces 211. The interior wall surfaces 211 are recessed from the respective perimeters 291, 291′ a distance “d” corresponding to collar portion 290. Drive pin 250 extends transversely away from drive motor 220 towards an adjacent profile wheel 131N to engage the adjacent profile wheel. However, the profile wheels 1311 . . . 131N shown in FIG. 5 require spacing therebetween along axle 132 since the opposed surfaces 280 and 281 are not recessed from the respective perimeters a distance “d” as shown in FIG. 6. This spacing is required between the profile wheels to allow the pins 258, 268 to rotate and engage an adjacent profile wheel. Thus, an array 130 having profile wheels of FIG. 5 would be configured for placement upstream from a target wafer disposed on platen 114 where the angular spread of the beam 10 means that the blocking profiles at the target wafer generated by the upstream profile wheels will overlap. Thus, the gaps between the profile wheels caused by the pins 258, 268 would not affect the profile. However, if an array 130 including a plurality of profile wheels having the configuration shown in FIG. 5 was located close to the target wafer, the blocking profiles would not overlap. The profile wheel configuration of FIG. 6 avoids this by having the respective drive pin 258 and pickup pin 268 recessed a distance “d” within the collar 290. In this manner, adjacent profile wheels 1311 . . . 131N may be positioned more closely since the respective drive pins 258 and pick-up pins 268 are recessed at least partially within collar portions 290 the distance “d” allowing the profile wheels to rotate. Thus, the profile wheel of FIG. 6 is used when the array 130 is disposed downstream along the beam line closer to the target wafer where unwanted spacing between the profile wheels may compromise the desired uniformity profile. FIG. 7 illustrates an operational flow chart of an exemplary method of configuring the uniformity of an ion beam incident on a target wafer or substrate utilizing the beam blocking array 130 in an ion implanter in accordance with this disclosure. A beam profiler is used to measure the profile of an ion beam 10 at step 300. Typically, a beam profiler measures beam current in incremental areas over the cross-sectional area of the ion beam 10 to obtain a beam profile in the form of a two-dimensional map of beam current density. The two-dimensional map may be configured as an X-Y array of pixels, each of which contains a measured value of beam current density in an incremental area of the beam cross section. The two-dimensional map of beam current density is utilized to confirm that the ion beam profile is within specification. If the beam profile is not within specification, the beam blocking array 130 is used to block particular locations of the beam 10 to obtained the desired profile. In particular, the beam profile measured by the profiler at step 300 is supplied to a controller at step 310. The controller compares the measured beam profile with the position of the drive wheel obtained from the encoder 240 at step 320. Based on this comparison, the controller calculates the current density of the profile to be reduced across the ion beam 10. Drive motor 220 is engaged at step 330 which rotates drive wheel 210 at step 340. Drive wheel 210 in turn rotates the profile wheels 1311 . . . 131N such that a portion of the ion beam is blocked at a location corresponding to a respective one of the profile wheels 1311 . . . 131N at step 350. In particular, drive wheel 210 rotates profile wheels 1311 . . . 131N in a first direction until the last profile wheel 131N which is the furthest away from the drive motor 220 is in the desired location with respect to the ion beam. The remaining profile wheels 1311 . . . 131N-1 are rotated in a second direction opposite the first direction until all the profile wheels have been set corresponding to a respective location of the ion beam to block that portion of the beam. The amount or degree to which each profile wheel blocks the ion beam changes as the wheels rotate about axle 132. Again, the beam profiler measures the beam profile at step 300 and the process continues until the desired beam profile is obtained. In this manner, a desired profile of an ion ribbon beam is obtained by positioning a beam blocking array along the path of an beam and mechanically blocking portions of the beam to varying degrees across the width of the beam. While the present invention has been disclosed with reference to certain embodiments, numerous modifications, alterations and changes to the described embodiments are possible without departing from the sphere and scope of the present invention, as defined in the appended claims. Accordingly, it is intended that the present invention not be limited to the described embodiments, but that it has the full scope defined by the language of the following claims, and equivalents thereof.
	abstract	An illumination optical system illuminates a surface to be illuminated. The illumination optical system includes a mirror having a surface effective to shape X-rays from a source into X-rays having an arcuate sectional shape, and an optical system for illuminating the surface to be illuminated, with the X-rays having an arcuate sectional shape from the mirror and in an oblique direction with respect to that surface.
047217383	abstract	A composition of matter including a polymer sensitized to the heating effects of microwave energy which comprises said polymer and a microwave sensitizing amount of a particulate, layered compound comprising a tetravalent atom and a pentavalent atom, selected from the group consisting of elements of Group V of the Periodic Table of the Elements, and having an atomic weight of greater than 30, dispersed throughout said polymer.
052215156	summary	BACKGROUND OF THE INVENTION The invention relates to a method for manufacturing grids to be incorporated in nuclear reactor fuel assemblies. It is particularly suitable for the manufacture of grids for assemblies whose framework comprises end pieces connected by guide tubes imprisoned in some of the cells of grids spaced apart evenly along the guide tubes, the other cells of the grids supporting the fuel rods of the assembly. Among such grids, reference will particularly be made to those whose major faces have a square shape and which comprise two sets of inner plates disposed in two orthogonal directions and interlocked together so as to define the elementary cells through which the fuel rods and the guide tubes pass. The inner plates are connected at their peripheries, to plates forming a belt. These component parts are secured together at their intersections to provide cohesion of the grids. Grids are also known, particularly for undermoderated reactor assemblies, which have a hexagonal cross-section and which comprise three sets of intersecting plates. Although particularly advantageous in the case of grids whose major faces have a square shape and whose cells are distributed above the nodes of the square lattice, the invention also applies to grids whose major faces have a hexagonal shape. The component parts of the grids have generally been secured together by manual brazing, which is a long and tedious operation and which, in addition, does not guarantee fully satisfactory reproducibility. Different methods and devices have however been proposed for welding the component parts together in a more or less automatic way. The invention relates to a method of the kind described in French Patent No. 2,522,560, in which the grid whose component parts are to be secured together is placed in a mechanical shaping and holding frame having passages for access to the points to be welded on the two major faces and on the sides of the grid; the frame containing the grid is gripped by a device, placed in a gas-tight chamber containing an inert gas atmosphere; weldings is carried out on one face at a time, using a laser beam orthogonal to the axis of rotation of the device, delivered by a laser generator placed outside the chamber and penetrating thereinto through a transparent window. The operation is then repeated on the other faces after manually repositioning the grid. By welding in an inert gas protective atmosphere, oxidation of the welds is avoided, which makes it possible, in particular, to apply the method not only to grids made of a high strength alloy, such as INCONEL, but also to grids whose components are made of a zirconium-based alloy. From a reading of French Patent No. 2,522,560, it seems that the protective atmosphere is obtained by scavenging, which involves considerable inert gas losses. The laser generator is held fixed during welding of a face, except for the welds carried out on the belt, for which the beam is deflected by means of a set of mirrors. The variety of accurate movements of the grids inside the chamber does not guarantee sufficient accuracy to be compatible with the very small size of the focus spot of the laser beam. In addition, the chamber must have a very large volume. FR-A-2532216 and EP-A-0102252 also describe a method for welding using a fixed laser generator. The laser beam is directed and focussed in an open chamber, subjected to argon scavenging, in which a grid is driven with orthogonal displacement movements. This method requires very considerable gas flows, collection of the gases polluted by the laser shots is difficult and controlling leaks in the open chamber requires very reduced clearances, which are difficult to obtain. U.S. Pat. No. 4,710,606 describes a welding robot in which a laser beam must be moved along five degrees of freedom by mirrors, which implies a great complexity of the mechanism and accumulation of tolerances. SUMMARY OF THE INVENTION An object of the invention is to provide a method for welding grids by laser beam of the type defined above, for sequentially and automatically making the required welds under conditions ensuring mechanical holding of the parts and absence of deformations and oxidation, while avoiding complex movements of a kind likely to adversely affect the welding accuracy due to cumulative lost motions. To this end, there is provided a method of the above-defined type, wherein, for making the welds on one face of the grid, the laser beam is moved in two directions orthogonal to each other and to the direction of the beam, while the frame is held stationary in the inert gas atmosphere chamber. It is consequently possible to distribute the movements between the laser beam, the required focussing optical system and the frame. Although this method is of particular advantage in the case of grids whose plates and belts are made of zirconium-based alloy, it can also be used in the case of grids whose component parts are made of a nickel-chrome base alloy or of steel. In a typical embodiment of the invention, each grid is placed first of all in a first chamber having a device for orientation of the grid about a first axis, and all welds are effected on the faces parallel to this axis; then the grid still contained in its frame, is placed in another chamber having a device for orientation about an axis which is parallel to the first one, but on which the grid is placed in an angular position perpendicular to that which it had in the first device, and welding is carried out on the major faces. This arrangement considerably simplifies the orientation devices, and thus provides greater positioning accuracy. On each face, welding is carried out in the same plane, i.e. without modification of the distance to the laser generator, except for particular welds, such as the corner welds of the grid. Since the grid remains in its frame from the beginning to the end of the welding operations, transfer from one chamber to the other does not disturb the relative position of the component parts of the grid. Unloading of one grid, loading of another grid to be welded and provision of inert atmosphere in a chamber may take place during welding of the grid occupying the other chamber, i.e. in overlapping time. In practice, the inert gas atmosphere may be created by pumping out the chamber until a primary vacuum is obtained, then filling with inert gas so that the atmosphere in which welding takes place contains less than 50 vpm of oxygen. To protect the windows from the alternating pressure stresses, they may be protected by lids during evacuation. It is often of advantage to form some at least of the welds by shots focussed in two points located on two parallel lines at a short distance from each other. To that end, the optical path of the laser beam may be defined by optical means, such as mirrors or lenses, which are vibrated. Then the vibrations of the focussing lens are synchronized with the laser pulses, so that each shot takes place when the optical path is oriented towards the point which is to receive the energy. The invention also provides an installation for laser welding the different component parts of the grid, comprising a controlled atmosphere enclosure having a device for angular adjustment about an axis, receiving a frame containing a grid to be welded and having a transparent window and comprising a welding laser source carried by a table with crossed movements in two orthogonal directions one of which is parallel to the axis of angular adjustment. BRIEF DESCRIPTION OF THE DRAWINGS The invention will be better understood from the following description of a particular embodiment, given by way of example.
	description	Any and all applications for which a foreign or domestic priority claim is identified in the Application Data Sheet as filed with the present application are hereby incorporated by reference under 37 CFR 1.57. This application claims priority to and the benefit of U.S. provisional application 61/753,851, entitled HETEROGENEOUS CORE DESIGNS AND THORIUM BASED FUELS FOR HEAVY WATER REACTORS, and filed Jan. 17, 2013, the entire contents of which are hereby incorporated by reference. Field The invention relates to core designs for thorium based fuels for heavy water reactors and more specifically to heterogeneous core designs for thorium based seed fuel and blanket fuel for channel-type heavy water reactors as well as thorium based fuel bundles for a heterogeneous core design. Description of the Related Art Research into the use of thorium as a new primary energy source has recently been explored. Thorium-232 (Th-232) is a naturally occurring isotope and is substantially more abundant than uranium. Although not fissile, upon absorbing a neutron will transmute to uranium-233 (U-233), which is an excellent fissile fuel material. Thorium fuel concepts therefore require that Th-232 is first irradiated in a reactor to provide the necessary neutron dosing. The U-233 that is produced can either be chemically separated from the parent thorium fuel and recycled into new fuel, or the U-233 may be usable in-situ in the same fuel form. Thorium fuels therefore require a fissile material as a driver so that a chain reaction (and thus supply of surplus neutrons) may be maintained. Fissile driver options are U-233, U-235 or Pu-239. It is possible, although difficult, to design thorium fuels that produce more U-233 in thermal reactors than the fissile material they consume (this is referred to as having a fissile conversion ratio of more than 1.0 and is also called breeding). Thermal breeding with thorium is possible using U-233 as the fissile driver, and to achieve this the neutron economy in the reactor has to be very good (i.e., low neutron loss through escape or parasitic absorption). The possibility to breed fissile material in slow neutron systems is a unique feature for thorium-based fuels. Another distinct option for using thorium is as a ‘fertile matrix’ for fuels containing transuranic elements such as plutonium. No new plutonium is produced from the thorium component, unlike for uranium fuels, and so the level of net consumption of this metal is rather high. In fresh thorium fuel, all of the fissions (thus power and neutrons) derive from the driver component. As the fuel operates the U-233 content gradually increases and it contributes more and more to the power output of the fuel. The ultimate energy output from U-233, and hence indirectly thorium, depends on numerous fuel design parameters, including: fuel burnup attained, fuel arrangement, neutron energy spectrum and neutron flux. The fission of a U-233 nucleus releases about the same amount of energy (200 MeV) as that of U-235. An important principle in the design of thorium fuel is that of fuel arrangements in which a high fissile (and therefore higher power) fuel zone referred to as the seed region is physically separated from the fertile (low or zero power) thorium part of the fuel referred to as the blanket region. Such an arrangement is far better for supplying surplus neutrons to thorium nuclei so they can convert to fissile U-233. Previous heavy water reactor core designs and associated fuel for channel-type heavy water reactors using thorium-based fuels have not been able to achieve simultaneously high fuel burnup, high fissile utilization and high conversion ratios, while also meeting design goals of high core-average power densities, meeting goals of operating limits on bundle power and maximum linear element ratings while keeping reactivity coefficients, such as for example coolant void reactivity, within desired values to enhance safety characteristics. Previous research in heavy water reactors have tended to focus on the design of homogeneous cores and heterogeneous fuel bundle designs that use neutron absorbing poisons to reduce void reactivity and has neglected to consider alternative design options. A thorium fuel based core design and/or a fuel bundle design that mitigates one or more various shortcomings is therefore in need. Thorium is an attractive fuel option to improve the sustainability of the nuclear fuel cycle, given the limited and unevenly distributed uranium reserves. As natural thorium does not contain a fissile isotope, implementation of thorium fuels in a reactor must involve a fissile component, generally either plutonium or uranium. The physical separation of a lower fissile blanket fuel and a higher fissile seed fuel into separate adjacent regions in a heterogeneous reactor core allows for the potential to improve the fissile utilization and increase the sustainability of the thorium fuel cycle. In one embodiment of the invention, there is provided a channel type heterogeneous reactor core for a heavy water reactor for burnup of thorium based fuel, the heterogeneous reactor core comprising at least one seed fuel channel region comprising seed fuel channels for receiving seed fuel bundles of thorium based fuel; and at least one blanket fuel channel region comprising blanket fuel channels for receiving blanket fuel bundles of thorium based fuel; wherein the seed fuel bundles have a higher percentage content of fissile fuel than the blanket fuel bundles. In an additional embodiment to that outlined above, the at least one seed fuel channel region and the at least one blanket fuel channel region are set out in a checkerboard pattern within the heterogeneous reactor core. In an additional embodiment to that outlined above, the at least one seed fuel channel region and the at least one blanket fuel channel region are set out in an annular pattern within the heterogeneous reactor core. In an additional embodiment to that outlined above, the seed fuel bundle comprises 35% or more UO2 and 65% or less ThO2. In an additional embodiment to that outlined above, the seed fuel bundle comprises 3% or more PuO2 and 97% or less ThO2. In an additional embodiment to that outlined above, the blanket fuel bundle comprises 30% or less UO2 and 70% or more ThO2. In an additional embodiment to that outlined above, the blanket fuel bundle comprises 2% or less PuO2 and 98% or more ThO2. In another embodiment of the invention, there is provided a fuel bundle for use in a channel type heterogeneous reactor core of a heavy water reactor, the fuel bundle comprising a central displacement tube; and a plurality of thorium based fuel pins surrounding the central displacement tube. In an additional embodiment to that outlined above, the central displacement tube is filled with ZrO2, MgO, BeO, graphite or stagnant D2O coolant. In an additional embodiment to that outlined above, there are 21 radially positioned thorium based fuel pins surrounding the central displacement tube. In an additional embodiment to that outlined above, there are 35 radially positioned thorium based fuel pins surrounding the central displacement tube. In an additional embodiment to that outlined above, the fuel bundle is a seed fuel bundle and the plurality of thorium based fuel pins comprises a homogeneous mixture of (PuO2+ThO2) with a PuO2 content of 3% or higher. In an additional embodiment to that outlined above, the fuel bundle is a seed fuel bundle and the plurality of thorium based fuel pins comprises a homogeneous mixture of (UO2+ThO2) with a UO2 content of 35% or higher. In an additional embodiment to that outlined above, the fuel bundle is a blanket fuel bundle and the plurality of thorium based fuel pins comprises a homogeneous mixture of (PuO2+ThO2) with a PuO2 content of 2% or less. In an additional embodiment to that outlined above, the fuel bundle is a blanket fuel bundle and the plurality of thorium based fuel pins comprises a homogeneous mixture of (UO2+ThO2) with a UO2 content of 30% or less. In an additional embodiment, the present invention provides for the use of a fuel bundle such as those embodiments outlined above in channel type heterogeneous reactor core of a heavy water reactor for burnup of thorium based fuel. A heterogeneous reactor core for a channel type heavy water reactor is provided. A channel-type heavy water reactor, similar to what is being currently used in nuclear power generation may be used as the initial basis for the design. The heterogeneous core comprises a lattice of channels for receiving seed or blanket fuel bundles in the channels as will be discussed below with reference to FIGS. 2-7. The core may contain from 25% to 84% seed fuel channels while the balance are blanket fuel channels. The nuclear fuel is in the form of short, (˜50 cm) or longer (>50 cm) fuel bundles made generally with one or two rings of fuel pins. It has been determined that to help minimize coolant void reactivity while maximizing fuel burnup and fissile utilization, the fuel bundle is designed to have only one or two rings of fuel pins, with a central displacer tube filled with stagnant coolant, or a solid moderator, for example graphite, or material with a low neutron scattering and low neutron absorption cross section, for example ZrO2 or MgO. The fuel bundles will be discussed in more detail below with reference to FIGS. 1A-1E. The nuclear fuel bundles are made from thorium, mixed with either plutonium or uranium, generally in oxide, carbine, silicide or a metallic-alloy form. As depicted in various non-limiting embodiments in FIGS. 2-7, an embodiment of the reactor core of the is a heterogeneous design with physically separate regions of seed fuel channels and blanket fuel channels arranged in a lattice. In FIGS. 2-7, seed channels are represented by an S and blanket channels are represented by a B. Seed fuel is made with higher concentrations of fissile fuel mixed with thorium and is used primarily to generate power and excess neutrons to drive blanket fuel. The blanket fuel is made with lower concentrations of fissile fuel mixed with thorium and used primarily to convert fertile thorium fuel into fissile fuel. There is some power generation by the blanket fuel. A seed channel, in one embodiment, is for seed bundles only while a blanket channel is for blanket bundles only. In the embodiments shown, each channel of the heterogeneous core has 12 bundles (either seed or blanket). As shown, the core may contain from 25% to 84% seed fuel channels, while the balance are blanket fuel channels. The core may have a lattice in a checkerboard-type arrangement of seed and blanket fuel channels such as those shown in FIGS. 6 and 7. Alternatively, the core may have a lattice in an annular arrangement of seed and blanket fuel channel regions with the outermost ring of the fuel channels adjacent to the radial reflector (not shown) of the core filled with blanket fuel channels such as those shown in FIGS. 2, 3, 4 and 5. It will be appreciated that there are several different permutations of heterogeneous seed/blanket core layouts which may be used or implemented and those shown in FIGS. 2 to 7 are not intended to be limited but rather illustrative of various embodiments of the concept of heterogeneous cores of the invention. The heterogeneous core allows for different and dynamic refueling strategies as the blanket fuel regions and the seed fuel regions can be refueled at different rates to achieve desirable burnup levels and core power distributions. Refueling strategies will be discussed in more detail below. The reactor core may be similar to current reactor cores such as the CANDU-6/EC-6 reactor which has 380 fuel channels with a square lattice pitch of 28.575 cm. Each channel thereof contains 12 fuel bundles, each approximately 50 cm long. Current CANDU cores use a homogeneous core of natural uranium (NU). Some more advanced designs use a single type of fuel and are still considered homogeneous. Shown in FIGS. 1A-1E are embodiments of fuel bundles for use in the channels of the heterogeneous core. As can be seen in the Figures, the fuel bundles include a central displacer tube to replace the central 8 fuel pins in a 43-element bundle, leaving outer rings of 14 and 21 fuel pins (FIGS. 1B and 1C). A further design, shown in FIGS. 1D and 1E includes a larger central displacer tube to replace the central 22 pins in a 43-element bundle leaving an outer ring of 21 fuel pins. Without wishing to be limited, the central displacer tube may be filled with ZrO2, MgO, BeO, graphite or stagnant D2O coolant. The purpose of the central displacer tube is to reduce coolant void reactivity (CVR). An advantage of the central displacer tube is that it helps to reduce the CVR, improving the safety characteristics of the lattice and the reactor during a postulated accident scenario, where there is a loss of coolant. The fuel pins of either the 21-element bundle or the 35-element bundle may be a combination of plutonium and thorium or low enriched uranium and thorium depending on whether the bundle is for use in a seed fuel region or blanket fuel region. In FIG. 1A, the inner 8 fuel pins are all the same, namely ThO2 and the outer 35 fuel pins are all the same, namely a homogeneous mixture of (PuO2+ThO2) or (UO2+ThO2). In the fuel bundles shown in FIGS. 1B and 1C, the 35 fuel pins are all the same, namely a homogeneous mixture of (PuO2+ThO2) or (UO2+ThO2) wherein the fuel bundle of FIG. 1B has central Zr-4 displacement tube filled with stagnant D2O coolant and the fuel bundle of FIG. 1C has a central Zr-4 displacement tube filled with ZrO2. In the fuel bundles shown in FIGS. 1D and 1E, the 21 fuel pins are all the same, namely a homogeneous mixture of (PuO2+ThO2) or (UO2+ThO2) wherein the fuel bundle of FIG. 1D has central Zr-4 displacement tube filled with stagnant D2O coolant and the fuel bundle of FIG. 1E has a central Zr-4 displacement tube filled with ZrO2. In the embodiments of fuel bundles wherein PuO2 is mixed with ThO2, the Pu is “reactor grade” Pu. In embodiments of fuel bundles wherein UO2 is mixed with ThO2, then the U is LEU (low enriched uranium), with a fissile content of about 5 wt % U-235/U in one non-limiting embodiment. The volume fraction of PuO2 in (Pu+Th)O2 may range from 1% to 13% in various non-limiting embodiments. The volume fraction of UO2 in (U+Th)O2 may range from 5% to 70% in various non-limiting embodiments. It will be appreciated that the mixture (volume fractions of either PuO2 or UO2 in (Pu+Th)O2 or (U+Th)O2) is dependent on whether the fuel is “seed” or “blanket” fuel. Seed fuel has a higher volume fraction of PuO2 or UO2 than blanket fuel. Typically, seed fuel contains fuel with 3% or higher PuO2 in (Pu,Th)O2, or 35% or higher UO2 in (U,Th)O2. The choice of LEU (in the non-limiting embodiment shown, 5 wt % U-235/U) for mixing with thorium (Th) is generally based on practical and economic considerations. 5 wt % U-235/U is readily available from existing enrichment facilities throughout the world as is therefore more commonly used. The choice of reactor grade Pu (generally about 0.67 wt % fissile Pu (Pu-239+Pu-241)) for mixing with Th is generally based on the assumption that most of the Pu inventory available in the world today is found in the spent fuel from light water reactors (LWRs). It is conceivable that one might use Pu from other sources, such as spent CANDU reactor natural uranium fuel, or Magnox reactor natural uranium fuel, or plutonium obtained from nuclear weapons stockpiles, or from a fast breeder reactor. In these other potential sources of plutonium, the fissile content will be different, probably higher. In principle, the plutonium from these alternative sources may be used in the heterogeneous reactor design as well, but given the assumption that the fissile plutonium content is higher, then the volume fraction of PuO2 in (Pu,Th)O2 would likely be lower to achieve the same level of burnup. Generally, a typical seed fuel will contain 35% UO2 (or more) and 65% ThO2 (or less), or it will contain 3% PuO2 (or more) and 97% (or less) of ThO2. Whereas a typical blanket fuel will contain 30% UO2 (or less) and 70% ThO2 (or more), or it will contain 2% PuO2 (or less) and 98% (or more) of ThO2. The fraction of the core's fuel channels that are seed channels can range from about 25% to about 84%. In most designs, the fraction is approximately 50% seed fuel channels and 50% blanket fuel channels as shown for example in FIGS. 2, 4 and 7. The core layout shown in FIG. 5 includes approximately 84% seed channels (320 channels) and 16% blanket channels (60 channels). An advantage of using more seed channels is that one can generate more power and achieve higher burnup while maintaining core reactivity. In addition, by using more seed the reactor may be operated at a higher power level, with a higher core-average power density. Typically, most of the previous CANDU core designs involving thorium based fuels have assumed a homogeneous core with one fuel type. The refuelling rates (and the core-average burnup of the fuel) depend on the choice of the fuel used (its initial enrichment), the desired radial and axial power distribution in the core, and the refuelling scheme. One refuelling scheme is a simple two-bundle shift, with bi-directional fuelling in alternating channels. Bundles are inserted from one side of the reactor, and are progressively moved to the other side until they reach the desired burnup. The objective in adjusting the exit burnup in each channel (and hence the refuelling rate) is to ensure that the maximum bundle power stays below ˜750 kW, and that the maximum channel power stays below ˜6,500 kW. However, it is also ideal to make the radial and axial power distribution as flat as possible, in order to maximize the power generated in the core, for economic advantage. The initial core designs used 35-element Pu/Th seed fuel that would achieve an approximate discharge burnup of 20 MWd/kg to 40 MWd/kg burnup. In most of the cases studied that meant using (3 wt % PuO2/97 wt % ThO2) for the seed to achieve a burnup of ˜20 MWd/kg. For core-average burnups closer to 40 MWd/kg, this means using (4 wt % PuO2/96 wt % ThO2). Most of the blanket fuel was either (2 wt % PuO2/98 wt % ThO2), burned to ˜20 MWd/kg, or (1 wt % PuO2/99 wt % ThO2) burned to 40 MWd/kg. Heterogeneous cores with LEU/Th fuel have not been tested yet, but they would use the same methods that were used in the analysis of the cores with Pu/Th fuel. There are two additional refuelling strategies to further improve the performance of the heterogeneous seed/blanket core, although these have not yet been tested: 1) To carry out axial shuffling of the fuel bundles in a given channel to help flatten the axial power distribution. This could be particularly useful in cores using seed fuel with higher levels of fissile enrichment (such as 5 wt % PuO2/95 wt % ThO2) and higher burnups (greater than 40 MWd/kg). The use of axial shuffling has been considered in the past by AECL in studies of CANDU reactor cores using SEU fuels (1.2 to 3 wt % U-235/U). 2) To send high enrichment, high-burnup seed fuel through a core twice or three times, somewhat analogous to what is done with batch refuelling in light water reactors. This is what would be called a 2TT (2 times through thorium) or 3TT (3 times through thorium) fuel cycle. For example, a seed fuel bundle which is estimated to have enough reactivity (and initial fissile content) to achieve a large discharge burnup will go through the CANDU core in three passes in three different channels. In addition, for example, a 35-element bundle might be made of (5 wt % PuO2/95 wt % ThO2) and lattice physics calculations indicate that it could achieve a final burnup of ˜54 MWd/kg. Instead of pushing the burnup of the fuel bundle from 0 to 54 MWd/kg in a single pass through the core, it can be divided up into two or three passes through the core. If divided into 3 passes, then the fuel would be burned from 0 to 18 MWd/kg in the first pass in one channel, 18 to 36 MWd/kg in the 2nd pass in another channel, and finally 36 to 54 MWd/kg in the third pass through another channel. A smaller change in the burnup between the inlet and exit of a given fuel channel will help flatten the axial power distribution, and permit a higher core power density, while staying within limits of peak bundle power and peak channel power. This type of refuelling scheme combines the on-line, bi-directional, continuous refuelling features of a CANDU reactor with the multi-batch zone refuelling schemes of a light water reactor (such as a PWR). Shown in FIGS. 8 to 16 are Tables 1 to 9 which set out geometry specifications and material specifications of the different fuel designs. Table 1 in FIG. 8 shows an embodiment wherein the reactor grade plutonium contains ˜52 wt % Pu-239 and ˜15 wt % Pu-241, giving a total fissile content of ˜67 wt % Pu-fissile/Pu. Table 2 in FIG. 9 shows the isotopic composition of LEU in oxide form. Thus, the fissile content is ˜5 wt % U-235/U and the balance of uranium is U-238 and U-234. Table 3 in FIG. 10 shows a description of different lattices tested. There are 10 different lattice designs, which are differentiated by geometry (5 geometry types) and fuel type (two fuel types, either (U,Th)O2 or (Pu,Th)O2) in the outer 35 or 21 pins. Only bundle designs 1 and 6 have 8 central ThO2 pins. All other pins are a mixture of either (U,Th)O2 or (Pu,Th)O2. Table 4 in FIG. 11 shows the dimensions of components for various lattices tested. The dimensions are given for a fuel pellet made of (Pu,Th)O2 or (U,Th)O2, or ThO2, the radius of the clad for the fuel element, the inner and outer radius for the central displacer tube, the inner and outer radius for the pressure tube (PT), the inner and outer radius for the calandria tube (CT). Table 5 in FIG. 12 shows the number of fuel pins and the pitch circle and radius, and the angular offset for the first fuel pin in the bundle. Note: bundle design 1a is the only one that has 4 rings of fuel pins (1+7+14+21). Bundle designs 1b and 1c do not have a central pin or an inner ring of fuel pins, only two outer rings of fuel pins (14+21). Bundle designs 1d and 1e have only a single outer ring of 21 fuel pins. Table 6 in FIG. 13 shows the material specifications for key components for various lattices tested. The type of material, its nominal operating temperature, and its nominal material mass density are given. The nominal purity of the heavy water moderator and the heavy water coolant are also specified. However, it should be pointed out that the purity of the heavy water in both the moderator and the coolant could be increased. Table 7 in FIG. 14 shows the value of the mass fractions for Pu-fissile (Pu-239+Pu-241) Pu, Th, and O in (Pu,Th)O2 for various volume fractions of PuO2 in (Pu,Th)O2. The fuels containing low volume fractions of PuO2 (e.g., 2% or less) are considered blanket fuel, while the fuels containing higher volume fractions of PuO2 (e.g. 3% or higher) are considered seed fuel. Also shown below is a sample set of core calculation results for two cores (1S-1B, and 84% Seed/16% Blanket) with different combinations of Seed and Blanket fuels. The data for the 1S-1B core design is shown in Table 8/FIG. 15. This shows the various performance characteristics of 5 different core designs, which differ in the type of seed and blanket fuel used. The data for the 84%-Seed/16% blanket core design is shown in Table 9/FIG. 16. This shows the various performance characteristics of 4 different core designs, which differ in the type of seed and blanket fuel used. The above described heterogeneous reactor core and fuel bundles are intended to be illustrative of the invention and are not intended to be limiting in any way. It will be appreciated that modifications and alterations to the design, function or use of the heterogeneous reactor core and fuel bundles may be made which are within the sphere of the invention contemplated and are within the scope of the claims.
	description	This application claims priority based on U.S. Provisional Patent Application Ser. No. 61/940,571 filed Feb. 17, 2014. The direct conversion of radioisotope beta (electron) emissions into usable electrical power via beta emissions directly impinging on a semiconductor junction device was first proposed in the 1950's. Incident beta particles absorbed in a semiconductor create electron-hole-pairs (EHPs) which are accelerated by the built-in field to device terminals, and result in a current supplied to a load resistor. These devices are known as Direct Conversion Semiconductor Devices, Beta Cells, Betavoltaic Devices, Betavoltaic Batteries, Isotope Batteries etc. These direct conversion devices promise to deliver consistent long-term battery power for years and even decades. For this reason, many attempts have been made to commercialize such a device. However, in the hopes of achieving reasonable power levels, the radioisotope of choice often emitted unsafe amounts of high energy radiation that would either quickly degrade semiconductor device properties within the betavoltaic battery or the surrounding electronic devices powered by the battery. The radiated energy may also be harmful to operators in the vicinity of the battery. As a result of these disadvantages and in an effort to gain approval from nuclear regulatory agencies for these types of batteries, the choice for radioisotopes has been limited to low-energy beta (electron) emitting radioisotopes, such as nickel-63, promethium-147 or tritium. Due to the fact that promethium-147 is regulated more stringently and requires considerable shielding and nickel-63 has a relatively low beta flux, tritium has emerged as a leading candidate for such a battery device. Tritium betavoltaic batteries, sometimes referred to as tritium betavoltaic devices or tritium direct conversion devices, have been promoted during the last thirty years. Tritium is a relatively benign radioisotope with low beta energy emission that can easily be shielded with as little as a thin sheet of paper. Tritium has a long track record in commercial use in illumination devices such as EXIT signs in commercial aircraft, stores, school buildings and theatres. It is also widely used in gun sights and watch dials, making it an ideal power source for the direct conversion devices. Unfortunately, tritium's beta emissions are so low in energy that it is has been difficult to efficiently convert it into usable electrical power for even the most low power applications, such as powering SRAM memory to prevent the loss of stored data. Several attempts have been made to produce useful current from a tritium betavoltaic battery. For example, polycrystalline or amorphous semiconductor devices have been considered for tritium betavoltaic batteries based on the assumption that such devices would allow batteries to be fabricated at a reduced cost. It is assumed that these devices could be manufactured in a thin-film like fashion and that tritium could be embedded within the polycrystalline or amorphous devices. However, this approach is extremely inefficient (much less than 1%) with respect to the beta energy emissions entering the semiconductor. The main reason for this low semiconductor conversion efficiency is the high dark current or leakage current of the semiconductor that acts as a negative current. This high dark current competes with the betavoltaic current produced by collection of EHPs created via the tritium beta particles impinging on the semiconductor. In short, the polycrystalline and amorphous semiconductors have a high number of defects resulting in recombination centers for the EHPs, which in turn significantly reduce the betavoltaic current and lead to very low efficiency for the battery. The best results for tritium betavoltaics have been achieved with single crystal semiconductor devices. Recent attempts have involved single crystalline semiconductor devices with a tritium source such as a tritiated polymer, aerogel or tritiated metal hydride placed in direct contact with a semiconductor junction device. Single crystalline semiconductors have longer carrier lifetimes and fewer defects resulting in much lower dark currents. To date, the highest reported efficiencies for tritium betavoltaic batteries were published in a reference text entitled: “Polymers, Phosphors and Voltaics for Radioisotope Microbatteries” edited by K. Bower et al. Single crystal semiconductor devices were exposed to tritium metal hydride sources on top of the semiconductors. Several homojunction semiconductor cells were utilized with the following results: Silicon Cells: Short Circuit Current=18.1 nA/cm^2 Open Circuit Voltage=0.162 Fill Factor=0.513 Tritiated Titanium Source=0.23 microwatts/cm^2 Efficiency=1.3% Aluminum Gallium Arsenide (AlGaAs) Cells: Short Circuit Current=58 nA/cm^2 Open Circuit Voltage=0.62 Fill Factor=0.751, Power=27 nW/cm^2 Tritiated Titanium Source=0.48 microwatts/cm^2, Efficiency=5.6% Silicon cells are a preferred choice due to their low cost. However, their low efficiency makes them a poor choice for even the most low power applications, such as SRAM memory devices. The performance of the AlGaAs homojunction cell is attractive with one of the highest reported efficiencies and would be suitable for powering an SRAM memory device through the stacking of tritiated metal hydride layers and AlGaAs homojunction cells. However, AlGaAs homojunctions cells are difficult to reproduce consistently with uniform dark currents across a semiconductor device due to the oxidation of the aluminum. As a result, AlGaAs is also an expensive option to scale up. Safety concerns over containment of the tritium based betavoltaic battery have emerged as another obstacle to commercialization of a tritium battery. In commercially available products such as tritium illumination devices (e.g. EXIT signs, gun sights and watch dials), the tritium is in gaseous form and contained within a glass vial. Many accidents involving tritium release due to the breakage of the tritium vials in EXIT signs have caused public concerns and resulted in costly clean-up operations. In the case of a tritium betavoltaic battery utilizing solid-state tritium metal hydride sources the risk for exposure is lower compared to gaseous tritium devices. However, the tritium metal hydride still involves a miniscule amount of tritium release when open to the environment at room temperature. Although several tritium based batteries have been proposed including direct conversion devices built within an integrated circuit, a method of effectively hermetically packaging the battery containing the tritium metal hydride has yet to be proposed. A major obstacle to hermetically sealing this type of battery is the risk associated with using a sealing process that involves high temperatures, i.e., above 200-300° C., where tritium is released from the metal hydride causing failure of the battery after sealing or worse, causing tritium exposure at the manufacturing facility and to the operator of the equipment for sealing the battery. In addition to the above listed obstacles, the texturing of a direct conversion semiconductor device to increase the surface area exposed to radiation emission has been proposed several times in the past. For example, on page 282 of the book entitled “Polymers, Phosphors and Voltaics for Radioisotope Microbatteries” edited by K. Bower et al., the use of porous silicon and tritium inserted into porous silicon holes was proposed as a means of increasing the surface area of the semiconductor device by 20 to 50 times, in contrast to the original planar semiconductor surface area. The following published patent applications and patents each propose a method of increasing the surface area of the semiconductor by textured growth of the semiconductor or a post-growth texturing method: US Patent Application Publication 2004/0154656 US Patent Application Publication 2007/0080605 U.S. Pat. No. 7,250,323 U.S. Pat. No. 6,949,865 Central to this approach is the hope that an increase in surface area exposed to radioisotope emissions will increase the power per unit volume of the direct conversion semiconductor device. The overall goal of this approach is to not only reduce the size of the direct conversion device but also to potentially reduce the cost associated with producing the equivalent surface area in a planar semiconductor device. The problem with such an approach arises when a relatively low energy radioisotope such as tritium is used. In this case, the incident power is quite small per unit area exposed and the dark current of the semiconductor device is a very significant factor in the overall efficiency of the device. For this reason, it is preferable to use single crystal semiconductors where device defects are minimized and the dark current is sufficiently low so that power can be produced efficiently. Unfortunately, alterations to the semiconductor surface, as proposed above, risk increasing lattice defects, resulting in a high number of recombination centers for EHPs. This creates a direct conversion semiconductor device with a low open circuit voltage and reduced short circuit current resulting in a low overall efficiency. In accordance with common practice, the various described features are not drawn to scale, but are drawn to emphasize specific features relevant to the invention. Like reference characters denote like elements throughout the figures and text. Before describing in detail the particular methods and apparatuses related to tritium direct conversion semiconductor devices, it should be observed that the present invention resides primarily in a novel and non-obvious combination of elements and process steps. So as not to obscure the disclosure with details that will be readily apparent to those skilled in the art, certain conventional elements and steps have been presented with lesser detail, while the drawings and the specification describe in greater detail other elements and steps pertinent to understanding the invention. The following embodiments are not intended to define limits as to the structure or method of the invention, but only to provide exemplary constructions. The embodiments are permissive rather than mandatory and illustrative rather than exhaustive. The present invention relates to a tritium direct conversion semiconductor device comprised of a single crystal semiconductor and a device structure with both a low dark current and high efficiency for power conversion of tritium's beta emissions into electrical power. It should be understood that the high efficiency and longevity (e.g. over 10 years) of the various device structure embodiments are suitable for use with other candidate radioisotopes for betavoltaic operations (e.g., promethium-147 and nickel-63). One embodiment of the present invention proposes a novel use of Indium Gallium Phosphide homojunction semiconductor 8 in conjunction with a tritiated metal hydride source 10, as illustrated in FIG. 1, for supplying power to a load 12. The tritiated metal hydride source (e.g., scandium tritide, titanium tritide, palladium tritide, magnesium tritide, lithium tritide, or any combination thereof etc.) is directly in contact with the semiconductor to generate electrical power at an efficiency of 7.5% or higher with respect to the beta electrons impinging on the Indium Gallium Phosphide homojunction. InGaP is one of the larger band gap materials and has only recently been used in a tritium based direct conversion battery. One embodiment uses a composition of the Indium Gallium Phosphide homojunction comprising In0.49Ga0.51P (subsequently referred to as InGaP). The band gap of this semiconductor is 1.9 eV and the materials production technology is well developed by the solar cell industry. The technology also lends itself to high quality growth with a low density of lattice defects and low dark current characteristics. In addition, InGaP may be mass produced with a high yield due to its manufacturing process maturity, thus lowering the cost of tritium betavoltaic batteries based on InGaP. InGaP device structures are grown by metal-organic-vapor-deposition (MOCVD) as is known by those skilled in the art. The description of this embodiment presents novel and non-obvious features that allow efficient conversion of tritium beta flux to electrical power. FIGS. 2 and 3 illustrate the physical structure and electron band diagram, respectively. Each layer has the same lattice constant as the GaAs substrate so that the number of dislocations generated by growth of the individual layers is minimized. The beta particles represented by arrowheads in FIG. 2 are released by the tritiated scandium material of FIG. 1. FIG. 2 illustrates the individual layers of the n/p homojunction semiconductor 8, comprising, from the bottom: a pGaAs substrate (notation for a p type GaAs substrate) a pGaAs layer (grown to establish crystal structure) a p+InGaP layer (a back surface field or minority carrier reflector) a pInGaP layer (base) an intrinsic InGaP layer (for preventing diffusion of dopants between the p-doped and n-doped layers) an nInGaP layer (emitter) an nInAlP layer (window layer closely matched to the nInGaP and cap layers that allows electrons to pass to the cap layer and reflects holes back to the emitter) an nGaAs cap layer (may be highly doped)If the dopant types are reversed from those set forth above, the structure may be referred to as a p/n homojunction semiconductor. There are several features of this structure that allow efficient betavoltaic energy conversion: (a) High quality, large band gap semiconductor junction resulting in a highly efficient device; (b) Back-surface field that reflects electrons back onto the junction field with a highly doped p+InGaP layer (can also be created by p-type InAlP or InAlGaP or ZnSe); (c) A lattice-matched n-type InAlP window layer to reflects holes back to the emitter leading to a low dark current (can also be created with a highly doped n+InAlGaP, ZnSe, AlAs, n+InGaP or AlAsP); (d) A GaAs Cap layer of about a few hundred angstroms or less covering the top surface; and (e) a 1000 to 3000 Å layer of intrinsic InGaP to act as a buffer to diffusion of the p type dopant (usually Zn) into the n-type emitter region. The features (a), (b) and (c) may be important for solar cell operation but their utilization in tritium betavoltaic application is considered novel in the present embodiment. The novel features (d) and (e) may be important for betavoltaic conversion, but is not necessarily used for photovoltaic energy conversion. All of these features allow the achievement of the low dark currents required for efficient betavoltaic energy conversion. The novel lattice-matched InAlP window layer prevents the formation of dislocations at the InAlP-InGaP interface, which would increase the dark current. The GaAs cap layer keeps the InAlP layer from oxidizing, the absence of which could introduce defects for EHP recombination at the InAlP-InGaP region. This cap layer, therefore augments hole reflections at that interface. The GaAs cap layer does not absorb a significant percentage of the beta flux, and therefore can be tolerated. It should be noted that the cap layer can be made out of other III-V materials or combinations of III-V materials that can function in a similar capacity. In another embodiment, the cap layer is about 50-100 Angstroms thick or less. According to this embodiment conductive gridlines are deposited on top of a thick (for example, about 3000 to 10000 Angstroms) cap layer and then the cap layer is removed by an etch process, except the cap layer material under the grid lines. To retain this cap layer material, the etch process can be timed so that about 50-100 Angstroms of cap layer material remains. Alternatively, an etch stop layer can be formed and located such that when the etchant reaches the etch stop layer about 50-100 Angstroms of cap layer material remains. In order to construct a thin 50-100 Angstrom cap layer for protection of the window a selective etch of the GaAs cap layer that stops at a thin InGaP layer may be used, leaving a thin layer of InGaP over the 50-100 Angstrom GaAs cap layer, and if desired, the InGaP layer may be selectively etched down to the GaAs cap layer. In solar cell operation the GaAs cap layer is typically removed except under the metal gridline contacts. This is required since a cap layer across regions between the metal gridline contacts would reduce the efficiency of the solar cell due to significant absorption of the solar photons. For this reason, the GaAs cap layer is etched away completely in a solar cell, except for the regions under the gridline metal contacts. Since in solar cell operation the remaining GaAs cap layer under the metal gridline contacts is a conduit for the electrons to the grid metal lines, the GaAs cap layer is normally doped to a high level of 10^19 ND/cm^3 in order to create good conduction for milliamps or higher current levels required in photovoltaic operations. High doping of the GaAs cap may unfortunately create defects in the n-type InAlP layer, which could increase the dark current. For betavoltaic operations, an increase in the dark current could cause a reduction of cell efficiency due to the high doping of this layer. This is not important for photovoltaic operations since the dark current is so low compared to the milliamp current levels generated in a solar cell photovoltaic operation, but it is extremely important for the betavoltaic operation where the current levels are in the range of nanoamps. For this reason, the novel application of a cap layer with reduced doping may be introduced. The betavoltaic GaAs cap layer doping may be reduced to a level of 10^18 ND/cm^3, or less, thereby reducing the number of defects that may result from diffusion of the GaAs Cap layer dopant into the n-type InAlP layer. The novel intrinsic InGaP layer is not used in photovoltaic operation but may be important for betavoltaic operation due to the fact that it helps achieve low dark currents. All layers of the InGaP device structure are grown at high temperatures (e.g. 500° C.-700° C.). In particular, the intrinsic layer, the n-InGaP emitter layer, the n-InAlP layer and the n-GaAs cap layer are all grown at high temperatures. During the time required for growth of these layers, the p-type dopant zinc in the p-InGaP layer will diffuse toward the n-type films. If the intrinsic layer is too thin and allows zinc to diffuse into the emitter layer and the InAlP region, the dark current will increase and the betavoltaic device performance is degraded. Thus, since low dark currents are critical for tritium betavoltaic energy conversion, the intrinsic layer must be thick enough to be an effective buffer to zinc diffusion. An intrinsic layer of approximately 1000-3000 Angstroms or more is sufficient to produce a low dark current, betavoltaic device. Although the present invention utilizes an intrinsic layer of InGaP that is 1000-3000 Å, it is also possible in one embodiment of the invention to remove the intrinsic layer or to use a substantially smaller intrinsic layer thickness of about e.g., 50-100 Å. It should be noted that the tritium InGaP betavoltaic structure presents novel and non-obvious features that provide a low dark current and a high voltage and collection efficiency. The following data was obtained with solid tritiated metal hydride sources (e.g. titanium tritide, scandium tritide etc.) and have the highest reported efficiency of 7.5% with respect to the incident beta radiation impinging on the InGaP homojunction. In particular, for a tritiated scandium source with a 250 to 500 nanometer thick scandium film and an InGaP homojunction as shown in FIG. 2 the following results were achieved: Short Circuit Current=45.2 nA/cm^2 Open Circuit Voltage=0.77 Fill Factor=0.79, Power=27.5 nW/cm^2 Tritiated Scandium Source=0.369 microwatts/cm^2, Efficiency=7.5% In yet another embodiment the dopants may be reversed for all layers to produce a p/n structure. In particular, starting from the bottom: a nGaAs substrate a nGaAs layer (grown to establish crystal structure) a n+InGaP layer (a back surface field or minority carrier reflector) a nInGaP layer (base) an intrinsic InGaP layer (for preventing diffusion of dopants between the p-doped and n-doped layers) an pInGaP layer (emitter) an pInAlP layer (window layer closely matched to the pInGaP and cap layers) that allows holes to pass to the cap layer and reflects electrons back to the emitter) a pGaAs cap layer (may be highly doped p-type) In general the present invention demonstrates that the intrinsic layer in tritium betavoltaic devices serves three important purposes: (a) it acts as a buffer to diffusion of dopant atoms from the base region into the emitter region; (b) it allows efficient collection of electron-hole pairs produced as a result of beta particle absorption; and (c) as a consequence, the base region can be heavily doped so that the built-in voltage can be maximized. The high dopant density in the base region (with reference to FIG. 2, the pInGaP layer) is novel to the betavoltaic structure. This is due to the fact that it is not necessary to have a finite diffusion length in the base region for efficient carrier collection; hence a relatively high dopant density can be used in the base region to maximize the built-in potential. Minimizing diffusion of dopant atoms from the base to the emitter and window layers is desirable for achieving a low dark current. With EHPs mainly produced in the emitter and high field intrinsic region, a large collection efficiency can be achieved. Tritium beta particle penetration in semiconductors is less than about one micron. Thus, it is clear that the emitter and window layers need to be very thin, preferably on the order of a few hundred Å so that most of the beta particle absorption occurs in the high field region in the depletion layer (with respect to FIG. 2, the intrinsic InGaP layer or in another embodiment a material region between a p-doped and an n-doped region). Homojunctions are typically formed by abruptly reducing one dopant (e.g., for n-type material) and immediately introducing the other dopant (e.g., for p-type material). The intrinsic regions formed in devices discussed herein are created by reducing one dopant input to zero followed by film growth with neither donors nor acceptors introduced to form the intrinsic layer, and then initiating introduction of the other dopant. Unless noted otherwise, in one embodiment all of the device structures considered herein have an intrinsic layer between the emitter (e.g., the nInGaP layer) and base region (the pInGaP layer). The thickness of the intrinsic layer is selected so that most of the beta particle absorption occurs in the emitter and intrinsic layers. The basic approach to solar cell fabrication does not typically include the intentional formation of a relatively wide intrinsic layer. However, since the tritium betas are absorbed in a few thousand Angstroms, there is great flexibility regarding an increased doping density in the base. In another embodiment, the betavoltaic structure may be alloyed with 1-3% Aluminum to achieve a slightly higher bandgap. The various approaches based on GaAs substrates are summarized in Tables 1, 2 and 3. Included are structures based on materials that lattice match GaAs but have larger electron bandgaps, namely, In(AlGa)P and InAlP. It should be noted that several betavoltaic structure embodiments utilize tunnel junctions to serve as a means of changing the dominant carrier from electrons to holes, or vice versa. Two different types of layers are utilized in the tunnel junction structures. These tunnel junctions involve a heavily doped n-layer adjacent to a heavily doped p-layer, referred to as n++ and p++ layers, respectively. Thicknesses are typically 100 A for both the n++ and p++ layers although they can range from approximately 50 A to 200 A. The dopant levels are typically 5E18 to 1E19 cm-3 for Zinc in p++ layers and similarly for Silicon doping in n++ layers. Other features of various materials and layers are set forth in the reference notes provided with the Tables below. TABLE 1InGaP Cells on GaAs Substratesp-GaAs Substratesn-GaAs Substratesn/p InGaPp/n InGaPn/p InGaPp/n InGaPLayer 1p-GaAs Subp-GaAs Subn-GaAs Subn-GaAs SubLayer 2p-GaAs (Est Crs Struc)p-GaAs (Est Crs Struc)n-GaAs (Est Crs Struc)n-GaAs (Est Crs Struc)Layer 3p InAlP (Reflector)p++GaAs (p-layer TJ)n++GaAs (p-layer TJ)n InAlP (Reflector)Layer 4pInGaP (base)n++GaAs (n-layer TJ)p++GaAs (n-layer TJ)nInGaP (base)Layer 5InGaP (i-Layer)nInAlP (Reflector)pInAlP (Reflector)InGaP (i-Layer)Layer 6nInGaP (emitter)nInGaP (base)pInGaP (base)pInGaP (emitter)Layer 7nInAlP (Window)InGaP (i-layer)InGaP (i-layer)pInAlP (Window)Layer 8n++GaAs cap LayerpInGaP (emitter)nInGaP (emitter)p++GaAs cap LayerLayer 9pInAlP WindownInAlP WindowLayer 10p++GaAs cap layern++GaAs cap layerNotes:1. Est Crs Struc refers to a layer to establish the crystal structure.2. Tj designates tunnel junction.3. InGaP refers to a compound InxGa(1 − x)P, where x = 0.48, which lattice matches GaAs.4. InAlP refers to a compound InxAl(1 − x), where x = 0.48, which lattice matches GaAs5. P materials will typically be doped with about 1E17 cm−3 of Zn, p+ up to about 1E18 cm−3, whereas those designated as p++ may be doped to a level of about 5E18 cm−3 Zn.6. N materials are doped with Si with the doping range also being in the range of about 1E17 cm−3 to 5E18 cm−3.7. Reflector layer refers to layer at an interface between the base layer and the substrate that reflects minority carriers so that recombination losses are minimized.8. Window layer refers to a layer adjacent the emitter that allows majority carriers pass to the cap layer and reflects minority carriers to minimize recombination losses.9. The cap layer is heavily doped so that it is very conductive. TABLE 2In(AlGa)P Cells on GaAs Substratesp-GaAs Substratesn-GaAs Substratesn/p In(AlGa)Pp/n (AlGa)Pn/p (AlGa)Pp/n (AlGa)PLayer 1p-GaAs Subp-GaAs Subn-GaAs Subn-GaAs SubLayer 2p-GaAs (Est Crs Struc)p-GaAs (Est Crs Struc)n-GaAs (Est Crs Struc)n-GaAs (Est Crs Struc)Layer 3p InAlP (Reflector)p++GaAs (p-layer TJ)n++GaAs (p-layer TJ)n InAlP (Reflector)Layer 4pIn(AlGa)P (base)n++GaAs (n-layer TJ)p++GaAs (n-layer TJ)nIn(AlGa)P (base)Layer 5In(AlGa)P (i-Layer)nInAlP (Reflector)pInAlP (Reflector)InGaP (i-Layer)Layer 6nIn(AlGa)P (emitter)nIn(AlGa)P (base)pIn(AlGa)P (base)pIn(AlGa)P (emitter)Layer 7nInAlP (Window)In(AlGa)P (i-layer)In(AlGa)P (i-layer)pInAlP (Window)Layer 8n++GaAs cap LayerpIn(AlGa)P (emitter)nIn(AlGa) P (emitter)p++GaAs cap LayerLayer 9pInAlP WindownInAlP WindowLayer 10p++GaAs cap layern++GaAs cap layerNotes:1. Est Crs Struc refers to a layer to establish the crystal structure.2. Tj designates a tunnel junction.3. InGaP refers to a compound InxGa(1 − x)P, where x = 0.48, which lattice matches GaAs.4. InAlP refers to a compound InxAl(1 − x), where x = 0.48, which lattice matches GaAs5. In(AlGa)P refers a compound Inx(AlyGa(1 − y))(1 − x)P with x = 0.48 and y = 0.2.6. P type materials will typically be doped with about 1E17 cm−3 of Zn, p+ up to about 1E18 cm−3, whereas those designated as p++ may be doped to a level of about 5E18 cm−3 Zn.7. N type materials are doped with Si with the doping range also being in the range of about 1E17 cm−3 to about 5E18 cm−3.8. A reflector layer refers to layer at an interface between the base layer and substrate that reflects minority carriers so that recombination losses are minimized.9. A window layer refers to a layer adjacent the emitter that allows majority carriers pass to the cap layer and reflects minority carriers to minimize recombination losses.10. The cap layer is heavily doped so that it is very conductive. TABLE 3InAlP Cells on GaAs Substratesp-GaAs Substratesn-GaAs Substratesn/p InAlPp/n InAlPn/p InAlPp/n InAlPLayer 1p-GaAs Subp-GaAs Subn-GaAs Subn-GaAs SubLayer 2p-GaAs (Est Crs Struc)p-GaAs (Est Crs Struc)n-GaAs (Est Crs Struc)n-GaAs (Est Crs Struc)Layer 3p+ InAlP (Reflector)p++GaAs (p-layer TJ)n++GaAs (p-layer TJ)n+InAlP (Reflector)Layer 4pInAlP (base)n++GaAs (n-layer TJ)p++GaAs (n-layer TJ)nInAlP (base)Layer 5InAlP (i-Layer)n+InAlP (Reflector)p+InAlP (Reflector)InAlP (i-Layer)Layer 6nIAlaP (emitter)nInAlP (base)pInAlP (base)pInAlP (emitter)Layer 7n+InAlP (Window)InAlP (i-layer)InAlP (i-layer)p+InAlP (Window)Layer 8n+GaAs cap LayerpInAlP (emitter)nInAlP (emitter)p+GaAs cap LayerLayer 9p+InAlP Windown+InAlP WindowLayer 10p++GaAs cap layern++GaAs cap layerNotes:1. Est Crs Struc refers a layer that establishes a crystal structure.2. Tj designates a tunnel junction.3. InGaP refers to a compound InxGa(1 − x)P, where x = 0.48, which lattice matches GaAs.4. InAlP refers to a compound InxAl(1 − x)P, where x = 0.48, which lattice matches GaAs5. P type materials will typically be doped with about 1E17 cm−3 of Zn, p+ up to 1E18 cm−3, whereas those designated as p++ may be doped to a level of about 5E18 cm−3 Zn.6. N type materials are doped with Si with the doping range also being in the range of about 1E17 cm−3 to about 5E18 cm−3.7. A reflector refers to layer at an interface between the base layer and substrate that reflects minority carriers so that recombination losses are minimized.8. A window refers to a layer adjacent the emitter that allows majority carriers pass to the cap layer and reflects minority carriers to minimize recombination losses.9. The cap layer is heavily doped so that it is very conductive. In other embodiments the GaAs substrate is replaced by a Germanium substrate. Growth of high quality GaAs films onto Ge requires the growth of a nucleation layer on the Ge. The general approach involves the growth of a GaAs at a relatively low temperature (LT) of about 550 C followed by growth of GaAs at a temperature more commonly used for GaAs, namely about 700 C. Growth of the LT GaAs leads to a relatively smooth GaAs layer which improves subsequent growth of the high temperature GaAs layer. In order to reduce the generation of anti-phase boundaries, which can lead to recombination centers, off-oriented Ge(001) substrates are used for growth of the low temperature GaAs layer. Although modifications of this approach have been developed, growth of GaAs nucleation layers resulting in high quality GaAs films on Ge substrates usually involve these two features: use of off-oriented Ge(001) substrates and the LT film of GaAs. In general layers in structures based on Ge substrates parallel those grown on GaAs substrates. Both n- and p-type Ge substrates are used and both types of substrates must first have nucleation layers grown in order to achieve the growth of high quality GaAs films. However, one unique feature must be dealt with when growing on p-Ge substrates. Growth of As- and P-containing films on p-Ge substrates results in the formation of an n-type layer on the surface of the Ge substrate. To counter the formation of this artifact layer, the first layer grown on the p-Ge substrate is heavily doped p-GaAs. This GaAs layer thus serves two purposes, it establishes a GaAs crystalline structure and the large Zn doping level (about 1E18 to about 5E18) mitigates the potential problem presented by the possible formation of an n-type layer on the p-type Ge substrate. In both cases of n- and p-type Ge substrates, Structures based on InGaP, In(AlGa)P and InAlP grown on Ge substrates are summarized in Tables 4, 5 and 6. TABLE 4InGaP Cells on Ge Substratesp-Ge Substratesn-Ge Substratesn/p InGaPp/n InGaPn/p InGaPp/n InGaPLayer 1p-Ge Subp-Ge Subn-Ge Subn-Ge SubLayer 2p+GaAs (Nucl Layer)p+GaAs (Nucl Layer)n-GaAs (Nucl Layer)n-GaAs (Nucl Layer)Layer 3p+GaAs (Est Crs Struc)p+GaAs (Est Crs Struc)n-GaAs (Est Crs Struc)n-GaAs (Est Crs Struc)Layer 4p InAlP (Reflector)p++GaAs (p-layer TJ)n++GaAs (n-layer TJ)n InAlP (Reflector)Layer 5pInGaP (base)n++GaAs (n-layer TJ)p++GaAs (p-layer TJnInGaP (base)Layer 6InGaP (i-Layer)nInAlP (Reflector)pInAlP (Reflector)InGaP (i-Layer)Layer 7nInGaP (emitter)nInGaP (base)pInGaP (base)pInGaP (emitter)Layer 8nInAlP (Window)InGaP (i-layer)InGaP (i-layer)pInAlP (Window)Layer 9n+GaAs cap LayerpInGaP (emitter)nInGaP (emitter)p+GaAs cap LayerLayer 10pInAlP WindownInAlP WindowLayer 11p+GaAs cap layern+GaAs cap layerNotes:1. Est Crs Struc refers to a layer for establishing the crystal structure.2. Nucl Layer refers to nucleation layer as discussed elsewhere herein.3. Layer 3 for the structures grown on p-Ge substrates is doped with Zn to a level of about 1E18 cm−3 to prevent formation of an artifact n-layer. The layer also establishes the GaAs crystal structure. For the p-Ge substrate and the n/p InGaP embodiment it may not be necessary to dope layer 3 to a p+ level, as a lower doping may be sufficient to establish the crystal structure without having to counteract the effects of the artifact layer.4. Tj designates a tunnel junction5. InGaP refers to a compound InxGa(1 − x)P, where x = 0.48, which lattice matches GaAs.6. InAlP refers to a compound InxAl(1 − x), where x = 0.48, which lattice matches GaAs7. P materials will typically be doped with about 1E17 cm−3 of Zn, p+ up to about 1E18 cm−3, whereas those designated as p++ may be doped to a level of about 5E18 cm−3 with Zn.8. N materials are doped with Si with the doping range of about 1E17 cm−3 to about 5E18 cm−3.9. A reflector layer refers to layer at an interface between the base layer and substrate that reflects minority carriers so that recombination losses are minimized.10. A window layer refers to a layer adjacent the emitter that allows majority carriers to pass to the cap layer and reflects minority carriers to minimize recombinations.11. The cap layer is heavily doped so that it is very conductive. TABLE 5In(AlGa)P Cells on Ge Substratesp-Ge Substratesn-Ge Substratesn/p In(AlGa)Pp/n In(AlGa)Pn/p In(AlGa)Pp/n In(AlGa)PLayer 1p-Ge Subp-Ge Subn-Ge Subn-Ge SubLayer 2p+GaAs (Nucl Layer)p+GaAs (Nucl Layer)n-GaAs (Nucl Layer)n-GaAs (Nucl Layer)Layer 3p-GaAs (Est Crs Struc)p-GaAs (Est Crs Struc)n-GaAs (Est Crs Struc)n-GaAs Est Crs Struc)Layer 4pInAlP (Reflector)p++GaAs (p-layer TJ)n++GaAs (n-layer TJ)nInAlP (Reflector)Layer 5pIn(AlGa)P (base)n++GaAs (n-layer TJ)p++GaAs (p-layer TJ)nIn(AlGa)P (base)Layer 6In(AlGa)P (i-Layer)nInAlP (Reflector)pInAlP (Reflector)In(AlGa)P (i-Layer)Layer 7nIn(AlGa)P (emitter)nIn(AlGa)P (base)pIn(AlGa)P (base)pIn(AlGa)P(emitter)Layer 8nInAlP (Window)In(AlGa)P (i-layer)In(AlGa)P (i-layer)pInAlP (Window)Layer 9n+GaAs cap LayerpIn(AlGa)P (emitter)nIn(AlGa)P (emitter)p+GaAs cap LayerLayer 10pInAlP WindownInAlP WindowLayer 11p+GaAs cap layern+GaAs cap layerNotes:1. Est Crs Struc refers to a layer to establish crystal structure2. Nucl Layer refers to nucleation layer as discussed elsewhere herein.3. Layer 3 for the structures grown on p-Ge substrates is doped with Zn to a level of about 1E18 cm−3 to prevent formation of the artifact n-layer. The layer also establishes GaAs structure.4. Tj designates a tunnel junction5. InGaP refers to a compound InxGa1 − xP (x = 0.48), which lattice matches GaAs.6. In(AlGa)P refers to a compound Inx(AlyGa(1 − y))(1 − x)P, where x = 0.48 and y = .2, which lattice matches GaAs.7. P type materials will typically be doped with about 1E17 cm−3 of Zn, p+ up to about 1E18 cm−3, whereas those designated as p++ may be doped to a level of about 5E18 cm−3 Zn.8. N type materials are doped with Si with the doping range being in the range of about 1E17 cm−3th about 5E18 cm−3.9. A reflector layer refers to layer at an interface between the base layer and a substrate that reflects minority carriers so that recombination losses are minimized.10. A window layer refers to a layer adjacent the emitter that allows majority carriers pass to the cap layer and reflects minority carriers to minimize recombination losses.11. The cap layer is heavily doped so that it is very conductive. TABLE 6InAlP Cells on Ge Substratesp-Ge Substratesn-Ge Substratesn/p InAlPp/n InAlPn/p InAlPp/n InAlPLayer 1p-Ge Subp-Ge Subn-Ge Subn-Ge SubLayer 2p+GaAs (Nucl Layer)p+GaAs (Nucl Layer)n-GaAs (Nucl Layer)n-GaAs (Nucl Layer)Layer 3p+GaAs (Est Crs Struc)p+-GaAs (Est Crs Struc)n-GaAs (Est Crs Struc)n-GaAs (Est Crs Struc)Layer 4p+InAlP (Reflector)p++GaAs (p-layer TJ)n++GaAs (p-layer TJ)n+ InAlP (Reflector)Layer 5p InAlP (base)n++GaAs (n-layer TJ)p++GaAs (n-layer TJ)nInAlP (base)Layer 6InAlP (i-Layer)n+InAlP (Reflector)p+InAlP (Reflector)InAlP (i-Layer)Layer 7nInAlP (emitter)nInAlP(base)pInAlP (base)pInAlP (emitter)Layer 8n+InAlP (Window)InAlP (i-layer)InAlP (i-layer)p+InAlP (Window)Layer 9n+GaAs cap LayerpInGaP (emitter)nInAlP (emitter)p+GaAs cap LayerLayer 10p+InAlP Windown+InAlP WindowLayer 11p+GaAs cap layern+GaAs cap layerNotes:1. Est Crs Struc refers to a layer to establish crystal structure2. Nucl Layer refers to a nucleation layer as discussed in elsewhere herein.3. Layer 3 for the structures grown on p-Ge substrates is doped with Zn to a level of about 1E18 cm−3 to prevent formation of an artifact n-layer. The layer also establishes GaAs structure.4. Tj designates tunnel junction5. InGaP refers to a compound InxGa(1 − x)P, where x = 0.48, which lattice matches GaAs.6. InAlP refers to a compound InxAl(1 − x), where x = 0.48, which lattice matches GaAs7. P type materials will typically be doped with about 1E17 cm−3 of Zn, p+ up to about 1E18 cm−3, whereas those designated as p++ may be doped to a level of about 5E18 cm−3 Zn.8. N type materials are doped with Si with the doping range also being in the range of about 1E17 cm−3 to about 5E18 cm−3.9. Reflector refers to layer at an interface between the base layer and the substrate that reflects minority carriers so that recombination losses are minimized.10. A window refers to a layer adjacent the emitter that allows majority carriers pass to the cap layer and reflects minority carriers to minimize recombination losses.11. The cap layer is heavily doped so that it is very conductive. It should be noted that in one embodiment a betavoltaic cell may be grown on each side of the semiconductor substrate. For instance, for either GaAs or Ge substrates a combination of a p/n and n/p betavoltaic structure (e.g. InGaP, InAlP, InAlGaP or others) may be grown on opposing sides of the substrate to create a bilateral betavoltaic structure (i.e. double-side betavoltaic structure) yielding twice the voltage of a single-sided counterpart. Just as in the case of a single-sided betavoltaic, each side of the bilateral cell will require a source of beta flux impinging on the respective junctions. The use of tunnel junctions may be required to allow the current to flow in a series arrangement between the two betavoltaic cells. In one embodiment of the present invention, the tritium source is a tritium metal hydride (sometimes referred to as a metal tritide), that is in contact with the top surface of the betavoltaic structure as shown in FIG. 1. The metal tritide may be formed by depositing one or a combination of hydride forming metals (e.g. scandium, titanium, magnesium, palladium, lithium etc.). Thicknesses of the metal tritide layer are typically less than one micron and can be as thin as 50-100 nanometers. The metal layer may be placed on top of the betavoltaic cell by directly depositing on top of the betavoltaic cell's active area (e.g. window or cap layer) through methods known in the art (e.g. evaporation, electro deposition etc.). Alternatively, the metal tritide layer may be deposited on a separate thin substrate (e.g. ˜25 microns to ˜500 microns or in the millimeter thickness range) that is mechanically connected to the betavoltaic cell's active area via pressure, epoxy or spot welding. The metal tritide is typically formed by exposure to tritium gas at pressures ranging 0.25 to 20 Bar and temperatures ranging approximately 100° C. to 600° C. for durations ranging minutes to days. A layer of palladium ranging from approximately 1 nanometer to 500 nanometers may be deposited over (i.e. capping-off) a scandium, titanium, magnesium or lithium metal or other tritide forming metal in order to reduce the tritium loading temperature and stabilize the tritium within the metal matrix after the tritide has been formed. The metal tritide layer may also be formed by an in-situ evaporation of the metal in the presence of tritium. Bi-directional metal tritide sources (i.e. with betas emanating from opposing surfaces) may be utilized in this invention. For example, the metal tritide may be formed as a film on top of the betavoltaic cell's active area allowing for a second cell to be place in direct contact with its active area as in FIGS. 6-7. In another embodiment of the present invention the contact lines on the top surface of the betavoltaic homojunction can be very thin and on the perimeter of the semiconductor. This contact ring is used to collect the current from the semiconductor while providing a minimal shadowing effect to the radioactive source's beta flux that impinges on the surface of the semiconductor. The contact ring for the betavoltaic semiconductor may be formed in the same manner as solar cell industry uses to make contact gridlines on the solar cell semiconductor. However, the betavoltaic cell contact ring is substantially different from a solar cell where a series of gridlines are uniformly covering the surface of the semiconductor and can cover approximately 5-10% of the semiconductor surface. This uniform coverage creates a shadowing effect resulting in a proportional loss of power from the solar cell. In contrast the betavoltaic cell's contact ring may be reduced to a small perimeter (e.g. outlining a 1 cm×1 cm cell or 3 cm×3 cm cell etc.) or it may be just a set of contact points or lines. This configuration may be necessitated by the low magnitude of current collected from the betavoltaic device that is in the nanoamp to microamp per square centimeter range as opposed to solar cells where the range is more in the milliamp per square centimeter range. Thus, whereas solar cells require relatively low series resistance (<1 ohm per square centimeter of cell area) by the inclusion of more contact line coverage, betavoltaic cells can function efficiently with much greater values of series resistance. In yet another embodiment of this invention a thin GaAs cap layer is grown to a desired thickness (e.g. 50-100 Å or less) and uniformly covers the betavoltaic window. In this configuration, the contact metal gridlines for current collection are replaced with a tritium metal tritide deposited uniformly over the GaAs cap layer. In this configuration, the tritium metal tritide serves as both a metal contact collector and a beta-source emitter resulting in less shadowing of betas impinging on the betavoltaic cell and a simpler construction of the betavoltaic cell. As an example, the contacts in a betavoltaic semiconductor can result in a shadow coverage that is much less than about 1%, thereby providing a higher efficiency betavoltaic battery. Specific shadow coverage and thicknesses of contact ring, lines or dots required by a betavoltaic semiconductor is dictated by consideration of sheet conductance of the top surface cell layers, namely, the cap, window and emitter layers. As noted above, the sheet resistance for a tritium betavoltaic cell can be relatively large (e.g. >100 Ohms per square centimeter). In all embodiments of the present invention it may be desirable to shield the edges of the betavoltaic structure from beta particles. This constitutes another novel aspect of the present invention. As is known in the art, if the energy of a beta particle is large enough, the particle can cause the displacement of an atom in a crystalline semiconductor. Atomic vacancies can act as a recombination center for EHPs in semiconductors and can cause degradation of betavoltaic efficiencies. Fortunately, the threshold for atomic displacement in semiconductors is typically greater than 250 keV. Therefore, tritium beta particles as well as beta particles from Promethium-147 and Nickel 63 do not cause degradation of semiconductor diode properties as a result of beta absorption within the bulk of the material. However, low energy betas can create dangling bonds along the junction periphery, which can cause shunting currents or carrier recombination at the junction edges. If the edges are not properly shielded or protected from the beta flux, the betavoltaic device performance/efficiency may degrade. As illustrated in FIGS. 4-7, the junction edges may be protected by the keeping the tritium source within the perimeter contact metal gridlines at a distance such that the beta particle cannot reach the edges of the semiconductor. Furthermore, the metal perimeter contact gridlines act as a physical barrier to the beta flux, thus preventing the beta particles from hitting the edge of the device. It should be understood that protection of the edges may be accomplished through a variety of means such as all forms of physical barriers (e.g. deposited metal barriers, polymers, insulators etc.) or simply physical distance acting as a barrier to beta particles impinging on the betavoltaic semiconductor's edges. In all embodiments of the present invention, the voltage and current may be scaled up via the stacking of betavoltaic semiconductors and tritium sources (betavoltaic cells). Betavoltaic cell layers may be stacked vertically or arranged horizontally and configured electrically in series or parallel. Electrical connection can be established by utilizing through-vias as power lead contacts across betavoltaic cell layers, by using current-channeling interposers (e.g. flexible circuit cards) in between betavoltaic cells or groups of cells, or by many other methods common in the art. It should be noted that varying stacking configurations produce varying voltage and current outputs from the betavoltaic composite device as illustrated in the approaches to connecting betavoltaic cells in series and parallel configurations in FIGS. 4-7. Arranging multiple (N) layers of n/p cells in series with unidirectional beta sources is illustrated in FIG. 4. If it is assumed that all cells have identical properties, namely, the same values for short circuit current (Isc), open circuit voltage (Voc) and maximum power (Pmax), and assuming the contacts between devices are ideal, the characteristics for the series stack of N cells are:(Isc)stack=Isc,(Voc)stack=N×Voc, and (Pmax)stack=N×Pmax Electrical connection between cells can be established by a soft metal such as indium or a deposited peripheral strip of gold or another appropriate metal. Electrical contact can be made by contact pressure between metals, solders, electrically conductive epoxies, and other methods well known in the art. FIG. 4 describes an approach where the electrical connections are made on the periphery of cells. Although n/p cells are shown in FIG. 4, the same approach can be used for p/n cells. FIGS. 5A and 5B illustrate a novel approach for combining n/p and p/n cells in series with bidirectional beta sources, i.e. sources that emit beta particles in two directions as shown. This approach allows for the efficient use of a tritium layer in a bidirectional capacity. Contacts can be formed as discussed above for the series stack. If the cells have identical properties, except for polarity, the two cell unit provides:(Isc)unit=Isc,(Voc)unit=2×Voc and (Pmax)unit=2×Pmax FIGS. 6A and 6B illustrate a configuration for combining two n/p (or p/n) cells in parallel and coupled to a bidirectional source. In this case, characteristics of the two cell unit are:(Isc)unit=2×Isc,(Voc)unit=Voc and (Pmax)unit=2×Pmax An example where cells are arranged in a stack but connected in parallel electrically is depicted in FIG. 7 where such a structure of n/p (or p/n) cells are each coupled to unidirectional beta sources. Assuming there are N identical cells, each having Isc, Voc, and Pmax as cell parameters,(Isc)stack=N×Isc,(Voc)stack=Von, and (Pmax)stack=N×Pmax Joining methodologies of electronic component stacking (e.g., multi-chip stacking) such as, solder connections, wire bonding, and other conductive adhesive materials and techniques, can be utilized to join combinations of the configurations listed in FIGS. 4-7. This allows for a broad variety of design interconnectability, thus achieving betavoltaic batteries with a variety of current and voltage specifications. An embodiment of the present invention includes a method of hermetically sealing a direct conversion semiconductor battery with a tritium metal hydride source at low temperatures. During construction of the battery and sealing there is no leakage of tritium from the metal hydride due to high temperature sealing methods, such as glass frit seals or solder seals, and it poses no risk of tritium exposure to the operator sealing the battery. Additionally, the hermetic battery design and the sealing method allow for high throughput manufacturing and low contamination of tritium within the manufacturing facility. Hermetic packaging and sealing techniques for integrated circuits are widely used in the semiconductor industry to prevent dirt, moisture, particulates and ionic impurities from entering the integrated circuit package and causing corrosion of the circuit elements and interconnects. In an embodiment of the present invention a combination of these techniques and packaging designs prevents tritium from exiting the battery package. That is, the role of hermetic packaging and sealing for integrated circuits is reversed in the case of the tritium battery, from contamination entering the IC package to preventing radioactive contamination from exiting the tritium battery package. In this embodiment of the present invention, the battery package is comprised of a ceramic or metal package housing containing electrode pins or leads from an internal area of the package to an external area of the package. These leads serve as conduits of electrical power for the battery and are connected to a load on a circuit board or other device. The leads are hermetically attached and sealed via glass frits or commonly used techniques for hermetic sealing of leads. Although the lead sealing methods involve high temperature processes above 300° C., the leads are sealed on the battery housing prior to containment of the tritium metal hydride. Note, the package may take any form currently in use for IC packages, i.e. PIN device leads, leadless package, surface mounts, etc. The direct conversion semiconductor is placed or bonded within the ceramic or metal package and is connected to internal areas of the leads via wire bonds or other commonly used techniques. The tritiated metal hydride source, comprising either scandium or titanium or another suitable metal, is placed in contact with the direct conversion semiconductor. A combination of direct conversion semiconductors and tritium metal hydride source layers in series or parallel may be connected within the package. Additionally, the tritium metal hydride source layers may be deposited directly onto the direct conversion semiconductor. Also, the direct conversion semiconductors may be formed as epilayers that are approximately 5-50 microns in thickness. In one embodiment, the present invention uses a Kovar lid or Kovar step lid that closes the tritium battery package. If a ceramic package is used a side brazed Kovar seal ring must be attached using techniques commonly known in the art. Note, the Kovar seal ring is attached prior to inserting the tritiated metal hydride. The final step in completion of the betavoltaic is the sealing of the Kovar step lid to the metal package or the ceramic side brazed package with a Kovar seal ring. See FIGS. 8A (bottom view) and 8B (side view). FIG. 8A illustrates a lid step and a lid edge. The Kovar lid is sealed with a resistance or laser welder that uses localized heating well below 200° C. to hermetically weld the lid to the package. The preferred method for welding is a parallel seam welder, which is inexpensive compared to laser welding and offers a high throughput. Note, the most common method in the IC industry for hermetic sealing is the solder weld using a belt furnace. This method involves temperatures of approximately 360° C., well above the threshold for tritium containment. Testing of the tritium battery package seal is achieved by enclosing the parallel seam-sealer and the unsealed tritium battery package within a helium glove box environment. Helium is flowed across the unsealed package and the Kovar lid is then placed on the package. The sealing is performed under a helium atmosphere enclosing a helium environment within the tritium battery package. The tritium battery package is then placed in an ultra-sensitive helium detector with detection levels up to 10^-11 cc/second under a 1 atmosphere differential. A leak rate of 10^-8 cc/second under a 1 atmosphere differential is considered hermetic for the tritium battery package and easily achieved using the above method. Additionally, lower hermiticity requirements are still acceptable as long as tritium leakage is within acceptable regulatory limits. There are benefits to the operation and longevity of the betavoltaic device that are directly derived from sealing the device in an inert atmosphere. Namely, the prevention of oxidation or corrosion reactions involving both the weld joint between the lid and package, as well as oxidation that forms on the surfaces of the actual components of the betavoltaic device can be mitigated. Elimination of trapped oxygen and humidity though the use of a ultra-high-purity, very low humidity, inert gas prevents the possibility for generating an oxide product in the weld seal, which would produce an opportunity for tritium leakage out of the package, or humidity and oxygen leakage into the package. Another approach to testing of the hermetic seal may be achieved with a helium bombing system where the tritium battery package is enclosed in high-pressure helium environment. Depending on the size of the leaks within the tritium battery package the helium gas will enter the package. The package is then removed from the high-pressure environment and inserted in the ultra-sensitive helium detector unit to detect leakage rates. In another embodiment the containment of tritium and radiation emanating from the tritium metal hydride is contained within individualized tritiated direct-conversion semiconductor dies or epilayer dies. These direct-conversion dies and tritium metal hydrides can be supplied with appropriate encapsulation that serves to contain radiation. Encapsulation in the form of discrete, conformal coatings that can be applied through numerous techniques, such as dipping/immersion process, chemical/physical vapor deposition techniques, (e.g. potting, sputtering, evaporation, etc.). These coatings are applied as thin films and can be metallic or vitreous in nature, providing some modest structural support and robustness to the direct conversion dies, while still providing an important, necessary, and effective barrier to the emission of beta particles arising from tritium decay and containment of the tritium radioisotope. Encapsulation is conducted to safeguard against any radiation leakage, but would be accomplished in a conformal manner so as to leave contact leads exposed as necessary for integration into device housings and maintain geometric requirements for the dies. These dies thusly encapsulated are then facile candidates for regulatory general and/or exempt licensure; in this manner, the encapsulated materials could easily be transported or handled without any risk of radiation exposure and without any need for specialized radiation materials training. For example, the encapsulated tritium betavoltaic dies could be shipped to an OEM integrator for inclusion in an integrated circuit package without a hermetic seal. One aspect of the present invention involves increasing the surface area per unit volume in a direct conversion device without increasing the dark current, via a texturing method. Instead of texturing a surface of the betavoltaic semiconductor, an epitaxial liftoff (ELO) process is employed to remove an intact epilayer containing the betavoltaic semiconductor device. The ELO process used may be any of the techniques known to those skilled in the art. This epilayer can be made substantially free of surface defects that may harm the betavoltaic device and thus increase dark current of the device. The epilayer is approximately 0.1 microns to 5.0 microns thick, but can be as thick as 50 microns, and is usually coupled to a backing layer that may comprise a metallic layer (e.g. gold, copper, aluminum, titanium, scandium, platinum, silver, tungsten, and other alloys) or a polymer material (e.g. polyimide, Kapton, etc.). The composite epilayer, comprising the epilayer and backing layer, is approximately 5-50 microns, or as thick as 100 microns, and is flexible. Conversely, the backing layer may serve a dual purpose as a metal tritide that can approximately double the creation of EHPs within betavoltaic epilayer. This is due to the nascent thinness of the device and the high diffusions lengths of the charge carriers in InGaP and other III-V structures allowing for betavoltaic operation for betas entering through the base layer. Moreover, a metal tritide that is formed on both sides of a betavoltaic epilayer provides for a symmetric distribution of forces under thermal expansion providing structural integrity to the betavoltaic epilayer. As described previously, the metal tritide can also be sealed with barrier layers (e.g. metallic, polymer, semiconductor, ceramic etc.) preventing the diffusion or migration of tritium or tritium species out of the metal tritide and providing shielding against radiation emanation. Furthermore, the composite epilayer and tritium metal hydride (i.e., betavoltaic epilayer dies) comprises a thin betavoltaic device that may be stacked in series or parallel configurations as described in the teachings of the various embodiments of the invention. A single composite epilayer and tritium metal hydride can be as thin as 1 to 10 microns and have an approximate power range of 0.1 to 0.2 microwatts. Furthermore, via stacking of these individual layers the power density can reach as high as 100-2000 microwatts/cm^3, thereby achieving an increase in surface area per unit volume resulting in a significant increase in power per unit volume. In one embodiment the tritium metal hydride film may be deposited directly on one or both sides of the tritium betavoltaic epilayer. In another embodiment, the tritium metal hydride may be formed on a separate thin substrate or thin foil (e.g. less than 100 microns thick) and is physically attached to the composite epilayer containing the betavoltaic device. In one embodiment, the betavoltaic epilayer is comprised of a III-V semiconductor with a betavoltaic semiconductor device structure. The betavoltaic structure may have any of the constructions or combinations described herein. For example, the betavoltaic epilayer may have a p/n or n/p structure with or without a highly doped base, a Cap layer to protect the device from oxidation. The betavoltaic epilayer with a betavoltaic device in its structure may be selectively etched/released from the III-V substrate via an intermediary sacrificial layer (e.g. AlGaAs, AlAs) as is known in the art. The sacrificial layer can have a thickness ranging from about 1 nm to about 200 nm. Once the sacrificial layer has been removed via etching, the epilayer and backing layer together are released. In doing this, the betavoltaic device thickness is reduced from standard semiconductor wafer thickness to less than 50 micron thickness. Furthermore, a cost reduction occurs due to the fact that the substrate may be reused to grow another epilayer, thereby reducing the cost of the base substrate material of the semiconductor device. Other methods of releasing the betavoltaic epilayer include etching the substrate with an etch-stop near the betavoltaic epilayer structure or wafer thinning using methods known in the art (e.g. grinding and polishing). It should be understood that any III-V direct conversion device may be formed into an epilayer through this liftoff process by utilizing a selective etch process to release the epilayer. Conversely, the epilayer may be released by etching away the substrate completely while leaving the epilayer intact. This may be accomplished by using a selective etch process that terminates or slows down near an etch-stop adjacent to the betavoltaic epilayer. Similarly, the substrate may be grinded and polished up to or near the betavoltaic epilayer. It should be noted that in both the case of releasing the betavoltaic epilayer via a selective etch of a sacrificial layer and when completely etching away the substrate, the substrate may be undoped/insulating. Typically, the substrate is doped in order to make a back contact to the betavoltaic cell, but in the case of a betavoltaic epilayer the contact may be directly made to the released betavoltaic epilayer either through a metal contact or a metal tritide acting as a contact that is connected to a doped back surface field or a doped buffer layer. The use of undoped/insulating substrates can offer a further cost reduction over doped counterpart substrates. The various embodiments of the present invention allow construction of a single flexible epilayer tritium betavoltaic battery or a very thin betavoltaic battery involving a stack of tritium betavoltaic epilayer cells stacked in either a series or parallel electrical configuration. For example a thin epilayer tritium betavoltaic battery may be constructed with either the tritium metal hydride film connected to the epilayer or directly deposited on the epilayer. A thin betavoltaic epilayer battery may be connected to a lithium ion thin film battery available from companies such as Front Edge Technologies of Baldwin Park, Calif., Cymbet Corporation of Elk River, Minn. or Infinite Power Solutions from Littleton, Colo. These two batteries may be connected together as a joint film that may be pasted within an integrated circuit package to run the device periodically via power bursts from the lithium thin film battery. The tritium epilayer betavoltaic battery can trickle charge the lithium ion film battery. Periodically the film battery can discharge power bursts at milliwatt power levels and then be recharged via the trickle charging by the tritium epilayer betavoltaic battery. The tritium epilayer battery, due to its thinness and flexibility, may be inserted into the conformal coating of an integrated circuit and power the integrated circuit stealthily. It can also be combined with a lithium ion thin film battery into the conformal coating of an integrated circuit as a source of power for the integrated circuit. The tritium epilayer battery can also be placed within an integrated circuit's package, multi-chip-module or printed circuit board. Another approach of the present invention involves texturing the tritium metal hydride substrate to increase the surface area of the deposited tritium. Using this method, the substrate is textured to produce surface roughness and then a suitable metal (e.g. palladium, titanium, scandium) for tritium capture into the metal hydride deposited on the surface. Texturing the tritium metal hydride substrate rather than the semiconductor in the betavoltaic device may avoid creating defects on the semiconductor's surface that result in a high dark current and poor efficiency. The textured tritium metal hydride source is then placed in direct contact with the smooth semiconductor device's surface resulting in a higher density of tritium beta flux entering the semiconductor device. Note the tritium may be deposited on the textured substrate via any means known in the art. Some examples include aerogels and polymers that may be deposited directly onto the textured substrate surface. In one embodiment, a silicon substrate's surface is textured using a potassium hydroxide (KOH) etchant as is known in the art for texturing silicon solar cells to prevent reflection of sunlight. In this embodiment, square-based pyramids with approximately 10 micron tall peaks, as measured from the base, are formed on the surface. The resulting surface area is 1.8 times the original planar surface. As mentioned elsewhere herein, a suitable metal is then deposited on the surface with a thickness of approximately 0.1 to 1 micron and then treated with tritium. The metal to be tritiated is then deposited using methods known in the art. This results in a tritium metal hydride with increased surface area on a stainless steel surface. In another embodiment, a metal substrate (e.g. stainless steel or titanium) surface is mechanically roughened. Conversely, periodic triangular rows are grated through the surface via a laser or other suitable method that can increase the surface area. Note, if triangular rows are formed where the triangles are equilateral in nature, the surface can reach twice the surface area of the original planar surface. A tritiated metal hydride is then formed on the substrate. Some of the most secure processors and field programmable gate arrays (FPGA's) are using SRAM memory to store encryption keys. However current battery technologies depend on chemistries that are unreliable over long periods of time (i.e. several years) especially under wide temperature ranges, such as −55° C. to +125° C. The tritium betavoltaic batteries of the present invention are able to power the SRAM memory for periods of 15-20 years or more through these extreme temperatures. Note, the voltage of tritium betavoltaic batteries based on III-V compounds will fluctuate less in higher temperatures than silicon-based betavoltaic devices. The tritium based betavoltaic batteries of this invention allow soldier-to-base wireless communications and computer-to-base communication to be encrypted using FPGA's with encryption keys stored in SRAM as well as defense and telecom applications that experience a wide range of temperatures. Note, the tritium betavoltaic batteries are hermetically sealed batteries packaged in surface mount packages that may be soldered to circuit board with the FPGA's Another application of tritium based betavoltaic batteries of the present invention is for supplying power to anti-tamper volume protection for electronics and other devices that require protection from intruders. These type of batteries provide the critical longevity of more than 10 years for anti-tamper protection. Note, the temperature resilience of these batteries is critical to the longevity and reliability. In one embodiment a volume protection membrane from W. L. Gore is used on a circuit card to protect encryption keys stored in SRAM from a reverse engineering attack. The tritium betavoltaic batteries of this present invention may be hermetically sealed in a surface mount package and soldered on the circuit board to provide power to both the volume protection device, the anti-tamper trigger in the processor and the encryption keys held in SRAM. If an attack occurs on the volume protection device (i.e., W. L. Gore volume protection membrane), the tritium betavoltaic battery power allows the volume protection device to detect the attack and the anti-tamper trigger will erase all critical information residing in the electronics, including the encryption keys. Various layers are described herein as having a p-type dopant or an n-type dopant. Those skilled in the art recognize that the dopant types can be reversed (n-type doped layers replaced with p-type and p-type doped layers replaced with n-type) and the device will provide the same functionality. Also, certain embodiments have been described as having an intrinsic layer; depending on the dopant types, doping levels, and other factors, this intrinsic layer may not be required in all embodiments. In the Tables the intrinsic layer is referred to as the “i-layer.” This written description uses examples to disclose the invention, including the best mode, and also to enable any person skilled in the art to make and use the invention. The scope of the invention may include other examples that occur to those skilled in the art. Such other examples are intended to be within the scope of the claims if they have structural elements that do not differ from the literal language of the claims, or if they include equivalent structural elements with insubstantial differences from the literal languages of the claims. In addition to the described embodiments and the layers comprising those embodiments, it should be noted that other embodiments of the invention may comprise one or more material layers from one of the described embodiments used with one or more material layers from other ones of the described embodiments.
060211709	abstract	A nuclear reactor plant includes a reactor pressure vessel having a core barrel supported on a cylindrical holding structure fixed in the reactor pressure vessel. The core barrel has a lower part with at least one elastically resilient segment. A projection disposed on the segment engages under the holding structure in such a way that a wedge-like gap is formed between a lower surface of the holding structure and the projection. A wedge is braced in the gap. A method is also provided for mounting the core barrel.
	claims	1. A method for fabricating a sintered duplex nuclear fuel pellet, comprising the steps of:preparing a first powder composed of a material selected from the group consisting of UO2 and UO2—Er2O3, and a second powder composed of UO2—Gd2O3 and a sintering additive;producing a duplex compact consisting of an annular outer portion composed of the first powder and a cylindrical inner portion composed of the second powder; andsintering the duplex compact under a reducing gas atmosphere,wherein the sintering additive is manganese oxide of from 0.001% to 2% by weight based on the total weight of the cylindrical inner portion, and the density of the sintered duplex nuclear fuel pellet is more than 94% of theoretical value,wherein the sintering additive is added only to the second powder. 2. The method according to claim 1, wherein the Er2O3 content in the mixture UO2—Er2O3 constituting the first powder is 4% by weight or less. 3. The method according to claim 1, wherein the Gd2O3 content in the mixture UO2—Gd2O3 constituting the second powder is in the range of 4% to 15% by weight. 4. The method according to claim 1, wherein the reducing gas is hydrogen gas, or a mixed gas of hydrogen gas with at least one gas selected from water vapor, inert gases and carbon dioxide. 5. The method according to claim 1, wherein the duplex compact is sintered in the temperature range of between 1,600° C. and 1800° C. 6. The method according to claim 1, wherein at least part of the UO2 constituting the annular outer portion and the cylindrical inner portion is replaced by PuO2.
048287601	summary	BACKGROUND OF THE INVENTION The present invention broadly relates to a method of cleaning a spent fuel assembly which has been removed from service in a nuclear reactor. It particularly relates to a method of cleaning such a spent fuel assembly contaminated with a radioactive alkali metal and in which the fuel assembly comprises a spent fissile material selected from the group consisting of oxides and carbides of uranium and/or plutonium or a metallic uranium/plutonium alloy. The typical breeder reactor utilizes an alkali metal as a coolant. In operation of the reactor a primary coolant is circulated via a pump through a core of fuel assemblies and then through a heat exchanger from which thermal energy is extracted by indirect heat exchange with a secondary coolant. The primary coolant picks up a certain amount of radioactivity in passing through the core. All components in contact with this primary coolant also become contaminated with the radioactive constituents. Thus, any component in contact with such coolant after removal from service must be cleaned, not only to remove the chemically highly reactive coolant but also the radioactive coolant. All fuel handling of spent core assemblies must be done remotely in an inert gas filled fuel handling cell or facility such as a fuel storage building or housing, until the spent core assemblies are cleaned of sodium. Thereafter, they may be handled remotely in an air filled fuel handling storage facility or stored under water. One method utilized for cleaning such components is to wash them with an organic solvent such as alcohol or wash them with water. The disadvantage of these methods is that they produce a large volume of liquid, low-level radioactive waste with its attendant storage and disposal problems. Another technique utilized is to heat the contaminated parts to a sufficient temperature to volatilize the alkali metal and the radioactive constituents. It also has been proposed to place such contaminated components in a chamber maintained under a vacuum to removal alkali metal vapors and radioactive constituents. A fuel assembly removed from a breeder reactor presents a particularly unique problem. For example, during service the cladding of some of the individual fuel pins comprising the fuel assembly could have been cracked or ruptured such that some of the alkali metal will be within the cladding of the individual pins. In addition, the spent fuel assembly, after removal from service, will continue to generate some heat generally referred to as decay heat. Thus, if the fuel assembly was placed in a vacuum the individual pins could generate sufficient heat to rupture the cladding of those pins which were still intact with the release of more radioactive material. Clearly, the use of external heating would be equally inapplicable. Obviously there is a need for a procedure uniquely adapted to the cleaning of a spent fuel assembly which has been in contact with an alkali metal coolant. OBJECTS OF THE INVENTION It is an object of the invention to provide a cleaning and decontamination method particularly adapted for cleaning spent fuel assemblies. It is another object of the invention to provide a method for cleaning such fuel assemblies which is both economical and efficient in removing contaminants. It is another object of the invention to provide such a cleaning method which can accommodate increases in temperature resulting from decay heat of the fuel assemblies. It is another object of the invention to provide a fuel assembly cleaning method which utilizes a combination of heat and vacuum. Still another object of the invention is to provide such a method which also produces an offgas stream free of alkali metal and radioactive contaminants such that the offgas may be reused for additional cleaning. These and other objects of the invention will be more apparent from the drawings and following detailed description. SUMMARY OF THE INVENTION The present invention provides a method of decontaminating or cleaning a fuel assembly contaminated with a radioactive alkali metal. Typically, a fuel assembly comprises a plurality of elongated pressurized metallic pins or tubes containing a spent fissile material under pressure. The present invention is particularly suitable for fuel assemblies containing a fissible material selected from the group consisting of carbides and oxides of uranium and/or plutonium and which are going to be reprocessed. The method comprises a plurality of sequential steps. First, the fuel assembly is placed in a sealed chamber. A heated inert gas is passed through the chamber to heat the fuel assembly to a temperature sufficient to cause evaporation of the alkali metal but not so high as to affect the structural integrity of the individual metal pins. Thereafter the chamber is evacuated to a pressure of about less than 0.05 and preferably less than 0.005 mm of mercury and maintained at that pressure until the temperature of the fuel assembly increases to a level which could affect the structural integrity of the metal pin, typically about 1000.degree. F. Af that time a cool, inert gas is introduced through the chamber and the fuel assemblies to reduce the temperature of the fuel assembly back to a safe level, typically about 800.degree. C. The vacuum and cooling steps are repeated as often as required to insure removal of substantially all of the radioactive alkali metal. The decontaminated fuel assemblies may then be removed from the chamber and are suitable for water storage or shipment to a chemical reprocessing center. In accordance with certain preferred embodiments of the invention the inert gas utilized is argon and the gases removed from the sealed chamber are passed in indirect heat exchange relationship with a coolant for the condensation and removal of any vaporized alkali metal contained therein. Typically, the coolant will be one which is inert with respect to the alkali metal, for example, alcohol or other organic liquids. In accordance with another preferred embodiment, the gas removed from the chamber when it is being evacuated is also passed through a cryogenic trap to insure substantially complete removal of any remaining radioactive alkali metal from the gas. Typically, the alkali metal will be either sodium or a mixture of sodium and potassium. The method of the present invention is particularly applicable for use in a fuel storage building which is maintained in an inert atmosphere and generally is located adjacent to the reactor from which the fuel assemblies are removed. In such application the inert gas within the facility may be used for the heating and cooling, and after treatment for removal of sodium or other alkali metal, may be introduced into an existing gas cleaning facility which is provided within such storage buildings.
048516945	summary	FIELD OF THE INVENTION The present invention relates to the field of radiotherapy, and more particularly to a device with which a source holder can be introduced into and positioned in an applicator tube. BACKGROUND OF THE INVENTION In treating certain diseases such as breast cancers, it is known to introduce one or more tubes of small diameter, known as "applicator tubes" or "applicators", into the diseased organ, one or more radioactive sources being introduced into these applicators and allowed to act for a variable duration, generally on the order of one to two days. Most often, there are about 20 applicators involved, the number typically being between 5 and 30. The elements most recently used as the sources are iridium, cobalt, cesium and californium. These sources can be contained inside a tube called a "source holder", and it is this source holder that is introduced into the applicator. The sources may be in the form of grains, each of them several millimeters in length, or in the form of filaments of greater length. Hence the source holder may contain either a single filament or a plurality of grains disposed one after another, and separated from one another by spacers. The active length, that is, the length occupied by the sources, is equal to 150 mm at most. The source holder is mounted in the anterior portion of a movable cable along a guide tube that terminates in the applicator. The problems that arise are not only that of driving the cable between an initial or storage position, in which the source holder is for instance inside a block of lead, to a final position or treatment position, in which the source holder is located inside the applicator, but also the problem of positioning the source holder, because it must be located at a very precise region inside the applicator, as well as the problem of detection of obstacles. It is essential that any obstacle encountered by the cable be detected, before it reaches its final position, and that the site of this obstacle be determined so that rapid intervention is possible. Several solutions have been proposed so far. French Pat. No. 2 033 653 describes a device used especially in plesioradiotherapy, in which the source holder tube is driven by a metal cable. Controlling the position of the posterior end of this cable, that is, the end opposite that carrying the sources, is done with the aid of two end-of-course detectors, which detect the passage of this end. There is no provision for detection of obstacles in this device, and hence the cable is in danger of being blocked before its posterior end has passed the end-of-course detector corresponding to the final position, and thus it can happen that the sources are not located inside the applicator. French Pat. No. 2 536 531 relates to a method of controlling the positioning of a source holder in a device of this generic type. Control is accomplished by analysis of the induced current of the motor driving the cable; the presence of an obstacle over the course causes the induced current of the motor to change, which enables the detection of the presence of this obstacle. A variation in the intensity of the current represents an increase in the load moment, and detecting this variation makes it possible to define a displacement time of the cable on which the source holder is mounted. This displacement time is compared with a reference time, which enables determining whether the cable is occupying an extreme position along its course, or is in an intermediate position. Although the presence of an obstacle can be detected with such a method, the exact position of the obstacle cannot be determined, because a certain amount of slippage, or sliding, can take place between the tube and the roller of the driving motor, and the exact length of the cable that has unwound is not known. For the same reason, it is no longer known with certainty that when the cable is in its final position, the source holder is indeed in its position inside the applicator. European Pat. No. 0 152 124 also describes a device for driving and positioning a source holder in an applicator used in radiotherapy. In this device, the source holder is actuated via a metal cable, and the end of the cable opposite the sources is wound in a spiral over a groove provided on one face of a wheel driven by a rack. Positioning of the source holder in the applicator is done with the aid of a pneumatic system. Compressed air is sent through an opening located at one end of the guide tube, in which the cable moves, and the air exits by an opening located at the end of the guide tube closest to the applicator. The cable has a head that comes into contact with a stop when the source holder is in its place in the applicator, which prevents the air from leaving. A monitoring apparatus detects the resultant increase in pressure inside the tube sheath and stops the driving of the cable. There is no provision for an obstacle detection system in this device. As a result, if the cable is stopped by an obstacle prior to reaching its final position, the air will continue to exit via the second opening in the guide tube, and the cable will continue to unwind. It is therefore impossible to know whether there is an obstacle, or to detect the position of such an obstacle. SUMMARY OF THE INVENTION It is the object of the invention to overcome these disadvantages by providing a device for driving and positioning a source holder in an applicator used in radiotherapy with which the source holder cannot only be easily displaced but also positioned very precisely in the applicator, and with which an obstacle can be rapidly detected and the site of the obstacle can be determined easily and precisely. More precisely, the subject of the present invention is a device for driving and positioning at least one source holder in an applicator tube used in radiotherapy, the applicator tube having a first end and a second end, the second end being open, the device comprising the following: a guide tube having one end capable of being connected to the second end of the applicator, a cable adapted to be displaced longitudinally along the guide tube and the applicator between a first position and a second position, the cable having a first end in the vicinity of which the source holder can be mounted, and a second end, and means for driving the cable along the guide tube and the applicator, characterized in that according to the invention, these drive means include: a motor-driven roller capable of being put into tangential contact, with sliding, with the cable, means for causing the rotation of this motor-driven roller, and an encoder equipped with a roller capable of being put into tangential contact, without sliding, with the cable, this encoder being adapted for detecting an interruption in the rotation of this roller. The words "device for driving and positioning at least one source holder in an applicator tube" are intended to mean that the invention applies to any number of applicator tubes. It can be used for placing a single source holder into an applicator, but typically, a plurality of applicators (on the order of 20) is used in radiotherapy. There is only a single source holder per applicator, and each source holder contains one or more sources. Preferably the encoder used is an incremental encoder; that is, it emits a signal each time the roller is displaced by rotation about a given angle, that is, practically each time that the cable is displaced by a predetermined length. Advantageously, a detector past which the source holder moves is mounted on the cable, in the course of the movement of the cable, can be provided, in order to count the sources. With this detector, it is possible not only to count the sources mounted on a given cable, but also to verify that the number of sources mounted on this cable is correct. In another feature of the invention, the device further includes a first end-of-course detector in front of which the second end of the cable is located when the cable is in its first position. Advantageously, the device may include a second end-of-course detector as well, located in such a manner as to detect the passage of the second end of the cable if the cable moves beyond the second position. Finally, a second guide tube in which at least a portion of the cable can be displaced, can also be provided. This second guide tube has one end closed by a stop, and the second end of the cable comes into contact with this end, if the cable arrives on the near side of the first position. The invention will become more apparent from the ensuing detailed description, which is given solely by way of illustrative example and in no way limits the scope of the invention, taken in conjunction with the accompanying drawings.
	abstract	A control rod operation generates a rod insertion block signal during operation of a reactor. Four neutron detector assemblies including a plurality of LPRMs arranged in an axial direction of a core are arranged adjacent to a plurality of insertion selection control rods, respectively, which are simultaneously inserted into the core. Neutron flux ratio calculation units are arranged in each of the neutron detector assemblies, and ratios (neutron flux ratios BA/AA, CA/AA, and DA/AA) of an average LPRM signal of the respective LPRMs at positions B, C, and D to an average LPRM signal of the respective LPRMs at a position A which is closest to the control rod insertion end of the core are calculated. When the largest neutron flux ratio out of the neutron flux ratios exceeds a set neutron flux ratio, a rod insertion block signal which is generated by a local range rod insertion monitor is output.
	description	The present invention relates to a clamper and an in-channel-head operation device, and particularly, to a clamper capable of stably clamping and holding a tube member and an in-channel-head operation device having the clamper. In a steam generator of a nuclear plant, an in-channel-head operation device is used to carry out an operation inside a water chamber. Further, in recent years, there has been proposed an in-channel-head operation device that has a manipulator of which a front end portion is equipped with a working tool. Further, since a floor surface of the water chamber has a spherical shape, there has been proposed a configuration in which such an indoor operation device is installed while being suspended from a tube sheet surface of the water chamber in order to improve the workability of the in-channel-head operation device. As such an in-channel-head operation device, a technique disclosed in Patent Literature 1 is known. Here, a plurality of heat transfer tubes are opened and arranged on the tube sheet surface of the water chamber. Therefore, in such an in-channel-head operation device, there has been proposed a configuration in which a clamper is provided so as to clamp and hold the heat transfer tube by inserting a front end portion thereof into the heat transfer tube and the in-channel-head operation device is fixed to the tube sheet surface by the clamper. As an in-channel-head operation device having such a clamper, a technique disclosed in Patent Literature 2 is known. Patent Literature 1: Unexamined Japanese Patent Application No. 2007-183278 Patent Literature 2: Japanese Utility Model Registration No. 2503172 In the in-channel-head operation device, a manipulator increases in weight, and various force and moment act on the clamp mechanism by the operation of the manipulator during an operation inside the water chamber. For this reason, there has been a demand for the clamper to stably clamp and hold the heat transfer tube. Therefore, the invention is made in view of such circumstances, and it is an object of the invention to provide a clamper capable of stably clamping and holding a tube member and an in-channel-head operation device having the clamper. According to an aspect of the present invention, a clamper includes: a clamp mechanism which inserts an insertion portion into a tube member and clamps the tube member; and a lifting and lowering mechanism which lifts and lowers the clamp mechanism in the insertion direction of the insertion portion. The clamp mechanism includes a clamp body which includes the insertion portion, a cotter which protrudes from the insertion portion so as to come into friction-contact with the tube member and is disposed so as to be movable in a reciprocating manner in the insertion direction of the insertion portion, a piston rod which presses the cotter so that the cotter protrudes from the insertion portion when being pulled toward the opposite side to the insertion direction of the insertion portion, and a rod cylinder which is integrated with the clamp body and pulls the piston rod. In the clamper, the lifting and lowering mechanism lifts the clamp mechanism so as to insert the insertion portion of the clamp mechanism into the tube member. Then, when the rod cylinder pulls the piston rod toward the opposite side to the insertion direction of the insertion portion (toward the rear end portion), the cotter is pressed by the piston rod so that the cotter protrudes from the insertion portion. Then, the cotter is pressed against the inner peripheral surface of the tube member so as to come into friction-contact therewith, thereby clamping the tube member. Accordingly, there is an advantage that the tube member may be stably clamped. Advantageously, in the clamper, the piston rod passes through the rod cylinder and protrudes toward a rear end portion of the rod cylinder. In the clamper, when the piston rod does not come off from the cotter by the fitting thereto upon unclamping the tube member, the piston rod may be separated from the cotter by beating the rear end portion of the piston rod from the outside. Accordingly, there is an advantage that the unclamping process may be carried out in an emergency state. Advantageously, the clamper, further includes: a pressurizing device which is connected through a tube to a fluid chamber that pulls the piston rod by the pressurization of a working fluid among fluid chambers of the rod cylinder and pressurizes the working fluid of the fluid chamber; a check valve which is disposed on the tube; and a release valve which discharges the working fluid of the fluid chamber. In the clamper, when the pressurizing device pressurizes the working fluid of the fluid chamber, the rod cylinder is lowered, and the cotter clamps and holds the tube member. At this time, the check valve prevents the reverse flow of the working fluid from the fluid chamber while the release valve is closed. Thus, even when the pressurizing device is stopped during an operation inside the water chamber (upon clamping the tube member), the fluid pressure of the fluid chamber is maintained, so that the state where the tube member is clamped is appropriately maintained. Accordingly, it is possible to realize the fail-safe when air leaks from the tube. Advantageously, in the clamper, the clamp mechanism includes a cotter diameter decreasing structure which decreases the diameter of the cotter. In the clamper, since the cotter may be accommodated inside the insertion portion by decreasing the diameter thereof upon unclamping the clamper, there is an advantage that the insertion portion may be easily extracted from the tube member. Advantageously, in the clamper, the cotter diameter decreasing structure includes a tapered surface which is formed on the cotter and an elastic member which biases the tapered surface to the clamp body and decreases the diameter of the cotter by sliding the cotter along the tapered surface by the biasing force from the elastic member. In the clamper, there is an advantage that the cotter diameter decreasing structure may be realized by a simple configuration with the tapered surface of the cotter and the elastic member. Advantageously, in the clamper, the cotter diameter decreasing structure includes an annular elastic member which is fitted to the outer periphery of the cotter. In the clamper, there is an advantage that the cotter diameter decreasing structure may be realized by a simple configuration with the annular elastic member. According to another aspect of the present invention, an in-channel-head operation device includes any one of above described clamper. In the clamper according to the invention, the lifting and lowering mechanism lifts the clamp mechanism so as to insert the insertion portion of the clamp mechanism into the tube member. Then, when the rod cylinder pulls the piston rod toward the opposite side to the insertion direction of the insertion portion (toward the rear end portion), the piston rod presses the cotter so that the cotter protrudes from the insertion portion. Then, the cotter is pressed against the inner peripheral surface of the tube member so as to come into friction-contact therewith, thereby clamping the tube member. Accordingly, there is an advantage that the tube member may be stably clamped. Hereinafter, the invention will be described in detail by referring to the drawings. Furthermore, the invention is not limited to the embodiment. Further, constituents of the embodiment include a constituent which may be replaced or apparently replaced while maintaining the identity of the invention. Further, a plurality of modified examples described in the embodiment may be arbitrarily combined with each other within the scope which may be apparently supposed by the person skilled in the art. [In-Channel-Head Operation Device] A clamper 23 is applied to, for example, an in-channel-head operation device 1 which carries out an operation inside a water chamber in a steam generator 130 of a nuclear plant (see FIG. 15). The in-channel-head operation device 1 is a device which is carried to a water chamber 131 of the steam generator 130, is installed therein, and is remotely operated so as to carry out an operation inside the water chamber. The in-channel-head operation device 1 includes a base 2, an intermediate link 3, a manipulator 4, and a tool 5. The base 2 is a device which becomes the base of the in-channel-head operation device 1, and is installed in a tube sheet surface 137a of the water chamber 131. The base 2 clamps and holds heat transfer tubes 132 of the tube sheet surface 137a so as to be fixed to the tube sheet surface 137a. The intermediate link 3 is a component which connects the base 2 to the manipulator 4 so as to incline a reference axis of the manipulator 4 with respect to the base 2 (the tube sheet surface 137a). The manipulator 4 is a multi-axis manipulator, and is installed so as to be hung from the tube sheet surface 137a of the water chamber 131 through the base 2 and the intermediate link 3. The manipulator 4 may change its posture by the remote operation. The tool 5 is a tool which corresponds to the operation inside the predetermined water chamber 131, and is attached to a front end portion of the manipulator 4. The tool 5 is, for example, a maintenance working tool which is used in the maintenance operation inside the water chamber, and includes an inspecting tool, a cutting tool, a welding tool, and the like. In the in-channel-head operation device 1, the manipulator 4 is hung while being suspended from the tube sheet surface 137a of the water chamber 131, and carries out an operation inside the water chamber by moving the tool 5 through a change in posture while being rotated in this state. Accordingly, it is possible to realize an operation inside the water chamber in a wide range based on the base 2 of the tube sheet surface 137a. Further, the in-channel-head operation device 1 may move inside the water chamber 131 along the tube sheet surface 137a since the base 2 has a tube sheet walking function to be described later. Accordingly, since the work region inside the water chamber 131 is widened, the workability of the operation inside the water chamber 131 is improved. Furthermore, the operation inside the water chamber 131 includes, for example, an operation of welding an inlet tube stand 135, an outlet tube stand 136, the heat transfer tubes 132, a partition plate 134, and a tube sheet 137, an operation of inspecting a welded portion between the partition plate 134 and a water chamber glass portion, a maintenance operation, and the like. [Base] FIGS. 16 and 17 are perspective views illustrating a specific example of the base of the in-channel-head operation device described in FIG. 15. FIGS. 18 and 19 are diagrams illustrating an installation state of the base described in FIG. 16. In these drawings, FIGS. 16 and 18 illustrate a state where the base opens all wings, and FIGS. 17 and 19 illustrate a state where the base closes all wings. The base 2 includes a base body 21, four wings 22a and 22b, and a plurality of clampers 23a and 23b. The base body 21 is a rim-shaped casing. The four wings 22a and 22b are installed by being inserted into the base body 21. The wings 22a and 22b are driven by, for example, a telescopic ladder mechanism, and may slide in a direction perpendicular to each other with respect to the installation position of the base body 21 (see FIGS. 16 and 17). Further, the four wings 22a and 22b may slide in different directions, and are independently driven. The clampers 23a and 23b are mechanisms which insert the front end portions thereof into the heat transfer tubes 132 so as to clamp and hold the heat transfer tube 132. For example, in the embodiment, a set having three clampers 23a (23b) is installed in each end portion of each wing 22a (22b). Further, the clampers 23a (23b) are arranged in series so as to match the installation interval of the heat transfer tubes 132 in the tube sheet surface 137a. Furthermore, a specific configuration of the clampers 23a and 23b will be described later. In the base 2, the respective clampers 23a and 23b insert the front end portions thereof into the heat transfer tubes 132 so as to clamp and hold the heat transfer tubes 132, and hence the base 2 is fixed while being suspended from the tube sheet surface 137a (see FIGS. 18 and 19). Further, the base 2 moves the wing 22a (22b) in a telescopic manner so as to slide the position of the clamper 23a (23b), and sequentially changes the clamping position of the clamper 23a (23b) with respect to the heat transfer tubes 132, so that the base may move along the tube sheet surface 137a (tube sheet walking) (not illustrated). Furthermore, the walking logic according to the tube sheet walking of such the base 2 may be arbitrarily adopted within the scope which may be apparently supposed by the person skilled in the art. FIG. 1 is an axial cross-sectional view illustrating the clamper according to the embodiment of the invention. Furthermore, in the embodiment, the side of the heat transfer tube 132 (the side of the tube sheet surface 137a of the water chamber 131) will be called the upper side or the front end portion side of the clamper and the opposite side (the floor surface side of the water chamber 131) will be called the lower side or the rear end portion side of the clamper. The clamper 23 is a mechanism which inserts the front end portion thereof into the heat transfer tube 132 so as to clamp the heat transfer tube 132, and may be adopted as, for example, the clampers 23a and 23b of the base 2. The clamper 23 includes a clamp mechanism 231 and a lifting and lowering mechanism 232. The clamp mechanism 231 is a mechanism which inserts a front end portion thereof into the heat transfer tube 132 so as to clamp the heat transfer tube 132 by the friction-contact, and includes a clamp body 2311, a cotter 2312, a piston rod 2313, an elastic body 2314, and a rod cylinder 2315. The clamp body 2311 constitutes a body of the clamp mechanism 231. The axial front end portion of the clamp body 2311 is provided with an insertion portion 2311a which may be inserted into the heat transfer tube 132. Further, the clamp body 2311 includes a contact portion 2311b which may contact the tube sheet surface 137a of the heat transfer tube 132 while inserting the insertion portion 2311a into the heat transfer tube 132. For example, in the embodiment, the clamp body 2311 is formed as an elongated cylindrical member, and one end portion thereof is provided with the insertion portion 2311a. Further, the base of the insertion portion 2311a of the clamp body 2311 is provided with the flange-shaped contact portion 2311b, and in a state where the insertion portion 2311a is inserted into the heat transfer tube 132, the contact portion 2311b comes into plane-contact with an opening edge portion 132a of the heat transfer tube 132. Further, the insertion portion 2311a of the clamp body 2311 is provided with slits 2311c which correspond to the number of the installed cotters 2312. The cotter 2312 is a member which is pressed into the inner peripheral surface of the heat transfer tube 132 so as to come into friction-contact with the surface when clamping the heat transfer tube 132, and is formed of, for example, metal or the like. The cotter 2312 is installed in the clamp body 2311, and is disposed so that the cotter protrudes from the insertion portion 2311a of the clamp body 2311 and to be accommodated in the insertion portion 2311a. Further, the cotter 2312 is disposed so as to be movable in a reciprocating manner in the axial direction of the clamp body 2311 with respect to the insertion portion 2311a. For example, in the embodiment, the slit 2311c is formed in the insertion portion 2311a of the clamp body 2311, and the cotter 2312 is inserted and disposed in the slit 2311c. Further, the cotter 2312 is disposed inside the slit 2311c so as to be movable in a reciprocating manner in the radial direction and the axial direction of the clamp body 2311. The piston rod 2313 is a rod which drives the cotter 2312, and is inserted into the clamp body 2311 so as to be movable in a reciprocating manner in the axial direction of the clamp body 2311. The piston rod 2313 includes a tapered front end portion, and is disposed so as to contact the cotter 2312 while a tapered surface 2313a faces the rear end portion of the clamp body 2311. Further, when the piston rod 2313 is pulled toward the rear end portion of the clamp body 2311, the tapered surface 2313a presses the inner peripheral surface of the cotter 2312 so as to protrude from the clamp body 2311. Further, the rear end portion of the piston rod 2313 is provided with a piston portion 2313b. The elastic body 2314 is a member which disposes the cotter 2312 on the upper portion of the slit 2311c, and is formed by, for example, a coil spring or a rubber tube. The elastic body 2314 is inserted into the clamp body 2311, and is supported by the clamp body 2311 so as to be biased to the rear end portion of the cotter 2312. The cotter 2312 is held in the upper portion of the slit 2311c by the biasing force of the elastic body 2314. The rod cylinder 2315 constitutes a piston and cylinder mechanism which uses the piston rod 2313 (piston portion 2313b) as a piston. The rod cylinder 2315 is integrally formed with the rear end portion of the clamp body 2311, and drives the piston rod 2313 so as to displace in a reciprocating manner in the axial direction of the clamp body 2311. Further, the rod cylinder 2315 receives a fluid pressure from an external fluid pressure mechanism (not illustrated). Then, the piston rod 2313 is driven by the control of the fluid pressure. For example, in the embodiment, the rod cylinder 2315 is divided into a first fluid chamber 2315a and a second fluid chamber 2315b through the piston portion 2313b of the piston rod 2313. Then, the piston rod 2313 is pulled toward the rear end portion of the clamp body 2311 by the pressurization of the working fluid of the first fluid chamber 2315a. Accordingly, the tapered surface 2313a of the piston rod 2313 pressurizes the cotter 2312, so that the cotter 2312 protrudes from the clamp body 2311. Further, the piston rod 2313 is pressed toward the front end portion of the clamp body 2311 by the pressurization of the working fluid of the second fluid chamber 2315b. Accordingly, the tapered surface 2313a of the piston rod 2313 releases the pressurization toward the cotter 2312, so that the cotter 2312 is accommodated in the clamp body 2311. The lifting and lowering mechanism 232 is a mechanism which lifts and lowers the clamp mechanism 231, and includes a lifting and lowering cylinder 2321. The lifting and lowering cylinder 2321 constitutes a piston and cylinder mechanism which uses the clamp body 2311 (piston portion 2311d) as a piston. Further, the lifting and lowering cylinder 2321 is directly connected to the rod cylinder 2315 of the clamp mechanism 231. Further, the lifting and lowering cylinder 2321 is fixed to the wing 22a (22b) of the base 2. Accordingly, the clamper 23 is fixed and held by the wing 22a (22b) of the base 2 through the lifting and lowering cylinder 2321. The lifting and lowering cylinder 2321 receives a fluid pressure from an external fluid pressure mechanism (not illustrated). Then, the clamp body 2311 as the piston displaces in a reciprocating manner by the control of the fluid pressure, so that the clamp mechanism 231 is lifted and lowered. For example, in the embodiment, the lifting and lowering cylinder 2321 is divided into a first fluid chamber 2321a and a second fluid chamber 2321b through the piston portion 2311d of the clamp body 2311. Then, the clamp body 2311 is pressed toward the front end portion by the pressurization of the working fluid of the first fluid chamber 2321a, so that the clamp body 2311 is lifted. Further, the clamp body 2311 is pulled toward the rear end portion by the pressurization of the working fluid of the second fluid chamber 2321b, so that the clamp body 2311 is lowered. [Process of Clamping Heat Transfer Tube] FIG. 20 is a diagram illustrating a process of installing the base 2. FIG. 2 is a diagram illustrating a clamping process of the clamper described in FIG. 1. In these drawings, FIG. 20 illustrates a process (a base installing process) in which the assembled structure of the base 2 and the intermediate link 3 is carried to the water chamber 131 and is installed in the tube sheet surface 137a, and FIG. 2 illustrates a state where the clamper 23 of the base 2 clamps and holds the heat transfer tube 132 in the base installing process. In the process of installing the base 2, a pair of base carrying attachment fixtures 11 is attached to the tube sheet surface 137a of the water chamber 131, and the assembled structure of the base 2 and the intermediate link 3 is carried from a manhole 138 into the water chamber 131 (see FIG. 20). Then, the base carrying attachment fixtures 11 hold the assembled structure so as to be hung to the tube sheet surface 137a through a wire or belt 12. Subsequently, the base 2 inserts the front end portions of the clampers 23a and 23b (the insertion portion 2311a of the clamp mechanism 231) into the heat transfer tube 132 so as to clamp and hold the heat transfer tube 132. In the process in which the clamper 23 clamps the heat transfer tube 132, as the initial state, the clamper 23 is first disposed so that the axial direction is aligned to the vertical direction while the front end portion (the insertion portion 2311a of the clamp mechanism 231) faces the heat transfer tube 132 (see FIG. 2(a)). Further, the clamper 23 is fixed to the wing 22a (22b) of the base 2 in the lifting and lowering cylinder 2321 of the lifting and lowering mechanism 232. Further, in a state where the piston rod 2313 is pressed toward the front end portion of the clamp body 2311, the cotter 2312 is accommodated inside the slit 2311c of the clamp body 2311 while decreasing in diameter. Further, in a state where the elastic body 2314 is lengthened, the cotter 2312 is pressed upward toward the front end portion inside the slit 2311c of the clamp body 2311. Further, in a state where the lifting and lowering mechanism 232 lowers the clamp mechanism 231, the insertion portion 2311a is positioned at the side near the heat transfer tube 132. Next, the lifting and lowering mechanism 232 pressurizes the working fluid of the first fluid chamber 2321a of the lifting and lowering cylinder 2321 so as to lift the clamp mechanism 231 (see FIG. 2(b)). Accordingly, the insertion portion 2311a of the clamp mechanism 231 is inserted into the heat transfer tube 132. At this time, since the lifting and lowering cylinder 2321 is held at the side of the base body 21 (the wing 22a (22b)) and the base body 21 is held while being hung by the wires 12 of the base carrying attachment fixtures 11 (see FIG. 20), the height position of the lifting and lowering cylinder 2321 is constantly maintained. Further, in this state, the contact portion 2311b of the clamp body 2311 and the opening edge portion 132a of the heat transfer tube 132 (the tube sheet surface 137a) normally comes into close contact with each other (a gap g may be generated depending on the conditions). Next, the clamp mechanism 231 pressurizes the working fluid of the first fluid chamber 2315a of the rod cylinder 2315 (see FIG. 2(c)). Then, the piston rod 2313 is pulled toward the rear end portion of the clamp body 2311 so as to be lowered, and the tapered surface 2313a presses and opens the cotter 2312 so as to increase the diameter of the cotter 2312. Then, the cotter 2312 protrudes from the slit 2311c of the clamp body 2311, and the head portion of the cotter 2312 is pressed against the inner peripheral surface of the heat transfer tube 132 so as to come into friction-contact therewith. Accordingly, the cotter 2312 clamps and holds the heat transfer tube 132. Next, in a state where the cotter 2312 comes into friction-contact with the heat transfer tube 132, the clamp mechanism 231 further pressurizes the working fluid of the first fluid chamber 2315a of the rod cylinder 2315 (see FIG. 2(d)). Then, since the piston rod 2313 is fixed to the heat transfer tube 132 through the cotter 2312, the clamp body 2311 pulls the piston rod 2313 toward the rear end portion, so that the entire clamper 23 is lifted upward. Accordingly, a gap g between the contact portion 2311b of the clamp body 2311 and the opening edge portion 132a of the heat transfer tube 132 is closed so that the contact portion 2311b of the clamp mechanism 231 is pressed against the opening edge portion 132a of the heat transfer tube 132 so as to come into plane-contact therewith. Specifically, the rod cylinder 2315, the clamp body 2311 integrated with the rod cylinder 2315, the lifting and lowering cylinder 2321 engaging with the clamp body 2311, and the base 2 connected to the lifting and lowering cylinder 2321 are all pulled by the fluid pressure of the rod cylinder 2315 so as to be lifted. Further, at this time, the position of the slit 2311c moves toward the front end portion of the clamp body 2311 with respect to the cotter 2312 by the axial displacement of the clamp body 2311. Further, the elastic body 2314 between the clamp body 2311 and the cotter 2312 is deformed in a compressed state. Then, the plurality of clampers 23 (23a and 23b) which are installed in the base 2 clamp and hold the heat transfer tube 132 as described above, so that the base 2 is fixed to the tube sheet surface 137a in a suspended state (see FIG. 20). At this time, since the respective clampers 23 clamp and hold the heat transfer tubes 132 by causing the contact portion 2311b to come into close contact with the opening edge portion 132a of the heat transfer tube 132 (the tube sheet surface 137a) (see FIG. 2(d)), the base 2 is properly fixed to the tube sheet surface 137a. Accordingly, the in-channel-head operation device 1 is stably installed in the tube sheet surface 137a during the operation inside the water chamber. [Process of Unclamping Heat Transfer Tube] FIG. 3 is a diagram illustrating a process of unclamping the clamper described in FIG. 1. The same drawing illustrates a state where the clampers 23a and 23b unclamp the heat transfer tubes 132 in a process in which the in-channel-head operation device 1 is removed from the water chamber 131 (a removing process). When the clamper 23 unclamps the heat transfer tube 132, in a state where the clamper 23 clamps and holds the heat transfer tube 132 (see FIG. 3(e)), the clamp mechanism 231 pressurizes the working fluid of the second fluid chamber 2315b of the rod cylinder 2315 and depressurizes the working fluid of the first fluid chamber 2315a thereof (see FIG. 3(f)). Then, the piston rod 2313 is pressed toward the front end portion of the clamp body 2311 and the cotter 2312 is accommodated inside the clamp body 2311 while decreasing in diameter. Further, the compression state of the elastic body 2314 is released, so that the elastic body 2314 is restored. Accordingly, the clamp body 2311 may be extracted from the heat transfer tube 132. Next, the lifting and lowering mechanism 232 pressurizes the working fluid of the second fluid chamber 2321b of the lifting and lowering cylinder 2321 and depressurizes the working fluid of the first fluid chamber 2321a so as to lower the clamp mechanism 231 (see FIG. 3(g)). Accordingly, the insertion portion 2311a of the clamp body 2311 is extracted from the heat transfer tube 132, so that the clamping and holding of the heat transfer tube 132 is released. Then, the plurality of clampers 23 (22a and 23b) installed in the base 2 release the clamping and holding of the heat transfer tube 132, so that the base 2 may be separated from the tube sheet surface 137a. Further, since the base 2 is held so as to be hung by the wire 12 of the base carrying attachment fixture 11 in this state, the falling of the base 2 is prevented (see FIG. 20). Furthermore, in the embodiment, in a state where the clamper 23 clamps and holds the heat transfer tube 132, the piston rod 2313 passes through the rod cylinder 2315 and protrudes toward the rear end portion of the rod cylinder 2315 (see FIG. 3(e)). In such a configuration, when the piston rod 2313 does not come off from the cotter 2312 due to the fitting thereto upon unclamping the heat transfer tube 132, the piston rod 2313 may be separated from the cotter 2312 by beating the rear end portion of the piston rod 2313 from the outside. Accordingly, it is possible to perform the unclamping process in an emergency state. [Fluid Pressure Control Mechanism of Rod Cylinder] FIG. 4 is a schematic diagram illustrating a specific example of the clamper described in FIG. 1. FIG. 5 is a diagram illustrating an operation of the clamper described in FIG. 4. These drawings illustrate a configuration (FIG. 4) and an effect (FIG. 5) of the fluid pressure control mechanism of the rod cylinder 2315. As illustrated in FIG. 4, in the clamper 23, the clamp mechanism 231 includes a fluid pressure control mechanism 2316 which controls the fluid pressure of the rod cylinder 2315. The fluid pressure control mechanism 2316 includes a pressurizing pump 2316a, a check valve 2316b, and a release valve 2316c, and these are connected to the rod cylinder 2315 through tubes 2316d to 2316f. The pressurizing pump 2316a is a pump which pressurizes a working fluid (for example, air) of the rod cylinder 2315. The pressurizing pump 2316a is connected to the first fluid chamber 2315a of the rod cylinder 2315 through the tube 2316d, and is connected to the second fluid chamber 2315b of the rod cylinder 2315 through the tube 2316e. Further, the pressurizing pump 2316a may selectively pressurize the working fluids of the first fluid chamber 2315a and the second fluid chamber 2315b of the rod cylinder 2315. The check valve 2316b is disposed on the tube 2316d near the first fluid chamber 2315a of the rod cylinder 2315 so as to prevent the reverse flow of the working fluid from the first fluid chamber 2315a. The release valve 2316c is a valve which releases the working fluid of the first fluid chamber 2315a of the rod cylinder 2315 to the outside, and is installed in the first fluid chamber 2315a. The release valve 2316c is connected to the pressurizing pump 2316a through the tube 2316f, and is driven by the pressurizing pump 2316a so as to be opened and closed. In the process of clamping the heat transfer tube 132, the pressurizing pump 2316a pressurizes the working fluid of the first fluid chamber 2315a of the rod cylinder 2315 (see FIG. 5(a)). At this time, the release valve 2316c is closed. Thus, the fluid pressure of the first fluid chamber 2315a increases, so that the piston rod 2313 is lowered. Accordingly, the cotter 2312 increases in diameter so as to clamp and hold the heat transfer tube 132 (see FIG. 5(b)). At this time, the check valve 2316b prevents the reverse flow of the working fluid from the first fluid chamber 2315a. Thus, even when the pressurizing pump 2316a is stopped, the fluid pressure of the first fluid chamber 2315a is maintained, so that the state where the heat transfer tube 132 is clamped is appropriately maintained. By the check valve 2316b, it is possible to realize the fail-safe for maintaining the state where the heat transfer tube 132 is clamped, for example, when air leaks from the tube 2316d which connects the pressurizing pump 2316a to the first fluid chamber 2315a during the operation inside the water chamber. In the process of unclamping the heat transfer tube 132, the pressurizing pump 2316a pressurizes the working fluid of the second fluid chamber 2315b and depressurizes the working fluid of the first fluid chamber 2315a by opening the release valve 2316c (see FIG. 5(c)). At the same time, the fluid pressure of the second fluid chamber 2315b increases, so that the piston rod 2313 is lifted. Accordingly, the cotter 2312 decreases in diameter, so that the state where the heat transfer tube 132 is clamped is released (see FIG. 5(d)). Furthermore, in the embodiment, air is used as the working fluid of the rod cylinder 2315. For this reason, the working fluid of the first fluid chamber 2315a is directly discharged from the release valve 2316c into the water chamber 131. [Cotter Diameter Decreasing Structure] FIGS. 6 and 7 are diagrams illustrating a cotter diameter decreasing structure of the clamper described in FIG. 1. These drawings illustrate a structure which decreases the diameter of the cotter 2312 of the clamp mechanism 231 in the process of unclamping the clamper 23. In the clamper 23 illustrated in FIG. 6, the cotter 2312 includes a tapered surface 2312a, and the tapered surface 2312a is disposed toward the front end portion of the clamp body 2311 inside the slit 2311c of the clamp body 2311. Further, the elastic body 2314 is biased to the rear end portion of the cotter 2312 while being supported by the clamp body 2311. By the biasing force of the elastic body 2314, the cotter 2312 presses the tapered surface 2312a against the wall surface near the front end portion of the slit 2311c. In a state where the heat transfer tube 132 is clamped (see FIG. 3(e)), the piston rod 2313 is pulled so as to be lowered, and protrudes from the slit 2311c of the clamp body 2311 while increasing the diameter of the cotter 2312 (see FIG. 6). At this time, the elastic body 2314 is compressed. Next, in the process of unclamping the heat transfer tube 132, when the piston rod 2313 is lifted (see FIG. 3(f)), the elastic body 2314 presses the cotter 2312 against the wall surface near the front end portion of the slit 2311c (see FIG. 6). Accordingly, the cotter 2312 decreases in diameter while sliding inside the slit 2311c along the tapered surface 2312a (not illustrated). Accordingly, the cotter 2312 is accommodated inside the clamp body 2311. In the clamper 23 illustrated in FIG. 7, the cotter 2312 includes an annular elastic member 2312b. The elastic member 2312b is formed by, for example, an annular plate spring and a rubber-like O-ring, and is fitted to the outer peripheral surface of the cotter 2312 (the outside in the radial direction), so that an elastic force is applied to the cotter 2312 inward in the radial direction. In a state where the heat transfer tube 132 is clamped (see FIG. 3(e)), the piston rod 2313 is pulled so as to be lowered, and protrudes from the slit 2311c of the clamp body 2311 while increasing the diameter of the cotter 2312 (see FIG. 7). At this time, the elastic member 2312b is lengthened by an increase in the diameter of the cotter 2312. Next, in the process of unclamping the heat transfer tube 132, when the piston rod 2313 is lifted (see FIG. 3(f)), the elastic member 2312b is shortened and the cotter 2312 decreases in diameter (not illustrated). Accordingly, the cotter 2312 is accommodated inside the clamp body 2311. FIGS. 8 to 10 are diagrams illustrating a first modified example of the cotter diameter decreasing structure described in FIG. 6. These drawings respectively indicate a main enlarged view of the cotter diameter decreasing structure (see FIG. 8), an operation diagram in the clamping process (see FIG. 9), and an operation diagram in the unclamping process (see FIG. 10). The clamper 23 illustrated in FIG. 8 has the following difference compared to the clamper 23 illustrated in FIG. 6. That is, the insertion portion 2311a of the clamp body 2311 is separated from the clamp body 2311. Specifically, the insertion portion 2311a is formed by a tube member having the slit 2311c, and is disposed so as to be movable in a reciprocating manner in the axial direction while being inserted into the front end portion of the clamp body 2311. Further, the cotter 2312 is inserted into the slit 2311c of the insertion portion 2311a, and is disposed inside the slit 2311c so as to be movable in a reciprocating manner in the axial direction and the radial direction of the insertion portion 2311a. Further, the cotter 2312 includes the tapered surface 2312a, and the tapered surface 2312a is disposed so as to face the front end portion of the clamp body 2311. Further, the rear end portion of the cotter 2312 is provided with a support member 2317 which supports the cotter 2312. The support member 2317 is a tube member which has a flange-shaped front end portion, and is disposed so that the front end portion thereof contacts the rear end portion of the cotter 2312. Further, the support member 2317 is inserted into the insertion portion 2311a, and is disposed inside the slit 2311c so as to be movable in a reciprocating manner in the axial direction of the insertion portion 2311a. Further, an elastic body 2318 is disposed so as to be interposed between the rear end portion of the support member 2317 and the clamp body 2311. The elastic body 2318 is formed by, for example, a coil spring or a rubber tube, and is biased to the rear end portion of the support member 2317 while being supported by the clamp body 2311. By the biasing force of the elastic body 2318, the support member 2317 is held while pressing the cotter 2312 against the wall surface near the front end portion of the slit 2311c, and the insertion portion 2311a is held so as to protrude from the front end portion of the clamp body 2311 in the axial direction. In the clamping process, the lifting and lowering mechanism 232 lifts the clamp mechanism 231, so that the clamp mechanism 231 inserts the insertion portion 2311a into the heat transfer tube 132 (see FIGS. 9(a) and 2(b)). At this time, the contact portion 2311b of the clamp body 2311 and the opening edge portion 132a of the heat transfer tube 132 (the tube sheet surface 137a) come into close contact with each other (a gap g may be formed depending on the conditions). Next, the clamp mechanism 231 pulls the piston rod 2313 toward the rear end portion so as to be lowered (see FIGS. 9(b) and 2(c)). Then, the tapered surface 2313a of the piston rod 2313 presses and opens the cotter 2312, so that the cotter 2312 increases in diameter. Then, the cotter 2312 protrudes from the slit 2311c of the insertion portion 2311a, and the head portion of the cotter 2312 is pressed against the inner peripheral surface of the heat transfer tube 132 so as to come into friction-contact therewith. At this time, the support member 2317 displaces toward the rear end portion of the clamp body 2311 inside the slit 2311c, so that the cotter 2312 may displace while sliding along the tapered surface 2312a. Next, in a state where the cotter 2312 comes into friction-contact with the heat transfer tube 132, the clamp mechanism 231 further pulls the piston rod 2313 toward the rear end portion (see FIGS. 9(c) and 2(d)). Then, since the piston rod 2313 is fixed to the heat transfer tube 132 through the cotter 2312, the entire clamper 23 is lifted upward. At this time, since the clamp body 2311 and the insertion portion 2311a are separated from each other, the clamp body 2311 displaces in a sliding manner toward the front end portion in the axial direction with respect to the insertion portion 2311a. Accordingly, a gap g between the contact portion 2311b of the clamp body 2311 and the opening edge portion 132a of the heat transfer tube 132 is closed so that the contact portion 2311b is pressed against the opening edge portion 132a of the heat transfer tube 132 so as to come into plane-contact therewith. Further, the elastic body 2318 between the clamp body 2311 and the support member 2317 is deformed in a compressed state by the axial displacement of the clamp body 2311. In the unclamping process, when the piston rod 2313 is lifted while the clamper 23 clamps the heat transfer tube 132, the pressing force from the piston rod 2313 to the cotter 2312 is released (see FIGS. 10(d) and 10(e)). Then, since the support member 2317 presses the cotter 2312 against the clamp body 2311 (the wall surface near the front end portion of the slit 2311c) by the biasing force of the elastic body 2318, the cotter 2312 decreases in diameter while sliding inside the slit 2311c along the tapered surface 2312a. Accordingly, the cotter 2312 is accommodated inside the clamp body 2311. FIGS. 11 to 13 are diagrams illustrating a second modified example of the cotter diameter decreasing structure described in FIG. 7. These drawings respectively indicate a main enlarged diagram of the cotter diameter decreasing structure (FIG. 11), an operation diagram in the clamping process (FIG. 12), and an operation diagram in the unclamping process (FIG. 13). The clamper 23 illustrated in FIG. 11 has the following difference compared to the clamper 23 illustrated in FIG. 7. That is, the insertion portion 2311a of the clamp body 2311 is separated from the clamp body 2311. Specifically, the insertion portion 2311a is formed by a tube member having the slit 2311c, and is disposed so as to be movable in a reciprocating manner in the axial direction while being inserted into the front end portion of the clamp body 2311. Further, the elastic body 2318 is disposed so as to be interposed between the insertion portion 2311a and the clamp body 2311. Since the elastic body 2318 is biased to the insertion portion 2311a, the insertion portion 2311a is held so as to protrude from the front end portion of the clamp body 2311 in the axial direction. Further, the cotter 2312 is inserted into the slit 2311c of the insertion portion 2311a, and is disposed inside the slit 2311c so as to be movable in a reciprocating manner only in the radial direction of the insertion portion 2311a. Then, the annular elastic member 2312b which decreases the diameter of the cotter 2312 is fitted to the outer peripheral surface of the cotter 2312 (the outside in the radial direction). In the clamping process, the lifting and lowering mechanism 232 lifts the clamp mechanism 231, so that the clamp mechanism 231 inserts the insertion portion 2311a into the heat transfer tube 132 (see FIGS. 12(a) and 2(b)). At this time, the contact portion 2311b of the clamp body 2311 and the opening edge portion 132a of the heat transfer tube 132 (the tube sheet surface 137a) come into close contact with each other (a gap g may be generated depending on the conditions). Next, the clamp mechanism 231 pulls the piston rod 2313 toward the rear end portion so as to be lowered (see FIGS. 12(b) and 2(c)). Then, the tapered surface 2313a of the piston rod 2313 presses and opens the cotter 2312, so that the cotter 2312 increases in diameter. Then, the cotter 2312 protrudes from the slit 2311c of the insertion portion 2311a, so that the head portion of the cotter 2312 is pressed against the inner peripheral surface of the heat transfer tube 132 so as to come into friction-contact therewith. Next, in a state where the cotter 2312 comes into friction-contact with the heat transfer tube 132, the clamp mechanism 231 further pulls the piston rod 2313 toward the rear end portion (see FIG. 12(c)). Then, since the piston rod 2313 is fixed to the heat transfer tube 132 through the cotter 2312, the entire clamper 23 is lifted upward. Further, since the clamp body 2311 is separated from the insertion portion 2311a, the clamp body 2311 displaces in a sliding manner with respect to the insertion portion 2311a. Accordingly, a gap g between the contact portion 2311b of the clamp body 2311 and the opening edge portion 132a of the heat transfer tube 132 is closed, so that the contact portion 2311b is pressed against the opening edge portion 132a of the heat transfer tube 132 so as to come into plane-contact therewith. Further, the elastic body 2318 between the clamp body 2311 and the insertion portion 2311a is deformed in a compressed state by the axial displacement of the clamp body 2311. In the unclamping process, when the piston rod 2313 is lifted while the clamper 23 clamps the heat transfer tube 132, the pressing force from the piston rod 2313 to the cotter 2312 is released (see FIGS. 13(d) and 13(e)). Then, the elastic member 2312b is shortened so as to decrease the diameter of the cotter 2312 (not illustrated). Accordingly, the cotter 2312 is accommodated inside the clamp body 2311. Furthermore, in the second modified example of the cotter diameter decreasing structure, the elastic body 2318 is disposed so as to be interposed between the rear end portion of the insertion portion 2311a and the inner bottom portion of the clamp body 2311 (see FIG. 11). However, the invention is not limited thereto, and the elastic body 2318 may be disposed so as to be interposed between a cotter brace 2319 supporting the rear end portion of the cotter 2312 and the inner bottom portion of the clamp body 2311 (see FIG. 14). In the case of such a configuration, the biasing force of the elastic body 2318 presses the front end portion of the slit 2311c through the cotter brace 2319 and the cotter 2312, and hence the insertion portion 2311a may protrude from the clamp body 2311, whereby the same function may be obtained. [Effect] As described above, the clamper 23 includes the clamp mechanism 231 which inserts the insertion portion 2311a into the tube member 132 (the heat transfer tube 132) and clamps the tube member 132 and the lifting and lowering mechanism 232 which lifts and lowers the clamp mechanism 231 in the insertion direction of the insertion portion 2311a (see FIGS. 1 and 2). Further, the clamp mechanism 231 includes the clamp body 2311 which has the insertion portion 2311a, the cotter 2312 which protrudes from the insertion portion 2311a so as to come into friction-contact with the tube member 132 and to be movable in a reciprocating manner in the insertion direction of the insertion portion 2311a, the piston rod 2313 which presses the cotter 2312 so that the cotter protrudes from the insertion portion 2311a when being pulled toward the opposite side to the insertion direction of the insertion portion 2311a, and the rod cylinder 2315 which is integrated with the clamp body 2311 and pulls the piston rod 2313. In such a configuration, the lifting and lowering mechanism 232 lifts the clamp mechanism 231 so that the insertion portion 2311a of the clamp mechanism 231 is inserted into the tube member 132 (see FIG. 2). Then, when the rod cylinder 2315 pulls the piston rod 2313 toward the opposite side to the insertion direction of the insertion portion 2311a (toward the rear end portion), the piston rod 2313 presses the cotter 2312 so that the cotter protrudes from the insertion portion 2311a. Then, the cotter 2312 is pressed against the inner peripheral surface of the tube member 132 so as to come into friction-contact therewith, thereby clamping the tube member 132. Accordingly, there is an advantage that the tube member 132 may be stably clamped. Further, in the clamper 23, the piston rod 2313 passes through the rod cylinder 2315 and protrudes toward the rear end portion of the rod cylinder 2315 (see FIG. 1). In such a configuration, when the piston rod 2313 does not come off from the cotter 2312 due to the fitting thereto upon unclamping the tube member 132, the piston rod 2313 may be separated from the cotter 2312 by beating the rear end portion of the piston rod 2313 from the outside. Accordingly, there is an advantage that the unclamping process may be performed in an emergency state. Further, the clamper 23 includes the pressurizing device (the pressurizing pump 2316a) which is connected through the tube 2316d to the fluid chamber (the first fluid chamber 2315a) that pulls the piston rod 2313 by the pressurization of the working fluid among the fluid chambers of the rod cylinder 2315 (the first fluid chamber 2315a and the second fluid chamber 2315b) and which pressurizes the working fluid of the fluid chamber 2315a, the check valve 2316b which is disposed on the tube 2316d, and the release valve 2316c which discharges the working fluid of the fluid chamber 2315a (see FIG. 4). In such a configuration, the pressurizing device 2316a pressurizes the working fluid of the fluid chamber 2315a, so that the rod cylinder 2315 is lowered and the cotter 2312 clamps and holds the heat transfer tube 132 (see FIGS. 5(a) and 5(b)). At this time, the check valve 2316b prevents the reverse flow of the working fluid from the fluid chamber 2315a in a state where the release valve 2316c is closed. Thus, even when the pressurizing device 2316a is stopped during the operation inside the water chamber (when clamping the heat transfer tube 132), the fluid pressure of the fluid chamber 2315a is maintained, so that the state where the heat transfer tube 132 is clamped is appropriately maintained. Accordingly, it is possible to realize the fail-safe when the original pressure of the working fluid of the clamper 23 is interrupted. Further, in the clamper 23, the clamp mechanism 231 includes a cotter diameter decreasing structure which decreases the diameter of the cotter 2312 (see FIGS. 6 to 8 and FIG. 11). In such a configuration, since the cotter 2312 may be accommodated inside the insertion portion 2311a while decreasing in diameter when unclamping the clamper 23, there is an advantage that the insertion portion 2311a may be easily extracted from the tube member 132. Further, in the clamper 23, the cotter diameter decreasing structure includes the tapered surface 2312a which is formed in the cotter 2312 and the elastic body 2318 which presses the tapered surface 2312a against the wall surface near the front end portion of the slit 2311c so as to bias the cotter 2312 to the clamp body 2311, and also decreases the diameter of the cotter 2312 by sliding the cotter along the tapered surface 2312a by the biasing force from the elastic body 2318 (see FIGS. 6 and 8). In such a configuration, there is an advantage that the cotter diameter decreasing structure may be realized by a simple configuration with the tapered surface 2312a of the cotter 2312 and the elastic body 2318. Further, in the clamper 23, the cotter diameter decreasing structure is formed by the annular elastic member 2312b which is fitted to the outer periphery of the cotter 2312 (see FIGS. 7 and 11). In such a configuration, there is an advantage that the cotter diameter decreasing structure may be realized by a simple configuration with the annular elastic member 2312b. As described above, the clamper according to the invention is useful in that the tube member may be stably clamped. 1 IN-CHANNEL-HEAD OPERATION DEVICE 2 BASE 3 INTERMEDIATE LINK 4 MANIPULATOR 5 TOOL 11 BASE CARRYING ATTACHMENT FIXTURE 12 WIRE OR BELT 21 BASE BODY 22a, 22b WING 23, 23a, 23b CLAMPER 130 STEAM GENERATOR 131 WATER CHAMBER 132 HEAT TRANSFER TUBE (TUBE MEMBER) 132a OPENING EDGE PORTION 134 PARTITION PLATE 135 INLET TUBE STAND 136 OUTLET TUBE STAND 137 TUBE SHEET 137a TUBE SHEET SURFACE 138 MANHOLE 231 CLAMP MECHANISM 2311 CLAMP BODY 2311a INSERTION PORTION 2311b CONTACT PORTION 2311c SLIT 2311d PISTON PORTION 2312 COTTER 2312a TAPERED SURFACE 2312b ELASTIC MEMBER 2313 PISTON ROD 2313a TAPERED SURFACE 2313b PISTON PORTION 2314 ELASTIC BODY 2315 ROD CYLINDER 2315a FIRST FLUID CHAMBER 2315b SECOND FLUID CHAMBER 2316 FLUID PRESSURE CONTROL MECHANISM 2316a PRESSURIZING PUMP 2316b CHECK VALVE 2316c RELEASE VALVE 2316d to 2316f TUBE 2317 SUPPORT MEMBER 2318 ELASTIC BODY 2319 COTTER BRACE 232 LIFTING AND LOWERING MECHANISM 2321 LIFTING AND LOWERING CYLINDER 2321a FIRST FLUID CHAMBER 2321b SECOND FLUID CHAMBER
	description	The present invention relates to a strip for a nuclear fuel assembly spacer grid comprising interlaced strips defining a lattice of cells for receiving fuel rods and allowing flow of a coolant in a flow direction. U.S. Pat. No. 4,879,090 illustrates on FIG. 5 thereof a peripheral strip for a nuclear fuel assembly spacer grid, the peripheral strip comprising wall portions to delimit cells and on each wall portion a spring formed by a tab cut out in the strip and motion limiters formed as a pair of bosses embossed in the strip at a distance from the tab. In operation, a coolant fluid (e.g. water) flows axially upwardly through the cells of the spacer grid. The spring and the motion limiters provided on each wall portion protrude from the plane of the wall portion towards the center of the same cell delimited by the wall portion and partially obstruct the coolant fluid flow channel. An object of the invention is to provide a strip for a nuclear fuel assembly spacer grid limiting the flow resistance of the spacer grid whilst allowing suitable support for the nuclear fuel rods during the whole fuel assembly lifetime and good manufacturability. To this end, the a strip for a nuclear fuel assembly spacer grid for a nuclear fuel assembly spacer grid is provided comprising interlaced strips defining a lattice of cells for receiving fuel rods and allowing flow of a coolant in a flow direction, the strip being of the type comprising a wall portion for delimiting a cell, a spring formed in the strip and provided on the wall portion for biasing the fuel rod extending through the cell away from the wall portion, the spring comprising a cantilevered tab formed in the strip and a contact portion formed at least partially in the tab and protruding from the tab for contacting the fuel rod received in the cell. In other embodiments, the strip comprises one or several of the following features, taken in isolation or in any technically feasible combination: in a free state of the spring, the tab is inclined relative to the wall portion; the contact portion is formed exclusively in the tab; the contact portion is formed in the tab and in the wall portion, the contact portion stepping over a connection line between the tab and the wall portion; the contact portion is in the form of an arched bridge; the contact portion has an end connected to the tab and an end connected to the wall portion; the contact portion is a bulge formed in the tab; the contact portion is of elliptical perimeter; the contact portion is elongated in the flow direction; the tab extends downwardly in cantilevered fashion towards an upstream lower edge of the strip; the tab is delimited by an elongated curved slot and a connection line joining the two opposite ends of the slot; ends of the slot are circular and enlarged; the slot is U-shaped with branches diverging one from the other; a motion limiter formed in the strip on the wall portion to limit motion of the fuel rod received in the cell towards the wall portion against action of the spring; the motion limiter is provided upstream or downstream the spring in the coolant flow direction through the cell delimited by the wall portion; the motion limiter is a bulge formed in the wall portion; the motion limiter is elongated, namely of elliptical perimeter. A spacer grid is also provided comprising interlaced strips defining a lattice of cells for receiving fuel rods and allowing flow of a coolant axially upwardly through the spacer grid, at least one of the interlaced strips being a strip as defined above. A nuclear fuel assembly is provided comprising a bundle of fuel rods and an armature for supporting the fuel rods, the armature comprising at least one spacer grid as defined above. The nuclear fuel assembly 2 for a pressurized water reactor (PWR) illustrated on FIG. 1 comprises a bundle of nuclear fuel rods 4 and an armature 6 for supporting the fuel rods 4. The PWR fuel assembly 2 is elongated along an assembly axis L extending vertically when the fuel assembly 2 is disposed inside a nuclear reactor. The armature 6 comprises a lower nozzle 8, an upper nozzle 10, a plurality of guide-tubes 12 and a plurality of spacer grids 14. The guide-tubes 12 extend parallel to assembly axis L and connect the lower nozzle 8 to the upper nozzle 10 and maintain a predetermined spacing along assembly axis L between the nozzles 8, 10. Each guide-tube 12 opens upwards through the upper nozzle 10 for allowing insertion of a control rod into the guide-tube 12. The nuclear fuel assembly 2 for a boiling water reactor (BWR) illustrated on FIG. 2 is also elongated along an assembly axis L extending vertically when the fuel assembly 2 is disposed inside a nuclear reactor. The BWR fuel assembly 2 comprises a bundle of nuclear fuel rods 4, an armature for maintaining the fuel rods 4 and a tubular fuel channel 15 surrounding the bundle of fuel rods 4. The armature typically comprises a lower nozzle and an upper nozzle spaced along assembly axis L, at least one water channel 13 arranged within the bundle of fuel rods 4 and a plurality of spacer grids 14 distributed along the bundle of fuel rods 4. The fuel rods 4, the water channel 13 and the fuel channel 15 extend between the lower nozzle and the upper nozzle, with the water channel 13 and the fuel channel 15 connecting the lower nozzle and the upper nozzle. The water channel 13 extends parallel to the fuel rods 4. The water channel 13 is arranged for channeling a coolant/moderator flow separately from the bundle of fuel rods 4. The fuel channel 15 extends parallel to the fuel rods 4. The fuel channel 15 encases the bundle of fuel rods 4 and the water channel 13. The fuel channel 15 is arranged for channeling a coolant/moderator flow between and about the fuel rods 4. The PWR and BWR spacer grids 14 are distributed in spaced relationship along the fuel rods 4. Each spacer grid 14 extends transversely to the assembly axis L. Each fuel rod 4 comprises a tubular cladding, pellets of nuclear fuel stacked inside the cladding and caps closing the ends of the cladding. Each fuel rod 4 extends parallel to assembly axis L through the spacer grids 14 with being supported transversely and longitudinally relative to assembly axis L by the spacer grids 14. In operation, the fuel assembly 2 is placed in a nuclear reactor with the lower nozzle 8 resting on a bottom plate of the reactor and the assembly axis L being substantially vertical. A coolant flows upwardly along the fuel assembly 2 with flowing between the fuel rods 4 and through the nozzles 8, 10 and the spacer grids 14 as illustrated by arrows F on FIGS. 1 and 2. The spacer grids 14 may be similar to each other and one spacer grid 14 according to the invention will be further described with reference to FIGS. 3-7. As illustrated on FIG. 3, the spacer grid 14 comprises a plurality of interlaced metallic strips 16 defining a lattice of cells 18 each for receiving one fuel rod 4, only a few cells 18 being illustrated on FIG. 3. In a known manner, in the case of a spacer grid for a PWR fuel assembly, the interlaced strips 16 also define a plurality of cells for receiving PWR guide-tubes 12, the spacer grid 14 being secured to the guide-tubes 12, e.g. by welding. Similarly, in the case of a spacer grid for a BWR fuel assembly, the at least one BWR water channel 13 typically replaces one or several fuel rods 4 in the lattice, the interlaced strips define an aperture for receiving the water channel 13 and the spacer grid 14 is secured to the water channel 13, e.g. by welding. Only cells 18 for receiving fuel rods 4 are illustrated on FIG. 3 and in the following, the term “cell” refer to the cells 18 for receiving fuel rods 4. Each cell 18 is tubular and extends along a cell axis A. The cell axis A is to be parallel to the assembly axis L (perpendicular to FIG. 3) when the spacer grid 14 is assembled in the fuel assembly 2 (FIGS. 1 and 2). The cell axes A of the different cells 18 are parallel. Each cell 18 is delimited by four wall portions 20 of two pairs of intersecting strips 16, the strips 16 of each pair extending parallel to one another. One wall portion 20 of each pair of opposite wall portions 20 delimiting a cell 18 has an elastic spring 22 formed in the wall portion 20 and protruding in a free state towards the center of the cell 18, and the other wall portion 20 of each pair of opposite wall portions 20 has a rigid dimple 24 formed in the wall portion 20 and protruding towards the center of the cell 18. The springs 22 and dimples 24 provided on the wall portions 20 of each cell 18 are arranged such that a fuel rod 4 extending through the cell 18 is biased transversely by the springs 22 against the dimples 24 to support the fuel rod 4 transversely and longitudinally relative to the cell axis A. Each wall portion 20 delimiting two adjacent cells 18 (one on each side of the strip 16) has a spring 22 protruding on a face of the wall portion 20 in one of the cells 18 and a dimple 24 protruding on the opposite face of the wall portion 20 in the other cell 18. Each wall portion 20 delimiting only one cell 18 has either a spring 22 or a dimple 24. FIG. 4 illustrates a plurality of the wall portions 20 of a strip 16, each of these wall portions 20 being adapted to delimit two cells 18, one on each side of the strip 16. In operation, the coolant flows upwardly through each cell 18 in the flow direction F represented on FIG. 4 from an upstream lower edge 26 to a downstream upper edge 28 of the strip 16. The flow direction F is parallel to the cell axis A. Each wall portion 20 extends from the lower edge 26 to the upper edge 28. The wall portions 20 are separated by slits 30 provided on the lower edge 26 and extending substantially to the half-height of the strip 16 for engagement with a series of corresponding slits 30 provided on the upper edge 28 and extending substantially to the half-height of an intersecting strip 16. The strip 16 optionally comprises fins 32 protruding upwardly from the upper edge 28, each fin 32 being inclined relative to the cell axis A for imparting helical motion to the coolant fluid flowing through the cells 18 and enhancing heat exchange between the coolant and the fuel rods 4. The strip 16 comprises on each of the illustrated wall portions 20 a spring 22, a dimple 24 and a motion limiter 34 each formed in the strip 16 and thus being integrally one-piece with the strip 16. The spring 22 and the motion limiter 34 provided on each wall portion 20 protrude on the same face of the strip 16, whereas the dimple 24 protrudes on the opposite face of the strip 16. The dimples 24 are alternately disposed below and above the springs 22 on the adjacent wall portions 20. Interlaced strips 16 thus can be arranged such that a spring 22 provided on a wall portion 20 of a strip 16 delimiting a cell 18 faces a dimple 24 provided on the opposite wall portion 20 of another strip 16 delimiting the cell 18. The springs 22 of the strip 16 are identical and one spring 22 is further described with reference to FIGS. 5-7. The spring 22 illustrated on FIG. 5 comprises a flexible cantilevered tab 36 and a contact portion 38 cut out in the strip 16. The tab 36 is delimited in the strip 16 by an elongated curved slot 40 of closed contour. The tab 36 is delimited between the slot 40 and the connection line 46 joining the opposed ends 48 of the slot 40. The ends 48 are preferably circular and enlarged to limit mechanical local peak stresses. The line 46 is perpendicular to the cell axis A. The tab 36 is connected to the wall portion 20 along the line 46. The tab 36 extends downwardly in cantilevered fashion towards the upstream lower edge 26 and has an upper base 42 connected to the wall portion 20 and a lower free tip 44. The tab 36 is converging towards the free tip 44. The slot 40 is generally U-shape with diverging branches (or V-shape with a rounded tip). In a free state of the spring 22, the tab 36 is inclined relative to the wall portion 20 and extends downwardly and away from the wall portion 20 towards the center of the cell 18 delimited by the wall portion 20. The tab 36 is elastically flexible by elastic deformation of the tab 36 with rotation of the tab 36 around a rotation axis substantially coinciding with the line 46. The flexibility of the tab can be adjusted by adjusting the diameter of the ends 48 of the slot 40. The contact portion 38 is formed exclusively in the tab 36 and protrudes from the tab 36 opposite the wall portion 20 and towards the center of a cell 18 delimited by the wall portion 20. The contact portion 38 is integrally one-piece with the tab 36. According to the embodiment illustrated on FIGS. 4-6, the contact portion 38 is provided in the form of an arched bridge cut out in the tab 36. The contact portion 38 is elongated in the direction of the cell axis A, the two ends of the contact portion 38 connected to the tab 36 being aligned in the direction of the cell axis A. The contact portion 38 is formed as a lancing delimited between two openings 52 extending substantially parallel to each other in the direction of the cell axis A. The motion limiter 34 associated to the spring 22 is formed in the strip 16 at a distance from the spring 22 and protrudes from the wall portion 20 on the same side than the corresponding spring 22. The motion limiter 34 is a bulge of curved perimeter formed in the strip 16, for instance of circular perimeter as illustrated on FIG. 5 with a risen portion 56 and an apex 62. The motion limiter 34 is disposed below or above the corresponding spring 22 (FIG. 3). As illustrated on FIG. 7, the contact portion 38 contacts the outer surface of a fuel rod 4 extending through a cell 18 delimited by the wall portion 20 with the tab 36 being elastically deformed towards the wall portion 20. The spring 22 thus biases the fuel rod 4 away from the wall portion 20 (toward the right on FIG. 7) in contact with a dimple 24 provided on the opposite wall portion 20 delimiting the cell 18. In this configuration, the free tip 44 of the tab 36 extends substantially in the plane of the wall portion 20 and the height H of the contact portion 38 relative to the wall portion 20 is superior to the height h of the apex 62 of the motion limiter 34. There is a gap D between the apex 62 and the outer surface of the fuel rod 4. In operation, the coolant flows through the cell 18 and around the fuel rod 4 upwardly at high speed in the flow direction F parallel to the cell axis A. This causes transverse vibration of the fuel rod 4 inside the cell 18. Transverse vibrations may also occur during transportation from manufacturing plant to power plant and during handling of the fuel assembly 2. The motion limiter 34 is rigid and limits movements of a fuel rod 4 towards the wall portion 20 against the action of the spring 22. The motion limiter 34 thus avoids overstress of the spring 22 and namely plastic deformation thereof. The spring 22 formed in the strip 16 with a flexible cantilevered tab 36 and a rigid contact portion 38 enables to bias the fuel rod 4 with an appropriate transverse force while limiting the flow resistance. The tab 36 furnishes the biasing force when the free tip 44 of the tab 36 is retracted in the plane (or nearly) of the wall portion 20; in this position only the contact portion 38 protrudes from the wall portion 20. The contact portion 38 being elongated in the flow direction F enables to further limit the flow resistance and to provide an elongated contact zone with the fuel rod 4 for limiting fretting risks. The spring 22 including the tab 36 and the contact portion 38 is obtainable in a single stamping operation easily and at low cost. The alternative embodiment of FIGS. 8 and 9 differs from that of FIGS. 5 and 6 by the feature that the contact portion 38 is formed partially in the tab 36 and partially in the wall portion 20. The contact portion 38 is more elongated and steps over the line 46 joining the ends 48 of the slot 40 delimiting the tab 36. This increases the stiffness of the spring 22 as biasing the spring 22 causes simultaneous deformation of the upper base 42 of the tab 36 and of the upper end of the contact portion 38 about two parallel but different axes. In the alternative embodiment of FIG. 10, the motion limiter 34 is a bulge having an elliptical perimeter. The motion limiter 34 is elongated transversely to the flow direction F. Each motion limiter 34 is located above the corresponding spring 22 in line with the spring 22 in the flow direction. Each motion limiter 34 is located adjacent the base of the spring tab 36. Preferably, the motion limiter 34 is tangent to the connection line 46. The contact portion 38 is provided in the form of a closed bulge of curved perimeter. As illustrated on FIG. 10, the perimeter is advantageously an elliptical perimeter preferably elongated in the flow direction F. The dimples 24 are also provided in the form of bulges of elliptical perimeter elongated in the flow direction. In an alternative embodiment, the motion limiter 34 of FIG. 10 is combined with a spring 22 having a bridge-like contact portion 38 cut in the spring as in FIGS. 4-6. In another alternative embodiment, the motion limiter 34 of circular perimeter of FIGS. 4-6 is combined with the contact portion of FIG. 10 in the form of a closed bulge, namely of elliptical perimeter. In the different embodiments, the motion limiter 34 and the spring 22 provided in line, one upstream the other in the flow direction F, reduce flow resistance of the strip 16. The motion limiter 34 located above the spring 22 adjacent the connection line 46 stiffens the spring element 22 and lead to less local peak stresses at the ends 48 of the slot 40. These effects are even increased with the motion limiter 34 of elliptical perimeter of FIG. 10. In the alternative embodiment of FIG. 11, the upstream lower edge 26 of the strip 16 is zigzag-shaped such that it is low at the center of each wall portion 20 and high at the junction between the wall portions 20 where interlaced strips 16 intersect each other. As a result, a spacer grid 14 may be formed with interlaced strips 16 crossing at cross points 66 at a level higher than the lower points 64, whereby debris possibly present in the coolant fluid are guided transversely towards the cross points 66 at corners of the square shaped cells 18 where the space between the inner surface of the cells 18 and the fuel rods 4 is larger. The debris are thus prevented from damaging the fuel rods 4. In an alternative embodiment, the lower edge 26 of the strip 16 is zigzag-shaped such that the upstream lower edge 26 is alternatively at a high level and at a low level at the junction between the wall portions 20. As a result, the interlaced strips 16 may be assembled to provide cross points 66 at a high level and cross points 66 at a low level arranged in staggered rows, with the same benefit. The lower edge 26 may present a wave shape instead of a zigzag shape. The invention is applicable to spacer grids for a PWR (Pressurized Water Reactor) fuel assembly or to spacer grids for a BWR (Boiling Water Reactor) fuel assembly as illustrated and also to spacer grids for a VVER (Water-Water Energetic Reactor) fuel assembly.
050874110	claims	1. Device for underwater recovery and elimination of radioactive waste, in a well, said device comprising (a) means (4) for aspiration and delivery of said water from said well; (b) a column (8) suspended in vertical position from a handling and lifting means (3) situated above said well; (c) an aspiration head (10) attached to an end of said column (8) remote from said handling means (3) and consisting of first and second hollow walls (32, 33) assembled to one another; (d) a filter (39) pierced by openings separating an internal volume of said aspiration head (10) into a first part comprising an aspiration opening (42) communicating with an exterior and a second part connected via a conduit (5) to said aspiration and delivery means (4); and (e) remotely demountable connecting means (45) connecting said first and second hollow walls (32, 22) of said aspiration head (10). 2. Device according to claim 1, wherein said hollow walls (32, 33) are of frustoconical shape and assembled to one another along major bases of said walls by means of flanges (34, 35). 3. Device according to claim 2, wherein said flanges (34, 35) respectively comprise pins (37) and centering openings (44) in corresponding locations. 4. Device according to claim 1, wherein said filter (39) is constituted by metallic wires disposed in accordance with a pattern, and wherein it is supported, within said aspiration head (10), on a filter support (38) comprising a network of openings of large dimensions (52). 5. Device according to claim 1, wherein said hollow walls (32, 33) are connected to one another by flexible connecting means (40), enabling relative displacement of said walls to enable extraction of said filter (39) from said aspiration head (10). 6. Device according to claim 1, wherein said means for connecting said hollow walls (32, 33), enabling said walls to be remotely demounted, comprise at least two eyelet screws (46) pivotably mounted on one of said hollow walls (33) and each having a tightening nut (49) threaded thereon, an opening (50) in said second hollow wall (32) in a position corresponding to a position of a said eyelet screw (46) for introduction of said eyelet screw by pivoting into said opening (50) and screwing of said nut (49) into contact with a part of said second hollow wall (32). 7. Device according to claim 1, comprising means (56) for ejection of said filter (39), said ejection means comprising a cylinder (57) solid with one of said hollow walls (32), a rod (58) mounted for sliding movement within said cylinder (57) and terminating in a pusher (59), a thrust means (61) and a return spring (62) disposed around said rod (58) and intercalated between one end of said cylinder (57) and said thrust means (61). 8. Device according to claim 1, comprising an aspiration nozzle (42) fixed on said second hollow wall (33) at a location of said opening placing said second part of the internal volume of said aspiration head in communication with an exterior of said aspiration head (10). 9. Device according to claim 8, wherein said nozzle (42) comprises at least one internal recess (43) at an end of said nozzle. 10. Device according to claim 1, wherein said column (8) comprises a plurality of successive sections (8a, 8b, 8c) in an axial direction of said column, said sections being connected by removable assembly means (12, 14, 15). 11. Device according to claim 10, wherein said column comprises three successive sections (8a, 8b, 8c), in a vertical direction, two upper sections (8a, 8b) each consisting of a tube of light alloy sealingly closed at ends thereof, and a lower section (8c) being open to be filled with water from said well. 12. Device according to claim 10, wherein each one of said sections (8a, 8b, 8c) comprises, at one end thereof, means for attaching a handling bar (9, 20).
042773636	claims	1. A method for processing a mixture of air and rare gases, at least some portion of which is radioactive, as gaseous effluents arising from the reprocessing of irradiated fuels, wherein said method comprises concentrating the rare gases in solution in liquid argon, as a liquefied mixture, by cryogenic distillation of light gases, a major portion of which is nitrogen, from the liquefied mixture. 2. A method according to claim 1, wherein said method comprises a cryogenic distillation from the concentrated and liquefied mixture to recover xenon. 3. A method according to claim 1, wherein the concentration of the rare gases is carried out at a maximum pressure of 18 bar. 4. A method for processing effluent gases arising from the reprocessing of irradiated nuclear fuels and comprising a mixture of gases containing at least radioactive krypton and xenon in air, wherein said method successively comprises (1) removing any impurities from said mixture, (2) concentrating xenon and krypton in solution in liquid argon as a liquefied mixture by distillation of light gases, a major portion of which is nitrogen, from the liquefied mixture, (3) liquefying the mixture of argon, xenon and krypton separated in step (2) and removing argon therefrom by cryogenic distillation thereof to obtain a mixture remaining of xenon and krypton, and (4) liquefying the mixture of xenon and krypton obtained in step (3) and separating the xenon from the krypton by cryogenic distillation thereof. 5. A method according to claim 4, wherein part of the krypton which has been separated from the xenon and obtained during step (4) is introduced into the mixture obtained at the end of step (1) before proceeding to the concentration of the xenon and krypton in solution in liquid argon in step (2). 6. A method according to claim 4, wherein step (1) comprising removing any impurities comprises an initial stage of removal of hydrocarbons by catalytic oxidation, a second stage of removal of oxygen and nitrogen oxides by catalytic reduction, and a third stage of removal of carbon dioxide gas and of water vapor by passage through a molecular sieve and/or silica gel. 7. A method according to claim 4, wherein the separation of xenon and krypton in step (4) is carried out at a pressure of approximately 3 bar. 8. A method according to claim 4, wherein the removing of argon from the mixture of argon, xenon and krypton in step (3) is carried out at a pressure between the pressure at which step (4) is conducted and the pressure at which step (2) is conducted.
	abstract	A vessel for the shipment of uranium hexafluoride includes a cylindrical wall closed by pair of approximately semi-ellipsoidal heads welded to form a sealed container. A service valve is located in one end. The valve is covered by a removable, watertight valve protection cover assembly. The vessel also includes a test port by means of which the integrity of the valve protection cover assembly may be tested after the cylinder has been filled with uranium hexafluoride and the valve protection assembly has been installed. The valve protection assembly is shaped so that it fits within the envelope of the standard 30B cylinders. The distal end of the valve protection assembly is recessed from a plane defined by the open end of the surrounding chime by at least one half inch, and preferably by xc2xe inch or more. Accordingly, the fits within an overpack already approved by the NRC and used by shippers of uranium hexafluoride.
	abstract	A storing device includes a radioactive material container, a radioactive material container outer shield and a ring member. The container includes a lower cup portion, an upper cup portion which is securely engaged to the lower cup portion, and a chamber with a cushion member mounted in a top and a bottom end thereof, which is formed inside the container for storing radioactive material. The outer shield includes a base portion and a lid which is securely covered on the base portion, the base portion further including a room to accommodate the radioactive material container. The ring member is pivotedly mounted to a top end of radioactive material container, and is rested between the lid and the upper cup portion. When the lid is opened, the ring member will stand out of the base portion for a tool to conveniently move out the container from the outer shield.
	description	The present application claims priority from Japanese patent application serial No. 2010-024007 filed on Feb. 5, 2010, the content of which is hereby incorporated by reference into this application 1. Field of the Invention The present invention relates to a steel concrete structure and a reactor containment facility employing it. 2. Description of the Related Art In an advanced boiling water reactor (ABWR), a reinforced concrete containment vessel (RCCV) is employed. The RCCV has such a structure that airtightness is retained by lining the inner surface with liners made of steel thereby leakage of radioactive material to the outside in an accident is prevented, and the load of an earthquake, pressure and the like presumed in the design is endured by reinforced concrete. On the other hand, in recent years, reactor containment vessels employing a steel concrete structure with the strength and anti-leakage function of the same level of those of the RCCV have been proposed (refer to JP-A-2008-275368, JP-A-2008-249329, JP-A-2007-204953, and JP-A-H9-41489). In the reinforced concrete structure, the work schedule tended to be prolonged by on-site works such as setting of reinforcing bars, a formwork for placing concrete, placing concrete and the like. On the other hand, in the steel concrete structure, concrete is filled inside a space sealed by steel plates. Also, the steel plates are fixed with the concrete through a plurality of studs arranged inside the steel plates to compositely bear a load, and therefore the reinforcing bars are not necessary. Furthermore, because the steel plates also function as the forms in placing concrete, the setting work of the reinforcing bars and the formwork on site are not required, and the construction period can be shortened. As a problem of the steel concrete structure, deformation by thermal load when the inside of the containment vessel is abnormal can be cited. Due to the sharp thermal expansion of the steel plate, difference of the thermal elongation occurs against the concrete and the steel plate possibly buckles. In addition, the moisture contained in the concrete may evaporate due to the heat when abnormality occurs, the steam inside the sealing steel plates may be concentrated between the rear surface of the heated steel plates and the concrete, the back pressure may be applied to the rear surface of the steel plates, and the steel plates may be bent beyond the initially installed flat plane to be antiplane-deformed causing drop of the buckling strength against the compressive stress. Further, because the concrete was filled between the two steel plates, the condition of the concrete could not be confirmed visually after placing the concrete. Furthermore, the purpose of arranging the studs in the steel plates is to secure the strength after placing the concrete, and the studs do not function as strengthening members before placing the concrete. The present invention aims to discharge the steam generated by evaporation of the water content of the concrete to the outside of the steel plates and to prevent or inhibit the back pressure from applying to the steel plates, aims to visually confirm the condition of the surface of the concrete even during the service period, and aims to secure the strength of the steel plates even before placing the concrete. The present invention is characterized in that, in a steel concrete structure for a reactor containment facility including a pair of steel plates composed of an inner steel plate exposed to a high temperature environment and an outer steel plate mutually opposingly arranged at an interval, a plurality of studs arranged on a surface of each steel plate in touch with concrete, and the concrete filled between the steel plates, a discharging opening for discharging steam is provided in the outer steel plate. Also the steel concrete structure for a reactor containment facility is characterized in that a steam discharging pipe in touch with the rear surface of the inner steel plate and leading steam to outside the concrete through a pipe opening arranged in the outer steel plate is provided, and a gap is provided between the pipe opening and the steam discharging pipe. Also the steel concrete structure for a reactor containment facility is characterized in that increased thickness parts are provided in the vicinity of the discharging opening and pipe opening. Also the steel concrete structure for a reactor containment facility is characterized in that vertical partitions connecting the opposing inner steel plate and outer steel plate are provided, horizontal reinforcements for preventing buckling are provided in the inner steel plate and the outer steel plate, and openings for letting concrete flow in are provided in the horizontal reinforcements. Also the steel concrete structure for a reactor containment facility is characterized in that the horizontal reinforcements are formed of horizontal partitions connecting the opposing inner steel plate and outer steel plate, and openings for letting concrete flow in are provided in the horizontal partitions. Also, the present invention is characterized by including a reactor containment vessel having the steel concrete structure with the configuration described above. Also, the present invention is characterized by including a reactor containment vessel internal structure employing the steel concrete structure with the configuration described above. The steel concrete structure according to an aspect of the present invention allows to prevent or inhibit antiplane deformation of steel plates due to the back pressure by providing a discharging opening for discharging steam in the outer steel plate in the steel concrete structure for a reactor containment facility including a pair of steel plates composed of an inner steel plate exposed to a high temperature environment and an outer steel plate mutually opposingly arranged at an interval, a plurality of studs arranged on a surface of each steel plate in touch with concrete, and the concrete filled between the steel plates. Further, observation of the concrete during the service period is also allowed. Embodiments according to the present invention will be described below referring to the drawings. FIG. 1 is a vertical cross-sectional view showing a part of the steel concrete structure of Example 1. In the steel concrete structure for a containment vessel including an inner steel plate 1, an outer steel plate 2, an upper steel plate 20, a lower steel plate 21, studs 3 and concrete 4, a discharging opening 8 is provided. The discharging opening 8 has the function of discharging the steam inside the concrete, and allows observation of the condition of the surface of the concrete during the service period. With the constitution, the steam inside the concrete is gradually discharged to the outside of the sealed steel concrete structure through the discharging opening 8, thus antiplane deformation of the inner steel plate 1 is avoided in particular. FIG. 2 is a perspective view showing a part of the steel concrete structure. In order to explain the internal construction, the concrete is not shown in the figure. FIG. 3 is a vertical cross-sectional view of the steel concrete structure shown in FIG. 1. FIG. 2 and FIG. 3 show the steel concrete structure including the inner steel plate 1, the outer steel plate 2, the upper steel plate 20, the lower steel plate 21, a rear steel plate 22, the studs 3 and the concrete 4 provided with a steam discharging pipe 6 and a pipe opening 5 through which piping penetrates. The steam discharging pipe 6 includes small openings 7 at its tip for letting the steam flow in. The steam discharging pipe 6 is joined to the inner steel plate in advance when the steel plate is to be installed, and the concrete is filled thereafter between the steel plates 1 and 2. When the concrete is filled, in order to prevent the concrete from flowing into the steam discharging pipe 6 through the small openings 7, a small openings closing fixture formed of a small diameter pipe and the like that can be inserted to and pulled out from the steam discharging pipe 6 is used. When the temperature of the inner steel plate 1 rises, the water content inside the concrete evaporates in a state the temperature of the concrete reaches approximately 100° C., becomes steam, is concentrated in the vicinity of the rear surface of the steel plate 1, and is discharged to the outside of the containment vessel from the small openings 7 through the steam discharging pipe 6. FIG. 4 is a vertical cross-sectional view of the steel concrete structure similar to that shown in FIG. 3, and shows that the increased thickness parts 9 are arranged around the pipe opening 5 and the discharging opening 8 and the steel plate thickness is increased. The increase of the plate thickness aims to reinforce the strength of the outer steel plate missed due to the openings. FIG. 5 is a perspective view showing a part of the steel concrete structure having the horizontal reinforcements. The structure includes the steel plates 1 and 2, the studs 3, the concrete 4, the pipe opening 5, horizontal reinforcements 10, and vertical partitions 12. For simplicity, the concrete and the vertical partitions are not shown in the figure. FIG. 6 is a transverse cross-sectional view of the steel concrete structure shown in FIG. 5, and the reinforcements 10 are provided with openings 11 to improve workability when the concrete is placed. The horizontal reinforcements 10 function as stiffener rings for the steel plates, and inhibit buckling of the steel plates before and after placing the concrete. FIG. 7 is a perspective view showing a part of the steel concrete structure having the horizontal partitions. The structure includes the steel plates 1 and 2, the studs 3, the concrete 4, the pipe opening 5, the horizontal reinforcements 10, and the vertical partitions 12. For simplicity, the concrete and the vertical partitions are not shown in the figure. FIG. 8 is a plan view of the steel concrete structure shown in FIG. 7, and horizontal partitions 13 are provided with openings 14 to improve workability when the concrete is placed. The openings 11 and 14 can be of arbitrary shapes considering workability irrespective of the shape illustrated in the figure. The horizontal partitions 13 function as the stiffener rings for the steel plates, and inhibit buckling of the steel plates before and after placing the concrete. The steel concrete structures of the Examples 1-5 are normally assembled into a cylindrical shape to be utilized as a reactor containment vessel, however they also may be of a vertical flat plane structure of a disk-shape as shown in FIG. 9 and FIG. 10. FIG. 9 shows a case in which the partitions are arranged radially and the concrete is placed inside, whereas FIG. 10 shows a case in which the partitions are arranged in a lattice-like formation and the concrete is placed inside. In this case, with respect to a pair of vertically arranged steel plates 15, partition 16 function as horizontal partitions arranged orthogonal thereto, whereas partition 17 function as vertical partitions. The concrete placed in the space formed by them is not shown in the figure. These steel concrete structures can be used for a reactor containment vessel requiring pressure resistant and anti-leakage functions. Also, they can be used for a reactor containment vessel internal structure as well.
	description	This application is a U.S. Divisional application of U.S. patent application Ser. No. 12/051,087, filed Mar. 19, 2008. The contents of this application are incorporated herein by reference. The invention relates to an improved method for irradiating a target with a beam of energetic electrically charged particles, wherein a pattern definition means having a plurality of apertures transparent to said particles is used, and by illuminating said pattern definition means with said beam, which traverses the pattern definition means through said apertures, a patterned beam is formed consisting of a corresponding plurality of beamlets, and said patterned beam is formed into an image on the location of the target, said image comprising the images of at least part of the plurality of apertures; the aperture images have a first width, wherein said target moves relative to the pattern definition means along a path within an image plane in which the images of apertures are formed (or in a direction generally perpendicular to a propagation direction of the patterned beam). A method of this kind and the pertinent particle-beam apparatus is disclosed in the U.S. Pat. No. 6,768,125, which is hereby incorporated into the present disclosure as relevant prior art. That patent describes a charged-particle lithography and processing method and apparatus dubbed PML2 (short for “Projection Mask-Less Lithography”) which realizes a multi-beam direct write concept and uses a programmable aperture plate system (APS) as a pattern definition (PD) device for structuring a particle beam, which is extracted from a single source of electrically charged particles. Particle lithography and processing is used in semiconductor production and micro-structuring applications. In particular direct patterning by ion-beam irradiation is a promising concept for the future industrial fabrication of nano-scale devices with high resolutions, in multi-aperture device. The implementation of a multi-beam projection optical system based on a programmable multi-aperture plate allows a significant improvement of the achievable productivity in comparison with focused single beam systems. (In this disclosure ‘optical’ always is meant as ‘particle-optical’.) The reasons for the improved productivity are, firstly, the parallelism of the process using a plurality of beams and, secondly, the increased current which can be imaged to a substrate at the same resolution. Both are made possible by a significantly reduced Coulomb interaction in the beam. Furthermore, the moderate current density related to the projection optical system results in an enhanced process rate when precursor gases are used for beam-induced chemical processes. As compared with a focused beam system, also the reduced heating effect due to extreme beam intensity can be reduced or avoided. In the PML2 layout as disclosed in the U.S. Pat. No. 6,768,125 and related prior art the position of the patterned beam on the target is held fixed on the corresponding pixel positions only for the duration one pixel of the target takes to travel under the optical system. Then the patterned beam jumps to the position of the next pixel which is adjacent to the previous one. In this manner, each beamlet covers adjacent pixels on the target. This strategy requires a rapid loading rate of the pattern information, which may cause an upper limit for the processing speed of the method. The diagram of FIG. 19 illustrates this prior-art writing strategy, for the example of one line of pixels to be exposed on a target such as a resist-covered semiconductor wafer. The pixels of this one line are irradiated (exposed) by a number of beamlets (respectively denoted by capital letters A, B, C, . . . ), and each beamlet irradiates a sequence of adjacent pixels during successive time steps. The amount of exposure dose imparted to each pixel is symbolized by hatched and cross-hatched fields. The cross-hatched fields denote pixels that are completely illuminated (according to a programmed pattern; subsequent exposures are depicted symbolically stacked on top of each other while in reality they simply add up), whereas simply hatched fields indicate pixels in the course of exposure. The dose for every pixel is successively summed up while the resist moves below the PD device forming the beamlets (see the discussion of maskless multi-beam writing discussed below in relation to FIG. 1). Each beamlet has the same time interval in which it contributes a partial exposure dose to the substrate, and consequently contributes the same amount of dose. Every beamlet in a line is used to expose every pixel in that line. In order to switch on and off the beamlets through the apertures in the PD device in accordance with this exposure method, the PD device apertures require the pattern information at a corresponding data rate, and the image information has to be distributed to every single aperture on the APS. The first two frames of FIG. 19 depict the situation at the beginning and the end of one time interval for the exposure of one pixel (per beamlet), denoted t=0:0 and 0:1, respectively. It is clear from the depiction that the beamlets are moved along with the target so their positions are locked on the (moving) position of the pixels during the pixel exposure. When the exposure of the pixel is finished after t=0:1, all beamlet are repositioned so as to lock on the position of the next pixels, respectively; the third frame t=1:0 shows the beginning of the next pixel exposure. The repositioning of the beamlets occurs in a very short time, much shorter than the complete time lag Tw between consecutive pixel exposures. (Tw is equal to the time between frames 0:0 and 1:0.) The mutual distance of beamlets is 5w, i.e., five times the width of the beamlets; but with the prior-art method other integer values are equally well suitable as discussed in the U.S. Pat. No. 6,768,125. The diagram of FIG. 20 illustrates the timing of the prior art writing process of FIG. 19. At the top the beam deflection x is shown as a function of time t; the beam deflection describes a sawtooth-like function over time. The transient oscillations until the deflection is stable after each repositioning are also shown. Below the beam deflection x the exemplary states pA, . . . , pE of five apertures of the APS (as examples out of a large number of apertures, typically many thousands) are shown. Each beamlet A to E will be projected onto the target only when the respective state pA to pE is low (non-energized, beamlet switched on) as further explained below with reference to FIG. 1. It can be seen that during each pixel exposure proper the states of the apertures are unchanged. It follows that during one period of the sawtooth function only a one-bit information per aperture can be processed. The prior-art approach makes it difficult to find a suitable on-chip data storage layout that is efficient and can handle the required data rates. Currently the use of shift registers is the most promising solution for this problem (see US 2005/0242303 A1). The present invention provides a method which offers high throughput rates in combination with a PD design that is easy to implement to overcome the deficiencies of the prior art. The invention is directed to a target irradiating method as set out at the beginning, wherein during pixel exposure periods the location of the beam image is moved along with the target, with the pixel exposure periods having a duration within which the relative movement of the target covers a distance, the distance of advance, which is greater than said first width as measured on the target, preferably at least a multiple of the first width (i.e., at least 2.0 times and more preferably 3 times the first width or more); and after the pixel exposure periods the location of said beam image is changed (repositioned on the target), so as to generally compensate the movement of the location of the beam image during the pixel exposure periods with regard to the location of the PD means. The invention offers a writing strategy that offers an optimal writing speed over the target and lends itself to an electronic layout which is realizable using standard CMOS technologies and allows a simplified realization of the blanking circuitry in the PD means. The invention provides a PD device for electron beam writing which makes it possible to write with the speed only limited by the e-beam density and the APS transitivity. Power consumption and data rate to the APS can be handled within the UMC 0.25 μm process. Currently used architectures would fail in these two aspects by many orders of magnitude. In order to ensure that the beamlets expose the complete surface of the target, the number of beamlets that expose different pixels will be chosen such that the first width multiplied with the number of said beamlets is equal to (or even greater than) said distance of advance. In the usual case that there are groups of redundant beamlets (viz., the beamlets which are in a same line as measured along the direction of the relative movement of the target form groups, ‘redundancy groups’, within which the beamlets have a distance from each other which, when measured on the target, is an integer multiple of said distance of advance) the number of these redundancy groups is chosen such that the first width multiplied with the number of said groups is equal to or greater than said distance of advance. In a preferred embodiment of the invention, the relative movement of the target is a continuous linear movement for durations comprising multiple pixel exposure periods. Usually, it is the target which is moved while the PD device and the optical system are kept fixed. The target movement may preferably be a scanning motion along scanning sweeps in a primary direction. Furthermore it is advantageous if outside of said pixel exposure periods the patterned beam is blanked out in a manner that it does not reach the target, but is absorbed at an absorbing means before it can reach the target. In a suitable choice of the relative movement of the target, the distance made during periods of time between the starts of consecutive first intervals amounts to a second width being an integer multiple of said first width. In this case, the separation distances of two beamlets which are in a same line as measured along the direction of the relative movement of the target may be integer multiples of said first width, and at least some of said integer multiples may be chosen so as to be relatively prime to the integer multiple of said second width. One simple way for this is that said integer multiples differ by one unit. In a preferred embodiment the pattern definition means comprises a plurality of electrostatic beamlet deflectors, each of which is associated with a respective aperture of the pattern definition means and is adapted to deflect the beamlet traversing the respective aperture by an amount dependent on an individual control signal, including an amount sufficient to deflect the beamlet off its nominal path (thus effectively switching off that beamlet so it does not reach the target). In this case, the apertures of the pattern definition means are controlled in a time-dependent manner according to a pattern to be formed on the target by means of a plurality of said control signals. In this preferred embodiment, preferably, the apertures are logically grouped into disjoint groups, and the control signals of the apertures of each group are applied through a common control line in a time-staggered manner. Suitably, the apertures of each group may be arranged in the pattern definition means along a direction of orientation generally perpendicular to the direction of relative movement of the target. The apertures of each group may then extend over one half of a breadth of an area covered by the apertures of the pattern definition means, as measured along the direction of orientation of generally (but not necessarily) perpendicular to the direction of relative movement of the target, with the control signals being fed to said area of apertures from two opposite sides. Also in this preferred embodiment, the pattern definition means may comprise data storage means, where data for activating the control signals are fed to and stored in; allowing that said data are individually read out to accordingly activate the control signals. The data storage means may be double-ported, wherein during each pixel exposure period data for controlling the apertures are read out to the control signals through a first port of said storage means, and data concerning the consecutive pixel exposure period is stored through a second port of said storage means. Another profitable development of this preferred embodiment is to group the apertures of the pattern definition means into groups of mutually redundant apertures, wherein for each group the control signals associated with the apertures of the group are generated from a common signal which is copied to the control signals in turn. In this case, the common signal may be copied to the control signals with respective time delays in correspondence with the spatial offset of the respective apertures along the direction of relative movement of the target. In order to realize gray levels (see US 2005/0242303 A1) the common signal of a group may include information of a gray value to be realized through the apertures belonging to the group, and then the gray value is expanded into a sequence of non-active and active control signals in accordance with the gray value which defines the relative number of non-active control signals over the total number of control signals in the group. Expanding of the gray value may be performed before or after reading out the data from the storing means, and the expanding method may advantageously comprise at least one of the following methods: expansion into an evenly-spaced sequence, expansion into a pseudo-random sequence, expansion using linear feedback shift registers, expansion using different offsets for the individual control signal bits; or a combination thereof. Of course, the invention can also be combined with other known writing strategies, such as an additional global blanking signal which asynchronously masks the blanker flip flop outputs (see WO 2007/112465). Furthermore, the plate realizing the array of apertures may be doubled and operated in parallel so that the apertures are arranged in (at least) two parallel arrays, displaced to each other by a distance along the beam. Another aspect of the invention relates to positioning a target using a beam of energetic electrically charged particles, in particular in conjunction with a target irradiating method as discussed here, wherein during the positioning method the target is moved relative to the pattern definition means along a path within an image plane in which the images of apertures are formed, wherein the relative movement of the target covers a distance of advance greater than the first width as measured on the target, while at least part of the beam impinges on markers provided on the target which then produce secondary radiation, and the secondary radiation is detected by a detection system with a suitably high sampling rate, so as to adjust the position of the target in a manner that the secondary radiation as measured obtains an extremal (maximum or minimum) value. With this method, preferably, the substrate stage moves the target at a continuous speed equaling a speed at which the target is moving during a subsequent exposure process according to the irradiation method described above. One suitable realization of the markers comprises structures which are part of a multi-layer structure, and these structures have been formed on the target in at least one foregoing process step. In particular, the markers may comprise components adapted to produce secondary radiation comprising characteristic Auger electron radiation, and the detection system employ an energy selective detector adapted to detect said Auger radiation. In another realization the markers may comprise components having a characteristic backscattering yield. A particular advantage is that the target positioning method and a subsequent particle-beam exposure exposure/irradiation process performed on the target can employ the same pattern definition means with the target moving relatively to the pattern definition means in a corresponding manner. However, the distance of advance of the target positioning method may be larger than that during the particle-beam exposure/irradiation process. The preferred embodiment of the invention discussed in the following is a development from the PML2-type particle-beam exposure apparatus with a pattern definition (PD) system as disclosed in the U.S. Pat. No. 6,768,125 (=GB 2 389 454 A) of the assignee/applicant, and with a large-reduction projecting system. In the following, first the technical background of the apparatus is rendered as far as relevant to the invention, then embodiments of the invention are discussed in detail. It should be appreciated that the invention is not restricted to the following embodiments or the particular layout of PD system, which merely represent one of the possible implementations of the invention; rather, the invention is suitable for other types of processing systems that employ a particle-beam with projection stages as well. PML2 System A schematic overview of a maskless particle-beam processing apparatus PML2 employing the invention is shown in FIG. 1. In the following, only those details are given as needed to disclose the invention; for the sake of clarity, the components are not shown to size in FIG. 1, particularly the lateral width of the particle beam is exaggerated. For more details, the reader is referred to the U.S. Pat. No. 6,768,125. As already mentioned, a particle beam generated by a particle source is used in the PML2 system. An illumination optical system forms the beam into a wide beam which illuminates a PD means having a regular array of apertures in order to define a beam pattern to be projected on a target surface. With each aperture, a small beam is defined, and the passage of each beam through an aperture can be controlled so as to allow (‘switch on’) or effectively deactivate (‘switch off’) the passage of particles of the beam through the respective apertures towards the target. The beam permeating the aperture array forms a patterned particle beam bearing pattern information as represented by the spatial arrangement of the apertures. The patterned beam is then projected by means of a particle-optical projection system onto the target (for instance, a semiconductor substrate) where an image of the apertures is thus formed to modify the target at the irradiated portions. The image formed by the beam is moved along a straight path over each die field; additional scanning of the beam in a direction perpendicular to the scanning direction is not necessary (except, where needed, to compensate for lateral travel motion errors of the target stage). The main components of the apparatus 100 are—in the order of the direction of the lithography beam lb, pb which in this example runs vertically downward in FIG. 1—an illumination system 101, a PD system 102, a projecting system 103, and a target station 104 with the target or substrate 14. The particle-optical systems 101, 103 are realized using electrostatic or electromagnetic lenses. The electro-optical parts 101,102,103 of the apparatus 100 are contained in a vacuum housing (not shown) held at high vacuum to ensure an unimpeded propagation of the beam lb, pb along the optical axis of the apparatus. The illumination system 101 comprises, for instance, an ion source 11, an extractor arrangement 11a defining the location of the virtual source, a particle filter/general blanker 12 and an illumination optics realized by a condenser lens system 13. The ions used can be, for instance, hydrogen ions or heavy ions; in the context of this disclosure heavy ions refer to ions of elements heavier than C, such as O, N, or the noble gases Ne, Ar, Kr, Xe. Apart from ions, the particles can be electrons (emitted from an electron gun) or, in general, other electrically charged particles can be used as well. The ion source 11 emits energetic ions of primarily a certain species, such as Ar+ ions having a defined (kinetic) energy of typically several keV (e.g. 5 keV at the PD system 102) with a comparatively small energy spread of, e.g., ΔE=1 eV; however, it is generally unavoidable that the source emits charged particles of other species as well. A velocity/energy dependent filter 12 serves to filter out such unwanted particle species; this device may also be used to blank out the beam as a whole during repositioning of the beamlets. By means of an electrooptical condenser lens system 13, the ions emitted from the source 11 are formed into a wide-area, substantially telecentric ion beam serving as lithography beam lb. The telecentricity of the beam is within a range of ±25 μrad deviation from the optical axis at the position of the PD device, resulting in a telecentricity range of ±5 mrad deviation from the optical axis at the position of the substrate, assuming here a 200× reduction system and equal particle energies at PD device and substrate. The lithography beam lb then irradiates a PD device which, together with the devices needed to keep its position, forms the PD system 102. The PD device is held at a specific position in the path of the lithography beam lb, which thus irradiates an aperture pattern formed by a plurality of apertures 21. As already mentioned, each of the apertures can be “switched on” or “open” so as to allow the beamlet passing through the respective aperture to reach the target; it is then said, the aperture is transparent to the incident beam. Otherwise, the aperture is “switched off” or “closed”, in which case the beam path of the respective beamlet is affected in a way that it will be absorbed or otherwise removed out of the beam path before it can reach the target; thus, the aperture is effectively non-transparent or opaque to the beam. The pattern of switched-on apertures is chosen according to the pattern to be exposed on the substrate, as these apertures are the only portions of the PD device transparent to the beam lb, which is thus formed into a patterned beam pb emerging from the apertures (i.e., in FIG. 1, below the PD system 102). The architecture and operation of the PD device, in particular with regard to its blanking plate, is discussed in detail below. In FIG. 1 only five beamlets (out of a large number) are shown in the patterned beam pb, of which the second beamlet from the left is switched off as it is absorbed on an absorbing plate 17; the other, switched-on beamlets pass through a central opening of the plate 17 and thus are projected onto the target. The pattern as represented by the patterned beam pb is then projected by means of an electro-optical projection system 103 onto the substrate 14 where it forms an image of the switched-on mask apertures. The projection system 103 implements a demagnification of, for instance, 200×. The substrate 14 is, for instance, a silicon wafer covered with a photo-resist layer. The wafer 14 is held and positioned by a wafer stage (not shown) of the target station 104. A detector 15 for secondary radiation can be used to detect the proper positioning of the substrate with respect to the beam; this is further discussed below (‘Marker Reading’). The projection system 103 is preferably composed of two consecutive electro-optical projector stages with a crossover c1, c2, respectively. The electrostatic lenses 30 used to realize the projectors are shown in FIG. 1 in symbolic form only as technical realizations of electrostatic imaging systems are well known in the prior art. The first projector stage images the plane of the apertures of the PD device to an intermediate image which in turn is imaged onto the substrate surface by means of the second projector stage. Both stages employ a demagnifying imaging through crossovers c1,c2; thus, while the intermediate image is inverted, the final image produced on the substrate is upright (non-inverted). The demagnification factor is about 14× for both stages, resulting in an overall demagnification of 200×. A demagnification of this order is in particular suitable with a lithography setup, in order to alleviate problems of miniaturization in the PD device. The electrooptical lenses are mainly composed of electrostatic electrodes, but magnetic lenses may also be used. As a means to introduce a small lateral shift to the image, i.e. along a direction perpendicular to the optical axis cx, deflection means 16 are provided in one or both of the projector stages. Such deflection means can be realized as, for instance, a multipole electrode system, as discussed in the U.S. Pat. No. 6,768,125. Additionally, a magnetic coil may be used to generate a rotation of the pattern in the substrate plane where needed. The lateral deflections are quite small in comparison to the lateral width of the patterned beam by itself. Even though the lateral deflections that are required with the invention are a multiple of the width of a single beamlet, that distance will still be at least one order of magnitudes below the beam width since the lateral dimension of one beamlet is considerably smaller than that of the beam by seen as a whole. By controlling the pattern formed in the PD system 102, an arbitrary beam pattern can be generated and transferred to a substrate. Suitably, a scanning stripe exposure strategy, where the substrate is moved under the incident beam, is utilized so a beam-scanning strategy is not required, where the position of the beam is perpetually changed and thus the beam is effectively scanned over the (more or less resting) target surface like in case of a single focused beam system. FIG. 17 shows a longitudinal section detail of one possible realization of the PD system 102 in a two-plate arrangement. The layout shown is known from prior art but is suitable to use with the invention nonetheless. A first plate 201 is an aperture plate having a set of apertures 21; only three out of a large number of apertures are shown. The plate 201, through its apertures 21, defines a corresponding number of beamlets b1, b2, b3. The apertures 21 are imaged to the target as mentioned earlier. The apertures 21 are arranged in a systematic manner forming staggered lines running parallel to a direction which corresponds to the relative movement of the images of the apertures over the target as described in the U.S. Pat. No. 6,768,125. In each line the offset between consecutive apertures is preferably a multiple of the width of the aperture, while the lines run immediately side by side such that the aperture images completely cover the target in the course of the scanning movement over the target. The second plate 202 of the PD system 102 is called blanking plate. It has a set of openings whose positions correspond to those of the apertures 21 in the aperture plate 201, but whose widths are greater so the beamlets pass through them without affecting the blanking plate material. The openings in the blanking plate 202 are provided with electrodes 221 so as to impart a small but sufficient deflection to the corresponding beamlet. Each beamlet can be deflected individually as explained in detail below. The blanking plate also comprises the circuitry for electronic controlling and electric supply of the electrodes. Further basic details of a PD device are discussed in the U.S. Pat. No. 6,768,125 and related prior art, whereas a preferred layout of the blanking plate circuitry according to the invention is discussed below. FIG. 18 illustrates the plan view arrangement of apertures on the PD device typical in prior art. The apertures are represented as cross-hatched squares, the orientation is chosen such that the horizontal direction coincides with the direction of (relative) substrate movement during exposure (cf. FIG. 19). Along this direction, the apertures are evenly spaced in rows, and between adjacent rows the apertures are offset, so as to realize a staggered arrangement, with the offset being an integer multiple of the aperture width w. In the example shown the offset is 3w, and the pattern of apertures repeats itself every three rows. Thus, the pattern repeats itself after 3×3w in the horizontal direction. As a consequence each aperture has an allotted cell space of 9w2, or 3w×3w visualized as squares surrounding the apertures. The writing strategy according to prior art used with this PD aperture arrangement was discussed above with reference to FIGS. 19 and 20. The regular arrangement shown in FIG. 18 may be interrupted by spaces with no apertures as discussed in U.S. Pat. No. 6,768,125, e.g. to offer space for the circuitry for the internal processing of the pattern data. Writing Strategy In contrast to known writing strategies (cf. FIG. 19) we propose a writing strategy that minimizes data transfers on and into the APS. We refer to this strategy as “trotting” strategy, because the movement of the individual beamlets was found to be reminiscent of the movement of the diagonal legs of a horse when it is trotting. FIG. 2 illustrates the principle concept of the “trotting” strategy according to the invention with a first, simplified example. A target 40, for instance a silicon wafer covered with a resist to be exposed, moves with a velocity v, while a number of beamlets (only five beamlets denoted A through E are shown for the sake of clarity) impinges on the target. Each beamlet passes the complete dose for one pixel to the resist during one pixel exposure cycle T1 lapsing between the uppermost and the lowermost frame in FIG. 2. Within the figure the cross-hatched fields denote pixels that are completely illuminated (according to the programmed pattern), while simply hatched fields indicate pixels which are currently exposed and thus are summing up the necessary dose. The target is moved with respect to the PD device (which is assumed to remain at a fixed position beyond the top of FIG. 2), but the writing method requires that the location where each beamlet impinges on the target stays locked on the position of a pixel during a respective pixel exposure cycle in spite of that relative movement, even for a movement of the target by a distance covering multiple pixels. As a consequence, the beamlets are deflected accordingly by means of the deflection system of the imaging optics (deflectors 16 in FIG. 1). The width of a pixel as exposed on the target is denoted w, the distance of two neighboring beamlets is N′w, wherein N′ is one more than the number N of beamlets writing in parallel on one line. The factor N+1 is one simple choice to ensure that N and N′ are coprime (relatively prime); any other number N′ which is coprime to N could be used instead. Such a choice of N′ and N will ensure the exposure of adjacent pixels through a run of N′ exposure cycles. In the case of FIG. 2, N=5 and N′=N+1=6. Starting from an initial position at the beginning of an exposure cycle, shown in the initial (uppermost) frame at t=0:0, the times are consecutively denoted as t=0:1 where the continuous motion of the target has progressed by one pixel width w, t=0:2 at two pixel widths 2w (not shown), and so on. At t=0:5 (t=0:N), the movement will have proceeded to a distance of five (N) pixel widths 5w, which marks the end of the actual exposure. After that, the beamlets are directed to another set of pixels as shown in the lowermost frame of FIG. 2 to start a next pixel exposure cycle at t=1:0 (which is a time point shortly after t=0:N; in other words t=0:N+δ=1:0, where δ denotes a small time interval used to switch the beamlets). The on/off-switching of every blanker in the PD device is controlled according to the programmed pattern, during each pixel exposure cycle. For example, if one pixel which is illuminated by its corresponding beamlet is to receive a dose of 25% of the maximum value, the blanker for that beamlet has to be opened for 25% of the pixel exposure cycle. It should be noted that the inclination of the beamlets in FIG. 2 (and, likewise, FIG. 19) is largely exaggerated; in a true depiction of a realistic geometry of a wafer process, the angles of inclination would be quite small and seemingly inconspicuous to the human eye. Moreover, it is also possible (by using, for instance a consecutive set of electrostatic multipoles) to shift the set of beamlets laterally without affecting their orientation; in that case, the angle of inclination would not change at all. Also, the range of deflection may suitably be chosen to be symmetric around the z axis. One typical value of the distance N′w between consecutive beamlets at the wafer is 180 μm. The positioning of the beamlets on the respective pixels is controlled via suitable ion-optical devices within the imaging system, for instance an electrostatic multipole. FIG. 3 (top frame) shows the functional relation of the spatial deflection x of the beamlets as a function of the time. Not all of the time can be used to open the blanker apertures since the beamlets and multipoles need a certain settling time, denoted Ts in FIG. 3, to settle after repositioning and transient oscillations. The settling duration Ts is a small fraction of the pixel exposure cycle T1. The remaining part of the pixel exposure cycle T1, the usable time Tu=T1−Ts is used for the exposure of pixels. The time interval Tu is the pixel exposure period within which to ensure that the appropriate dose is passed to the respective pixels. In FIG. 3 also the time points of the frames of FIG. 2 are denoted. Accordingly, the time interval between consecutive frames (like 0:0 and 0:1) is dt=Tu/N (here, =Tu/5), but the last two frames of FIG. 2 are separated by the duration Ts. The writing process as described in FIG. 2 implies no redundancy for the blanker layout. This implies the potential danger that if one blanker cell fails to work the corresponding pixel is always illuminated or not illuminated at all (depending on which way the blanker cell is defect). Redundancy can be added simply by doubling, tripling etc. the number of blanker cells that illuminate one pixel; in general, a K-multiplication. A redundant tripling of the aperture (i.e., K=3) is exemplarily illustrated in FIG. 4. Note that the distance of those redundant apertures is an (integer) multiple of Nw. For instance, it is possible to realize K sets of (non-redundant) N′ consecutive apertures, so K blocks of redundant apertures are writing within one line, and the offset between each of these K blocks is N'Nw. An alternative arrangement is to have K consecutive redundant apertures in each block, and arrange N′ blocks suitably (for instance with an offset of N'w between the last aperture of one block and the first aperture of the next block); the inter-block offset is then KNw. In a realistic implementation of the invention, the non-redundant and redundant writing as explained above with reference to FIGS. 2 and 4 will be combined. A simple example is illustrated in FIG. 5, with N=3, N′=4 and K=2. The beamlets are denoted A1, A2, B1, B2, C1, C2, . . . ; which is a combination of a capital letter A, B, . . . (denoting beamlets of different redundancy groups) with a digit 1 or 2 which denotes the individual redundant beamlets. Thus beamlets A1 and A2 are redundant and spaced apart by Nw=3w, while the spacing between the beamlets of subsequent redundancy groups, for instance between one A beamlet and one B beamlet, is a value (nN+N′)w, which is coprime with Nw. Returning to FIG. 3, the time Tu and the particle flow in the beam is chosen such that the maximum dose for one pixel, divided by a redundancy factor K′, can pass through one aperture within the duration Tu. The redundancy factor used here, K′, usually equals K, but may be chosen smaller if not all redundancy columns are activated. In contrast to the prior art method as shown in FIG. 20 where loading of information for each pixel is done individually and the deflection x is reset for each pixel, according to the invention the usable time Tu within each sawtooth cycle of deflection x covers multiple pixels. Furthermore a reloading of pixel information may occur multiple times within an cycle Tu according to the respective gray level of the pixel (see below ‘Data Encoding’), as shown for the pixel deflection signals pA to pE in FIG. 3. Architecture of the Blanker Plate In the following an architecture of the electronics for a blanking plate 202 is disclosed that allows to establish a writing process according to the above writing method principle. An overall block diagram of such a blanking plate 60 is depicted in FIG. 6. The blanking plate may physically be built from one single wafer die which includes logic circuits and blanking apertures, or it may be a stack or composite architecture of two or more parts. For example, a die that incorporates the logic circuitry may be contacted to a die incorporating the blanking electrodes. The blankable openings are arranged within an aperture field 61, according to 2R rows, each row having C apertures. Thus, the blanker cells can be viewed as C cell columns running perpendicular to the rows (vertical in FIG. 6; the direction of the rows is horizontal). The number C of blanker elements (apertures) in each row is preferably equal to N·N′ or an integer multiple of that number. The blanker cells are addressed in a column-wise fashion using two APS memory arrays 62N, 62S (N and S standing here for ‘north’ and ‘south’, respectively), which for instance are dual ported static RAM blocks allowing a high bit density. The memories are accessible from the external interface 63 of the blanking plate 60 in read and write mode. A control logic 64 interprets commands and data passed to it via the external interface 63 and accordingly performs addressing 65 of the memory arrays 62N,62S and provides the control signals 66 (clock, enable etc.) for the blanker cells within the aperture field 61. The north RAM data supplies the data and control signals to the upper R blanker rows and the south RAM is connected to the lower R blanker rows. In the embodiment shown here, the memory arrays 62N,62S are composed of a number of memory blocks 67 realized by individual memory units. Each of these blocks 67 is realized as dual ported RAM (DPRAM). Such DPRAMs are standard library elements of ASIC vendor libraries. It will be obvious that in other embodiments, the memory of the blanking plate 60 may be organized otherwise, for example in two memory banks (corresponding to arrays 62N,62S) or only one memory, or a number of memory blocks 67 may be realized by respective memory units, depending on the memory size of the memory units. Referring to FIG. 7, the RAM of each memory block 67 is built of static storage elements, which allow a very high storage density within the ASIC. The entire contents of each RAM block is addressable through two ports, labeled Port A and Port B, each of which is accessible through respective access means comprising an address bus, a data bus and additional control signals, as further explained below. The two ports allow simultaneous access of the information from two different locations. Note that the data width and address width does not necessarily need to be the same on both sides, the memory can—for example—be organized in 8 bits on Port A and as 32 bits on the Port B. FIG. 8 shows the design of a single half-column HC1, corresponding to the upper (north) half of one blanker column, according to the preferred embodiment of the invention. The blanker cells are denoted Blanker0, . . . , Blanker(R−1) and symbolically depicted as square areas with a small square (the aperture). The lower half of the column is realized symmetrically (mirrored at the horizontal middle axis of the aperture field 61), and further columns simply add to the left and right of the column shown. The blanker information is stored into the DPRAM 82 through Port A, and afterwards is transferred to the individual blanker cells through Port B as control signal data through the control lines 84. Every blanker cell is provided with a 1-bit memory (e.g., a flip flop) to store the current data provided from the DPRAM. In the layout of FIG. 8 the width bw of the control line data is 2 bits (bw=2), but it can have any other number as well. It is useful to choose the number R in a way that it is an integer multiple of bw. Higher values of bw allow a shorter reloading time of a complete column. The individual blanker cells are enabled to store the information provided from the DPRAM by means of enable-row signals er0, er2, . . . er(R−2) at the correct point of time, each enable-row signal activating a group of bw cells. Only one enable-row signal is active at a time, so the data on the control line 84 is loaded into the one correct blanker cell group; after that, the next blanker cell group is loaded, and so on. For example, if R=128 and bw=8, a column can be reloaded within 128/8=16 clock cycles. Within a 70 MHz design this relates to a time of 230 ns. Preferably, sufficient memory is provided to allow an “interleaved” reloading of the blanker cells: The memory of the DPRAM is divided into two address areas A0 and A1 so it is possible to reload one address area A0 via Port A and at the same time load the data stored in the other address area A1 into the blanker cells. In the next pixel exposure cycle the banks are switched, A1 is reloaded via Port A while A0 is loaded into the blanker cells. For efficient writing algorithms as well as for testing purposes it is useful to provide a static data setting on the control line 84 that allows switching all blanker cells into the blanking mode (switched-off apertures) at the same time. This is done by means of a “mask” block 83, activated by a mask-out signal. If the mask-out signal is set, the data of all blanker cells in the half-column shown will be set to 1, whereas in the normal working mode the mask block 83 is transparent to the data coming from Port B. Furthermore, it is possible to optionally realize a global blanker-enable signal, which allows switching all apertures of the aperture field simultaneously, for instance as a signal activating all mask-out and enable-row signals of all (half-)columns. The present layout also allows implementation of “gray levels”, i.e., exposure doses between no exposure (minimum dose, 0%) and full exposure (maximum dose at a pixel, 100%; see also US 2005/0242303 A1). The gray levels are realized as discrete steps from 0 to g−1, with g being an integer representing the 100% level. The number g of gray levels that can be realized may be realized in the present layout as follows. FIG. 9 shows a timing diagram for the controller operation with 8 gray levels (g=8, three bits), time shown proceeding along the horizontal axis. Conceptually, the time Tu is divided into g time slots, during some of which the pixel is activated according to the desired gray level. For example, with a gray level 5 out of 8, there are eight time slots in each Tu and five of them are set to switched-on. If the dose applied to one pixel within the time Tu is digitized into g gray levels, the blanking cells are reloaded g times within Tu. The blanker cell groups are labeled consecutively by enable-row designators, the gray level components are labeled g0 through g7. The complete loading cycle of the blanker cells is done within one “usable” cycle time Tu. The data is loaded into the blanker cells at g consecutive times (corresponding to time slots) and activated at the start of the respective time slots as shown in FIG. 9; simultaneously, the next set of data for the following exposure cycle is loaded. Note that the loading times are staggered between different blanker cell groups as denoted in FIG. 9. In the final part of Tu, each blanker cell group is switched off after the last gray level slot g7 expires; this is conveniently achieved by means of the mask block 83 controlled through the mask-out signal as shown in FIG. 8. The switched-off state continues until the first gray level slot g0 of the next exposure cycle is loaded and activated. The timing diagram shown provides an optimal dose preparation by the APS. In FIG. 3, below the depiction of the time-dependent deflection x, examples for data streams of five different apertures during the time Tu are illustrated. These data streams can be generated directly on the APS from the gray level data. This method is explained in more detail within the following section. Data Encoding In actual applications, the patterns to be generated comprise gray levels, realizing g gray levels ranging from 0 to 100% of a full exposure of a pixel. Since for g gray levels g bits will have to be stored within the SRAM, this incurs an overhead within the data. To lower the data rate via the external interface it is possible to send only the gray level G (G=ld(g) bits) to the APS and to provide a data expansion from a gray level (a G-bit number) to the g aperture bits inside the APS. This may be done before the data is stored within the DPRAM or when reading the data from the DPRAM and passing it to the blanker cells. In the case that the data are first expanded and then stored in the DPRAM, each gray level has to be converted into a vector of bits. Successive vectors (i.e. vectors of neighboring apertures) are arranged into a matrix. After collecting a certain number, for convenience the same number as gray levels are used to obtain a g×g matrix, the matrix has to be transposed (changing rows and columns) and written to the DPRAM. Expansion after storing in the DPRAMs (“late extraction”), on the other hand, saves memory area, and the controller does not need to change the data format from the data provided via the external interface. FIG. 10 depicts one basic logic layout realizing late extraction with redundant columns. Two corresponding half-columns are shown in FIG. 10 and denoted HC1.0 and HC2.0, respectively. (For the sake of clarity, only two columns which are redundant as explained with reference to FIG. 4 are shown in FIG. 10; the second column HC2.0 is representative for all further redundant columns. The remaining columns, in particular columns providing writing of further non-redundant pixels, are suppressed.) An additional time signal 840 is used, which is common to all columns and is updated after each gray-level slot. The DPRAM logic 871 for the first column HC1.0 derives, from the data received, the full data of width G·bw for the set of redundant half-columns. An extractor 841 expands the gray level information and extracts the single bit data (bw bits) for the first half-column 1.0 based on the time signal as further explained below. The full data of width G·bw are passed to the DPRAM of the next redundant column HC2.0. The DPRAM logic 872 of the half-column HC2.0 can therefore use the copied data in the next illumination cycle; only the address data are supplied from the control logic. The procedure is repeated in an analogous manner for further redundant half-columns (not shown) as with column HC2.0. Note that since the columns HC1.0, HC2.0 are designed for redundant writing, they will necessarily operate with identical gray-level data; the expansion of gray level information into individual on/off information is done in the respective extractors 841, 842. In the layout shown in FIG. 10, the DPRAM of column 1.0 also provides a buffering functions for the data input from the control logic. For the data expansion different algorithms can be used, as for example algorithms to generate pseudorandom sequences, distributing the “on” time of the redundant apertures equally over Tu in a quasi-stochastic manner. The time information 840 is used by the extractors 841, 842 for generating an exposure bit in accordance with the desired gray level and the actual time to obtain the bit sequence appropriate for the gray level. At every time step corresponding to a gray-level slot the extractor generates bw bits employing or emulating, for each of the bw rows, an extraction function f which takes the time information 840 (G bits) and the gray level (G bits) associated with the respective row and returns a bit signal (boolean function); in symbolical form: f=f(Time, Graylevel). The result of this function is then passed as a control signal to the corresponding aperture. In a simple embodiment the time information 840 is generated by a linear counter (not shown) which is incremented at each time step, and the extraction function f is a comparison f=compare (Time<Graylevel) In a preferred embodiment the time signal 840 is the output of a G-bit random number generator (e.g. linear feedback shift register LFSR) that generates the numbers 0 to g for the calculation of every bit g of the exposure control signal stream in a random order, but each number once for a set of g gray-level slots. Then, the function f can again be a comparison f=compare(Time<Graylevel), which directly relates to the control signal passed to the apertures. The use of such a random generator decorrelates the bit stream for an aperture in time. A further parameter can be incorporated in the comparison to decorrelate the data streams between individual pixels. For instance, every pixel obtains beside its gray level information also an individual number acting as a “sequence-key” that determines the actual bit sequence generated for the specified gray level. The sequence-key is kept unchanged during the whole exposure cycle. The sequence-key is used as an offset value added or subtracted (modulo g) to the value provided by random number generator before the comparison with the gray level is performed; i.e., symbolically, f=compare([(Time-Sequencekey) (mod g)]<Graylevel). Redundant Blanking Plate Configuration One possible (and presumably the most likely) type of failure in an APS system as shown here is a failure of a blanking cell due to mechanical or electrical damage, such that the corresponding beamlet cannot be blanked (always open error). By providing a stack of two blanking aperture plates positioned one on top of the other and electronically running in parallel, the probability of always open errors will be significantly reduced. This redundant configuration with two blanker plate devices 202a, 202b working in parallel is illustrated in FIG. 11. The two outer apertures shown in FIG. 11 are intact, whereas the two inner apertures have defects in the first or second blanker plate, respectively. Due to the redundant stacking of blanker plates, the proper operation of the two apertures is ensured. Aperture Arrangement FIG. 12 shows an example of the arrangement of apertures according to the invention. This drawing is to be seen in comparison to the prior-art arrangement of FIG. 18 discussed above. In the plan view of FIG. 12 the orientation is the same as in FIG. 18, and the directions of orientation coincide with those of FIGS. 6 and 8; the apertures are represented as cross-hatched squares. Like in FIG. 18, each apertures has an allotted cell space of 3w×3w. In contrast to prior art, an additional space is inserted between each block of three apertures in order to realize a spacing of N'w (here: 10w) which is coprime with the basic spacing within a row (here: 3×3w). FIG. 13 shows a variant of the arrangement including a redundancy of two (K=2). Each block of apertures is now doubled. The enlarged detail of FIG. 14 shows the elementary cell of this arrangement. As can be seen, the spacing of apertures within the redundancy group of apertures is an integer multiple of the basic spacing (namely, in this case 9w), whereas the spacing to the first aperture of the next block comprises an additional offset; here, the offset is 1w, so N'w=(1+9)w. Thus, within this cell the apertures have a spacing to provide the redundancy and the APS can be constructed by a concatenation of multiple instances of this elementary cell. Marker Reading The proposed architecture can also be used for reading of markers 41, 42 on the target 40. Such markers consist of a material that reflects a high percentage of the beam current or produces secondary radiation upon irradiation with the beam so these reflections can be measured by means of a detector 44 located laterally to the beam. Preferable types of secondary radiation may be, for instance, secondary electrons or Auger electron radiation. In particular, Auger radiation has the advantage that it provides a characteristic electron radiation at a specific energy with very low line width, so it is possible to tune the detector 44 accordingly. For instance referring to FIGS. 15 and 16, a two-stage marker reading may be realized in order to allow a coarse and a fine synchronization of the positioning between table and beam position. In the first stage (FIG. 15), for the purpose of finding the coarse position on the resist, a marker coding with a high autocorrelation peak may be used, such as lines arranged as bar codes or with a prime-number marker spacing. The PD device is then configured to illuminate the marker pattern 41 on the resist, and in the case the pattern thus programmed matches the pattern on the target, a correlation peak is observable in the secondary current. Optimum performance will be obtained when all markers have the width w, viz. equal to the beamlet width. The correlation peak of the coarse marker pattern 41 is used to determine the position of a fine marker 42, which is located with a distance to the coarse markers 41 to allow real-time processing of the detector data, with an accuracy±w. In the second stage (FIG. 16), the fine marker 42 may be used to increase the resolution of the marker finding procedure. Since the position of the fine marker is known to be in a field that is 2w wide, the PD device and the deflectors are configured to step through this field with a step width d according to the needed resolution while the relative movement of the target is preferably maintained to proceed at the same continuous motion and speed as during the pixel exposure process discussed above. A first beamlet 1 is positioned at the beginning of the 2w interval (x=0), a second beamlet 2 at x=d, and so on until the end of the interval is reached. Thus, the beamlets overlap and cause an intensity distribution that increases in steps to a maximum and then decreases again as shown in FIG. 16. Whenever the highest reflection current is measured, the maximum congruence between beamlets and marker has been found. This process can take place independently from any movement of the resist. The detector 44 used in the position synchronization should have a sufficiently high time resolution which corresponds to the spatial resolution (i.e., w or d) required divided by the velocity v of the relative movement of the target. Therefore, the sampling rate of the detector 44 is suitably at least smaller than 1/Tu. In order to increase the signal integration time, the time during which the beam is locked on the substrate may be chosen larger than the time of exposing one pixel (e.g. 10 Tu) when it is passing the region where the markers are located, without changing or reducing the scanning speed of the target. In other words, the distance of advance used in the position synchronization may be a multiple of the distance which is covered by the relative movement of the target during a typical pixel exposure period Tu during a subsequent (or foregoing) exposure process. This approach may appear to cause that some pixels in the area immediately following the marker region along the scanning direction cannot be completely exposed, but in fact this will be avoided since the markers will be sufficiently separated spatially from the surrounding lithography pattern (or microstructure pattern) to be formed. FIG. 15a illustrates the use of a variant of the markers, namely, buried markers 411. The markers 411 may have been produced in the course of preceding processing steps and covered by a coating material generated in an additional processing step. The structuring of the markers 411 may also have been achieved in situ by appropriate methods such as deep ion-irradiation where the ions pass a coating layer and substantially only affect a buried layer (which would be the markers). One suitable realization of the markers is metallic layers, preferably in a multi-layer structure, which are located beneath but close to the target surface. FIG. 15a shows a first stage as explained above referring to FIG. 15, but it will be clear that the buried markers 411 may be used for fine resolution positioning as shown in FIG. 16 as well. Summarizing, the invention enables to realize designs using fully synchronous design methods with only one clock. Although the apertures will not switch simultaneously (which is an advantage for the power consumption characteristics within the chip), the time an individual aperture is opened or closed is extremely accurate as it only depends on the accuracy of the clock (which is normally within the range of some ppm). Further advantages of the present blanking plate layout are: Reduced number of flip flops that is strongly reduced over the earlier PML2 concepts. Number of flip flop switching lower by a factor of in the order of several thousands. Power consumption is reduced drastically (in the same mode of operation). Switching of the flip flops is distributed over a larger time. Switching of the apertures is distributed over a larger time. Design is fully synchronous and uses only one clock. Writing strategy allows corrections of the optical system within the data stream. While preferred embodiments of the invention have been shown and described herein, it will be understood that such embodiments are provided by way of example only. Numerous variations, changes and substitutions will occur to those skilled in the art without departing from the spirit of the invention. Accordingly, it is intended that the appended claims cover all such variations as fall within the spirit and scope of the invention.
	description	The present application is a continuation application of U.S. Pat. No. 8,942,346, issued on, Jan. 27, 2015, the entire disclosure of which is hereby incorporated by reference. This application relates generally to systems and methods for obtaining and displaying an X-ray image. In particular, this application relates to systems and methods for using an X-ray collimator to generate an X-ray image in which one or more corners of an X-ray detector that is used to capture the image are not displayed as part of the image. In this manner, a relatively large view of the image can be displayed and rotated on a square or rectangular display device without changing the image's shape or size as the image is rotated. A typical X-ray imaging system comprises an X-ray source and an X-ray detector. The X-rays that are emitted from the X-ray source can impinge on the X-ray detector and provide an X-ray image of the object (or objects) that are placed between the X-ray source and the X-ray detector. In one type of X-ray imaging system, a fluoroscopic imaging system, the X-ray detector is often an image intensifier or, more recently, a flat panel digital detector. In many medical imaging applications, a collimator is placed between the X-ray source and the X-ray detector to limit the size and shape of the field of the X-ray beam. The collimator can shape or limit the X-ray beam to an area of a patient's body that requires imaging, preventing unnecessary X-ray exposure to areas surrounding the body part that is being imaged and protecting the patient from needless X-ray exposure. And because the collimator can limit the X-rays impinging on the X-ray detector near the body part being imaged, the collimator helps improve image contrast and quality. For example, the collimator can reduce excess X-rays from impinging on a flat panel digital detector, reducing or preventing image blooming or bleeding (which tend to occur when the detector is overloaded with X-rays). Thus, some conventional collimators can minimize X-ray exposure and maximize the efficiency of the X-ray dosage to obtain an optimum amount of data for diagnosis. This application relates to systems and methods for obtaining and displaying a collimated X-ray image. The methods can include providing an X-ray device having an X-ray source, a square or rectangular X-ray detector, and a collimator. The collimator can be sized and shaped to collimate an X-ray beam from the X-ray source that exposes a receptor region on the detector. The collimator can allow the X-ray image received by the X-ray detector to have any suitable shape that allows a relatively large view of the image to be displayed and rotated on the display device without changing the shape or size of the image as it rotated. In some instances, the collimator provides the image with superellipse shapes or cornerless shapes having four substantially straight edges with a 90 degree corner missing between at least two edges that run substantially perpendicular to each other (e.g., a squircle, a rounded square, rounded rectangle, a chamfered square, chamfered rectangle, etc.). The Figures illustrate specific aspects of the systems and methods for displaying collimated X-ray images. Together with the following description, the Figures demonstrate and explain the principles of the structures, methods, and principles described herein. In the drawings, the thickness and size of components may be exaggerated or otherwise modified for clarity. The same reference numerals in different drawings represent the same element, and thus their descriptions will not be repeated. Furthermore, well-known structures, materials, or operations are not shown or described in detail to avoid obscuring aspects of the described devices. Moreover, for clarity, the Figures may show simplified or partial views, and the dimensions of elements in the Figures may be exaggerated or otherwise not in proportion. The following description supplies specific details in order to provide a thorough understanding. Nevertheless, the skilled artisan would understand that the described systems and methods for obtaining and displaying collimated X-ray images can be implemented and used without employing these specific details. Indeed, the described systems and methods can be placed into practice by modifying the illustrated devices and methods and can be used in conjunction with any other apparatus and techniques conventionally used in the industry. For example, while the description below focuses on systems and methods for displaying collimated X-ray images that were created using a fluoroscopic X-ray device that obtains X-ray images in near real time, the described systems and methods (or portions thereof) can be used with any other suitable device or technique. For instance, the described systems and methods (or portions thereof) may be used with X-ray devices that produce traditional, plain X-ray images; with X-ray treatment procedures used in radiation therapy; in procedures for collimating gamma radiation; in nuclear medicine; and/or for a combination of different imaging and/or treatment techniques. As the terms on, attached to, connected to, or coupled to are used herein, one object (e.g., a material, an element, a structure, etc.) can be on, attached to, connected to, or coupled to another object, regardless of whether the one object is directly on, attached, connected, or coupled to the other object or whether there are one or more intervening objects between the one object and the other object. Also, directions (e.g., on top of, below, above, top, bottom, side, up, down, under, over, upper, lower, horizontal, vertical, etc.), if provided, are relative and provided solely by way of example and for ease of illustration and discussion and not by way of limitation. Where reference is made to a list of elements (e.g., elements a, b, c), such reference is intended to include any one of the listed elements by itself, any combination of less than all of the listed elements, and/or a combination of all of the listed elements. Furthermore, as used herein, the terms a, an, and one may each be interchangeable with the terms at least one and one or more. Additionally, the terms X-ray image, image, collimated image, and collimated X-ray image may refer to an X-ray image that is produced from a portion of an X-ray detector that is exposed to an X-ray beam that has been collimated with a collimator. As used herein, in some embodiments the term square may refer to a shape with four sides of equal length that also has four 90 degree corners. The term circle, in some embodiments, may refer to a closed plane curve having all points at a given distance from a common center point. The term squircle, in some embodiments, may refer to a Boolean intersection of a concentric circle and square, where the final shape has an area less than either the circle or the square. The term squircle, in other embodiments, may refer to a Boolean intersection of a square and a concentric circle whose diameter is greater than the length of the side of the square, but less than the diagonal of the square. The term mathematical squircle, in some embodiments, may refer to a specific type of superellipse with a shape between those of a concentric square and circle and may be expressed as a quadric planar curve or as a quadric Cartesian equation. A mathematical squircle, as opposed to the squircle shapes immediately above, maintains the tangent continuity between the circular corners with the flatter edges of a superellipse. The terms rounded square and rounded rectangle, in some embodiments, may respectively refer to a square or a rectangle with fillets breaking the corners (e.g., circular corners that are tangent to the edges of the square or rectangle). Additionally, in some embodiments the terms chamfered square and chamfered rectangle may respectively refer to a square and rectangle having any number of chamfers breaking their corners. As mentioned above, this application describes systems and methods for displaying collimated X-ray images. In some embodiments, the described systems and methods use a collimator to prevent an X-ray beam from impinging on one or more corners of an X-ray detector. The collimator can provide the image with any suitable shape that allows one or more corners of the X-ray detector that is used to obtain the image not to be displayed in the image. In some instances, the X-ray image has a perimeter with a (i) a superellipse shape and (ii) a cornerless shape with at least two substantially straight edges that run substantially perpendicular to each other, wherein such edges do not physically intersect with each other at a 90-degree corner. Some embodiments of a collimated image 10 are shown in FIG. 1. By having any of the described shapes, the collimated image can be shown on a relatively large portion of a display area of a display device (e.g., a square or rectangular monitor, screen, projector, TV, etc.), and the entire image can be viewed as it is rotated about its center, without requiring the image to be reshaped or resized. Thus, the described collimated image can maintain its size and geometry during rotation on the display device, while maximizing its on-screen, image size and the amount of the receptor area of the X-ray detector that is used to take the image. FIG. 2 shows some embodiments of a method 150 for displaying the described collimated X-ray images. Although this method can be modified in any suitable manner (including by rearranging, adding to, removing, modifying, substituting, and otherwise modifying various portions of the method), FIG. 2 shows those embodiments in which the method begins at 155 by providing an X-ray system 15. The X-ray system 15 can comprise any suitable X-ray device that is capable of capturing the described X-ray images 10. For example, the X-ray system can comprise a mobile X-ray device (e.g., an X-ray device comprising a C-arm, a mini C-arm, an O-arm, a non-circular arm, etc.), and a fixed X-ray device. By way of illustration, FIG. 3 shows an X-ray imaging system 15 comprises that a C-arm X-ray device 18. The X-ray system 15 can also comprise any component that allows it to take the collimated X-ray images 10. In some embodiments, FIG. 3 shows the X-ray imaging system 15 comprises an X-ray source 20, an X-ray detector 25, and a collimator 30. Any suitable X-ray source can be used, including a standard X-ray source, a rotating anode X-ray source, a stationary or fixed anode X-ray source, a solid state X-ray emission source, or a fluoroscopic X-ray source 35 (as shown in FIG. 3). Any suitable X-ray detector can be used, such as an image intensifier or a flat panel digital detector 40 (as shown in FIG. 3). Indeed, in some embodiments, the X-ray detector comprises a square or a rectangular flat panel detector. FIG. 3 shows some embodiments in which the collimator 30 comprises an X-ray attenuating material 45 that defines an aperture 50. The collimator 30 can comprise any suitable X-ray attenuating material 45 that allows it to collimate an X-ray beam. Some examples of suitable X-ray attenuating materials include tungsten, lead, gold, copper, tungsten-impregnated substrates (e.g., glass or a polymer impregnated with tungsten), coated substrates (e.g., glass or a polymer coated with tungsten, lead, gold, etc.), steel, aluminum, bronze, brass, rare earth metals, or combinations thereof. In some embodiments, however, the collimator comprises tungsten. The collimator 30 collimates an X-ray beam (not shown) so that a resultant image 10 comprises any suitable shape that does not include one or more corners of the X-ray detector 40 that is used to obtain the image. In some embodiments, however, the collimator provides the image with a shape corresponding to a shape of the aperture, wherein the image shape is a superellipse shape or a cornerless shape. A cornerless shape comprises a shape missing one or more 90 degree corners (i.e., two edges that run substantially perpendicular to each other without containing a 90 degree corner between those edges). The cornerless shape may contain corners with a degree less than 90 degrees. Some examples of such shapes include a rounded square, a rounded rectangle, a chamfered square, a chamfered rectangular, a rectangle with curved borders, a truncated circle, an octagon, a hexagon, or any other suitable shape. Where the aperture 50 has the shape of a superellipse, it can have any suitable characteristic that allows the shape of the aperture to be classified as a superellipse (as described above) and that allows the collimator 30 to prevent the X-ray beam from impinging on the corners of the X-ray detector 40. By way of example, the aperture can be a shape that is generated by a formula selected from: (i) (x−a)4+(y−b)4=r4, (ii) \|x−a\|n+\|y−b\|n=\|r\|n, and  ( x - a ) r a  n +  ( y - b ) r b  n ( iii ) wherein a, b is the center point; r is the minor; n is equal to 4; and ra and rb are the semi-major and semi-minor axes, respectively. FIG. 4 shows some embodiments in which the collimator 30 defines an aperture that has a shape of a superellipse. Furthermore, FIG. 4 shows the aperture is sized so that a portion of the collimator 30 overlaps (and thereby collimates) the corners 55 of a corresponding square X-ray detector 25 (e.g., a flat panel detector 40), wherein the perimeter 60 of a receptor area 62 of the detector 25 is illustrated by a dotted line. FIG. 5 shows one example of a collimated image 10 that has been taken with the collimator 30 of FIG. 4. In particular, FIG. 5 shows that image 10 has a shape of a superellipse, in which a first 51 and second 52 images edge, a second 52 and third 53 image edge, a third 53 and fourth 54 edge, and a fourth 54 and first 51 image edge, respectively, do not physically intersect at a 90 degree corner. Instead, FIG. 5 shows the image's corners 56 are trimmed (or collimated) so the image's first edge 51 and third edge 53 are each separated from the image's second edge 52 and fourth edge 54 by a non-linear (i.e., substantially curved) image border 65. Where the aperture 50 comprises the cornerless shape, the aperture can have any suitable characteristic that allows it to function as intended. In one example, FIG. 6 shows the aperture 50 comprises a first 91 and third 93 aperture edge, which run substantially parallel to each other, and which run substantially perpendicular to both a second 92 and a fourth 94 aperture edge. In another example, FIG. 6 shows that instead of physically intersecting at a 90 degree corner, one or more corresponding aperture edges that run perpendicular to each other (e.g., the second 92 and third 93 aperture edges) can be attached to each other with a border 63 that allows the collimator 30 to shield a 90 degree corner 55 of a corresponding detector 40. Additionally, while this disclosure focuses on using an aperture in which all of the apertures edges are substantially equal in length (e.g., an aperture having the appearance of a trimmed square), the skilled artisan will recognize that the aperture could be modified so that any two edges running parallel to each other may be longer or shorter than the other edges of the aperture (e.g., the aperture could have the appearance of a rectangle with trimmed corners). Where the aperture 50 comprises one of the described cornerless shapes, the aperture can be missing any suitable number of corners (e.g., one or more corners of the aperture can be filled in with an X-ray attenuating material), including 1, 2, 3, 4, or more. Indeed, FIG. 6 shows configurations where a portion of the collimator 30 shields two corners 55 (located diagonally from each other) of the X-ray detector's receptor area 62. Accordingly, FIG. 7 shows that an image 10 captured with the configuration of FIG. 6 contains two corners 56 that lack a 90 degree corner between two perpendicular edges of the image (e.g., between the image's second 52 and third 53 edges and between the image's first 51 and fourth 54 edges). Where the aperture 50 is missing one or more corners (e.g., contains an X-ray attenuating material that prevents the X-ray beam from impinging on one or more corners of a corresponding X-ray detector 40), the collimator can collimate the X-ray beam so that the resultant image 10 has any suitably shaped border 63 between adjacent aperture edges that run perpendicular to each other. Some examples of suitable borders include a border with the shape of an arc of a circle, a chamfered border, a rounded border, a convex border, a concave border, a zigzagged border, a curved border, an irregular border, etc. In this regard, FIG. 8 shows that in some embodiments in which all four borders 63 of the aperture 50 (and therefore borders of the image 63) comprise an arc-shaped border 66, the aperture 50 defines a squircle. FIG. 10 shows some configurations in which the each of the aperture's four borders 63 comprises a rounded border 67, the aperture 50 can comprise rounded square (or rectangle where applicable). Additionally, FIG. 12 shows some embodiments in which each of the aperture's four borders 63 comprises a chamfered border 68, the aperture 50 comprises a chamfered square (or rectangle where applicable). Images with shapes corresponding to the collimators 30 of FIGS. 8, 10, and 12 are respectively shown in FIGS. 9, 11, and 13. Additionally, where a border 63 (as described above) separates two substantially perpendicular edges (e.g., 91 and 92, 92 and 93, 93, and 94, and/or 91 and 94) of the aperture 50, the borders can be any suitable shape that allows the collimator to function as described herein. By way of example, FIGS. 14-16 illustrate additional embodiments in which the image 10 has the shape of a squircle, wherein the image 10 in FIG. 14 is predominantly square shaped, the image 10 in FIG. 16 is predominantly circular in shaped, and the image 10 in FIG. 15 has a shape between those shown in FIGS. 14 and 16. Where the collimator 30 shields a portion of the X-ray detector 25 (e.g. one or more of the detector's corners 55), the aperture can leave any suitable amount of the receptor area 62 exposed to X-rays from the X-ray source 20. This configuration allows an image 10 taken with the collimator to be rotated on a display device without being resized or reshaped. In some cases, the collimator allows less than an amount selected from about 100%, about 98.5%, about 94%, about 90%, about 87%, or about 80% of the detector's receptor area to be exposed to X-rays from the X-ray source. In other cases, the collimator allows more than an amount selected from about 78.5%, about 79%, about 80%, about 82%, about 84%, and about 85% of the detector's receptor area to be exposed to X-rays from the X-ray source. In yet other cases, the aperture can allow any suitable combination or sub-range of these amounts of the detector's receptor area to be exposed to X-rays. For example, FIG. 17 shows some embodiments in which the collimator (not shown) allows (from left to right) about 98.2%, about 93.7%, and about 86.1% of the detector's receptor area 62 to be exposed to X-rays (exposed area 105) and in which about 1.8%, about 6.4%, and about 17.9% of the receptor area 62, respectively is shielded by the collimator (unexposed area 110). In other words, FIG. 17 shows the trade-off between an on-screen image size and the detector utilization. The more square the image is, the smaller it has to appear on the screen in order to be rotatable. In some embodiments of the squircles described herein, the geometry could range between a full square (100% of the detector utilized) and a full circle (78.5% utilization). Returning to the method 150 in FIG. 2, after an image 10 has been taken of an object (as shown at 160), the method continues at box 165, where the collimated X-ray image is shown on a display device (e.g., a screen, monitor, tablet/handheld device, etc.). The image can take up any suitable amount of the display device's display area that allows the entire image to be viewed as it is rotated at least 45 degrees about its center, without being resized or reshaped. The height H (e.g., the distance between the first 51 and third 53 or second 52 and fourth 54 edges) of the image 10 can be any height that allows the entire image to be rotated on the display device 115 without the image being resized or reshaped. In some embodiments, the height H is greater than an amount selected from about 71.6%, about 75%, about 80%, or about 82.5% of the narrower of the width and length of the display area. In other embodiments, the height H of the image is less than an amount selected from about 100%, about 98%, about 95%, and about 90% of the height of the display area. In yet other embodiments, the image's height H falls between any suitable combination or sub-range of these amounts. For example, FIG. 17 shows that where the display device 115 comprises an HD display device (e.g., a device having a pixel resolution of 1080 pixels by 1920 pixels), the image 10 can have a height of about 841 pixels (where about 98.2% of the detector's receptor area is exposed), about 921 pixels (where about 93.7% of the detector's receptor area 62 is exposed), or about 1012 pixels (where about 78.5% of the detector's receptor area is exposed to X-rays). Because some embodiments of the aperture 50 can have any shape between a full circle (e.g., in which about 78.5% of a square detector are is utilized) and a true square (e.g., in which about 100% of the square detector is utilized), the widest portion D (e.g., a diagonal measurement) of the collimated image can be any length that allows the entire image to be rotated on the display device 115 without the image being resized or reshaped. In some embodiments, the widest portion D of the image is less than an amount selected from about 100%, about 99%, and about 96% of the width or length of the display's display area, whichever is narrower. In other embodiments, the widest portion D of the image is greater than an amount selected from about 85%, about 90%, and about 95% of the width or length of the display area, whichever is narrower. In still other embodiments, the widest portion D of the image can be between any suitable combination or sub-range of these amounts. In some instances, the shape of the aperture 50 helps provide a desired balance between the on-screen image size of the image 10 and detector utilization. By way of illustration, FIG. 17 shows that, in some cases, the more square the image 10 is, the smaller it has to be on the display device 115 in order to be entirely seen as it is rotated. In contrast, where the aperture has borders 63 that are arcs of a true circle, thereby providing a squircle image, the entire squircle can be rotated on the display device without clipping any part of the image and without rescaling the shape as long as the true circle could be fully displayed on the device. Returning to FIG. 2, the method 150 continues at box 165 where the collimated image is optionally shown on a display device 115. At box 170, the method 150 optionally includes a process of rotating the image clockwise and/or counterclockwise. FIGS. 18 through 20 show successive views of the image 10 being rotated counter-clockwise on a display device 115. As the image 10 is rotated, the entire image can be viewed on the display device 115, without any resizing or reshaping of the image. Thus, in some embodiments, the exposed area of the live image (and not the processed image) can be substantially equal to the displayed area. Where the collimated X-ray images 10 are shown, rotated, or otherwise manipulated on a display device 115, the display device can be used with any suitable computing environment. FIG. 21 describes some embodiments of one exemplary computing environment. These embodiments can include one or more processing units in a variety of customizable enterprise configurations, including in a networked or combination configuration. These embodiments can include one or more computer readable media, wherein each medium may be configured to include or includes thereon data or computer executable instructions for manipulating data. The computer executable instructions can include data structures, objects, programs, routines, or other program modules that may be accessed by one or more processors, such as one associated with a general-purpose modular processing unit capable of performing various different functions or one associated with a special-purpose modular processing unit capable of performing a limited number of functions. Computer executable instructions cause the one or more processors of the enterprise to perform a particular function or group of functions and are examples of program code means for implementing steps for methods of processing. Furthermore, a particular sequence of the executable instructions provides an example of corresponding acts that may be used to implement such steps. Examples of computer readable media (including non-transitory computer readable media) include random-access memory (“RAM”), read-only memory (“ROM”), programmable read-only memory (“PROM”), erasable programmable read-only memory (“EPROM”), electrically erasable programmable read-only memory (“EEPROM”), compact disk read-only memory (“CD-ROM”), any solid state storage device (e.g., flash memory, smart media, etc.), or any other device or component capable of providing data or executable instructions that may be accessed by a processing unit. With reference to FIG. 21, a representative enterprise includes modular processing unit 200, which may be used as a general-purpose or special-purpose processing unit. For example, modular processing unit 200 may be employed alone or with one or more similar modular processing units as a personal computer, a notebook computer, a personal digital assistant (“PDA”) or other hand-held device, a workstation, a minicomputer, a mainframe, a supercomputer, a multi-processor system, a network computer, a processor-based consumer device, a cellular phone, a smart appliance or device, a control system, or the like. Using multiple processing units in the same enterprise provides increased processing capabilities. For example, each processing unit of an enterprise can be dedicated to a particular task or can jointly participate in distributed processing. In FIG. 21, the modular processing unit 200 includes one or more buses and/or interconnects 205, which may be configured to connect various components thereof and enables data to be exchanged between two or more components. The bus(es)/interconnect(s) 205 may include one of a variety of bus structures, including a memory bus, a peripheral bus, or a local bus that uses any of a variety of bus architectures. Typical components connected by the bus(es)/interconnect(s) 205 include one or more processors 210 and one or more memories 215. Other components may be selectively connected to the bus(es)/interconnect(s) 205 through the use of logic, one or more systems, one or more subsystems and/or one or more I/O interfaces, hereafter referred to as data manipulating system(s) 220. Moreover, other components may be externally connected to the bus(es)/interconnect(s) 205 through the use of logic, one or more systems, one or more subsystems and/or one or more I/O interfaces, and/or may function as logic, one or more systems, one or more subsystems, and/or one or more I/O interfaces, such as one or more modular processing unit(s) 245 and/or proprietary device(s) 255. Examples of I/O interfaces include one or more mass storage device interfaces, one or more input interfaces, one or more output interfaces, and the like. Accordingly, embodiments of the described systems and methods embrace the ability to use one or more I/O interfaces and/or the ability to change the usability of a product based on the logic or other data manipulating system employed. The logic may be tied to an interface, part of a system, subsystem and/or be used to perform a specific task. Accordingly, the logic or other data manipulating system may allow, for example, for IEEE1394 (firewire), wherein the logic or other data manipulating system is an I/O interface. Alternatively or additionally, logic or another data manipulating system may be used that allows a modular processing unit to be tied into another external system or subsystem. For example, an external system or subsystem that may or may not include a special I/O connection. Alternatively or additionally, logic or another data manipulating system may be used wherein no external I/O is associated with the logic. Embodiments of the described systems and methods also embrace the use of specialty logic, such as for ECUs for vehicles, hydraulic control systems, etc. and/or logic that informs a processor how to control a specific piece of hardware. Moreover, those skilled in the art will appreciate that embodiments of the described systems and methods embrace a plethora of different systems and/or configurations that utilize logic, systems, subsystems and/or I/O interfaces. As provided above, embodiments of the described systems and methods embrace the ability to use one or more I/O interfaces and/or the ability to change the usability of a product based on the logic or other data manipulating system employed. For example, where a modular processing unit is part of a personal computing system that includes one or more I/O interfaces and logic designed for use as a desktop computer, the logic or other data manipulating system can be changed to include flash memory or logic to perform audio encoding for a music station that wants to take analog audio via two standard RCAs and broadcast them to an IP address. Accordingly, the modular processing unit may be part of a system that is used as an appliance rather than a computer system due to a modification made to the data manipulating system(s) (e.g., logic, system, subsystem, I/O interface(s), etc.) on the back plane of the modular processing unit. Thus, a modification of the data manipulating system(s) on the back plane can change the application of the modular processing unit. Accordingly, embodiments of the described systems and methods embrace very adaptable modular processing units. As provided above, processing unit 200 includes one or more processors 210, such as a central processor (or CPU) and optionally one or more other processors designed to perform a particular function or task. It is typically the processor 210 that executes the instructions provided on computer readable media, such as on the memory(ies) 215, a magnetic hard disk, a removable magnetic disk, a magnetic cassette, an optical disk, or from a communication connection, which may also be viewed as a computer readable medium. The memory(ies) 215 includes one or more computer readable media that may be configured to include or includes thereon data or instructions for manipulating data, and may be accessed by the processor(s) 210 through the bus(es)/interconnect(s) 205. The memory(ies) 215 may include, for example, ROM(s) 225, used to permanently store information, and/or RAM(s) 226, used to temporarily store information. The ROM(s) 225 may include a basic input/output system (“BIOS”) having one or more routines that are used to establish communication, such as during start-up of the modular processing unit 200. During operation, the RAM(s) 226 may include one or more program modules, such as one or more operating systems, application programs, and/or program data. As illustrated, at least some embodiments of the described systems and methods embrace a non-peripheral encasement, which provides a more robust processing unit that enables use of the unit in a variety of different applications. In FIG. 21, one or more mass storage device interfaces (illustrated as data manipulating system(s) 220) may be used to connect one or more mass storage devices 230 to the bus(es)/interconnect(s) 205. The mass storage devices 230 are peripheral to the modular processing unit 200 and allow the modular processing unit 200 to retain large amounts of data. Examples of mass storage devices include hard disk drives, magnetic disk drives, tape drives and optical disk drives. A mass storage device 230 may read from and/or write to a magnetic hard disk, a removable magnetic disk, a magnetic cassette, an optical disk, or another computer readable medium. The mass storage devices 230 and their corresponding computer readable media provide nonvolatile storage of data and/or executable instructions that may include one or more program modules, such as an operating system, one or more application programs, other program modules, or program data. Such executable instructions are examples of program code means for implementing steps for methods disclosed herein. The data manipulating system(s) 220 may be employed to enable data and/or instructions to be exchanged with the modular processing unit 200 through one or more corresponding peripheral I/O devices 235. Examples of the peripheral I/O devices 235 include input devices such as a keyboard and/or alternate input devices, such as a mouse, trackball, light pen, stylus, or other pointing device, a microphone, a joystick, a game pad, a satellite dish, a scanner, a camcorder, a digital camera, a sensor, and the like, and/or output devices such as a display device 115 (e.g., a monitor or display screen), a speaker, a printer, a control system, and the like. Similarly, examples of the data manipulating system(s) 220 coupled with specialized logic that may be used to connect the peripheral I/O devices 235 to the bus(es)/interconnect(s) 205 include a serial port, a parallel port, a game port, a universal serial bus (“USB”), a firewire (IEEE 1394), a wireless receiver, a video adapter, an audio adapter, a parallel port, a wireless transmitter, any parallel or serialized I/O peripherals or another interface. The data manipulating system(s) 220 enable an exchange of information across one or more network interfaces 240. Examples of the network interfaces 240 include a connection that enables information to be exchanged between processing units, a network adapter for connection to a local area network (“LAN”) or a modem, a wireless link, or another adapter for connection to a wide area network (“WAN”), such as the Internet. The network interface 240 may be incorporated with or peripheral to modular processing unit 200, and may be associated with a LAN, a wireless network, a WAN and/or any 260 connection (see FIG. 22) between processing units. The data manipulating system(s) 220 enables the modular processing unit 200 to exchange information with one or more other local or remote modular processing units 245 or computer devices. A connection between modular processing unit 200 and modular processing unit 245 may include hardwired and/or wireless links. Accordingly, embodiments of the described systems and methods embrace direct bus-to-bus connections. This enables the creation of a large bus system. It also eliminates hacking as currently known due to direct bus-to-bus connections of an enterprise. Furthermore, the data manipulating system(s) 220 enable the modular processing unit 200 to exchange information with one or more proprietary I/O connections 250 and/or one or more proprietary devices 255. Program modules or portions thereof that are accessible to the processing unit may be stored in a remote memory storage device. Furthermore, in a networked system or combined configuration, the modular processing unit 200 may participate in a distributed computing environment where functions or tasks are performed by a plurality of processing units. Alternatively, each processing unit of a combined configuration/enterprise may be dedicated to a particular task. Thus, for example, one processing unit of an enterprise may be dedicated to video data, thereby replacing a traditional video card, and provides increased processing capabilities for performing such tasks over traditional techniques. While those skilled in the art will appreciate that the described systems and methods may be practiced in networked computing environments with many types of computer system configurations, FIG. 22 represents an embodiment of a portion of the described systems in a networked environment that includes clients (265, 270, 275, 280, etc.) connected to a server 285 via a network 260. While FIG. 22 illustrates an embodiment that includes four clients connected to the network, alternative embodiments include one client connected to a network or many clients connected to a network. Moreover, embodiments in accordance with the described systems and methods also include a multitude of clients throughout the world connected to a network, where the network is a wide area network, such as the Internet. Accordingly, in some embodiments, the described systems and methods can allow a collimated image 10 to be taken in a first location and a user (e.g., a radiologist, technician, physician, etc.) to view, rotate, and otherwise manipulate the image from a second location. As previously mentioned, the described systems and methods can be modified in any suitable manner. In one example, where computer software is used to display the described collimated images 10 on a display device, the software can be used to clean up the images in any suitable manner. For instance, the software can be used to remove shadows, fuzzy lines, or to otherwise sharpen the image's edges. The described systems and methods for displaying collimated X-ray images 10 have several useful features. First, unlike some conventional methods that use a collimator to shield a relatively large amount of the detector's receptor area, some embodiments of the described systems and methods shield a relatively small amount of the detector's receptor area (as discussed above). Thus, some conventional methods are limited to using a collimator having a circular aperture with a circumference that falls completely within all of the perimeters of a four-sided flat panel detector. As a result, a relatively large amount of the receptor area in such conventional methods is not used. Second, while some conventional methods shrink an X-ray image as the image is rotated, some embodiments of the described systems and methods allow the image to be relatively large with respect to the display's display area and to be rotated while maintaining a substantially constant size and shape. And third, unlike some conventional methods for displaying an X-ray image that only show a small square image that can be rotated without being resized or reshaped, some embodiments of the described methods allow the described images 10 to use a relatively large amount of the display's display area without needing any resizing or reshaping. Thus, users of the described systems can see better detail on the collimated images than may be obtained through some other conventional methods. In addition to any previously indicated modification, numerous other variations and alternative arrangements may be devised by those skilled in the art without departing from the spirit and scope of this description, and appended claims are intended to cover such modifications and arrangements. Thus, while the information has been described above with particularity and detail in connection with what is presently deemed to be the most practical and preferred aspects, it will be apparent to those of ordinary skill in the art that numerous modifications, including, but not limited to, form, function, manner of operation, and use may be made without departing from the principles and concepts set forth herein. Also, as used herein, the examples and embodiments, in all respects, are meant to be illustrative only and should not be construed to be limiting in any manner.
H00006890	claims	1. A fuel pin for a liquid metal nuclear reactor comprising: a generally cylindrical cladding member; and metallic fuel material disposed within said cladding member, with at least a portion of said fuel material extending radially outwardly to the inner diameter of said cladding member, said fuel material defining at least one void space to facilitate swelling of said fuel material during fission. a protective housing; core region with fissile material therein, said core region being disposed within said housing; control means for absorbing neutrons disposed within said housing; a coolant system for circulating coolant liquid adjacent said core region; and a plurality of fuel pins, with at least a portion of each of said fuel pins being disposed within said core region, said fuel pins including metallic fuel material extending radially outwardly substantially all the way to the radial periphery of said fuel pin, said fuel material defining at least one internal void, radially inwardly spaced from the radial periphery of said fuel pin. 2. The fuel pin of claim 1, wherein said void space is filled with tag gas to permit sensing of a leak in the fuel pin. 3. The fuel pin of claim 1, wherein said void space is disposed radially inwardly spaced from said cladding, and the entire radial periphery of said fuel material extends to said cladding member. 4. The fuel pin of claim 2, wherein said void space is defined by a plurality of axially extending flutes disposed adjacent the radial periphery of said fuel material, adjacent said cladding. 5. The fuel pin of claim 1, wherein said void space includes a radially centrally disposed, axially extending channel. 6. The fuel pin of claim 1, wherein a plurality of said void spaces are spaced throughout said fuel material, with more of said void spaces being disposed toward the axial center of said fuel material than toward the axial ends thereof. 7. The fuel pin of claim 1, wherein a plurality of said void spaces are included, said void spaces being minute and spaced throughout said fuel material. 8. The fuel pin of claim 7, wherein more of said void spaces are disposed toward the axial center of said fuel material than toward the axial ends thereof. 9. The fuel pin of claim 1, further comprising a first plenum defined within said cladding member for receiving fission gases from said fuel material, and a first lock plug means disposed between said first plenum and said fuel material, for providing axial pressure against said fuel material and for permitting leakage of fission gases from said fuel material to said first plenum. 10. A fuel pin of claim 1, further comprising blanket material being disposed immediately axially adjacent each axial end of said fuel material. 11. The fuel pin of claim 1, wherein said fuel material is comprised of powder. 12. The fuel pin of claim 1, wherein said fuel material is comprised of minute spheroids. 13. The fuel pin of claim 1, further comprising a temperature sensitive tag gas capsule disposed within said cladding member, for releasing tag gas into said cladding member during reactor operation, which capsule includes an elongate tube having a top end and a bottom end and having a top end cap and a bottom end cap fixed to said top and bottom ends with at least one of said end caps being rupturable, and an axially extending, centrally disposed penetrator member which has a higher thermal coefficient of expansion than said tube, such that upon reactor temperature increase said penetrator member expands at a faster rate than said tube, causing rupture of said one end cap. 14. The fuel pin of claim 13, wherein said penetrator member is an elongate rod having one end fixedly mounted centrally to said top end cap, and wherein said tube is fixedly mounted to said bottom end cap, and wherein said bottom end cap includes a centrally disposed rupturable inner well with the other end of said rod extending within said well. 15. The fuel in of claim 14, wherein said inner well comprises an undercut portion of said bottom end cap. 16. The fuel in of claim 15, wherein the width of said inner well is at least as large as the width of said rod. 17. The fuel pin of claim 16, wherein the external configuration of said inner well is complementary to the external configuration of said penetrator member. 18. The fuel pin of claim 13, which further comprises a tag gas disposed in said capsule. 19. A liquid metal reactor comprising: 20. The reactor of claim 19, wherein said fuel pin has a pair of axial ends and a plurality of void spaces are included with a greater percentage of total void space being disposed toward an axial center of said fuel material than toward the axial ends thereof. 21. The reactor of claim 19, wherein at least one of said fuel pins further includes a temperature sensitive tag gas capsule disposed within said one fuel pin for releasing tag gas into said one pin during reactor operation, which capsule includes an elongate tube having a top end and a bottom end and having a top end cap and a bottom end cap fixed to said top and bottom ends with at least one of said end centrally disposed penetrator member mounted in one of said end caps, which penetrator member has a higher thermal coefficient of expansion than said tube, such that upon reactor temperature increase said penetrator member expands at a faster rate than said tube, causing rupture of one of said end caps.
	claims	1. A method of operating a nuclear reactor having a reactor core in which a plurality of reload fuel assemblies respectively having different infinite multiplication factors are arranged, said reload fuel assemblies including small-number cycle reload fuel assemblies having a larger infinite multiplication factor than a large-number operation cycle reload fuel assemblies having a smaller infinite multiplication factor than the larger infinite multiplication factor of said small-number cycle reload fuel assemblies, the small-number cycle reload fuel assemblies having been operated for a smaller number of operation cycles than the number of cycles of operation of said large-number cycle reload fuel assemblies, said method comprising: inserting control rods alternately in control cells of a first group and of a second group with respect to time, each control cell of the first group and the second group comprising four small-number cycle reload fuel assemblies having the larger infinite multiplication factor for a period longer than half of a period of an operation cycle. 2. A method of operating a nuclear reactor having a reactor core in which a plurality of reload fuel assemblies respectively having different infinite multiplication factors and a plurality of initial loading fuel assemblies comprising high-enrichment initial loading fuel assemblies and low-enrichment initial loading fuel assemblies with a lower enrichment factor than an enrichment factor of the high-enrichment initial loading fuel assemblies are arranged, said method comprising: inserting control rods alternately in control cells of a first group and of a second group with respect to time, each control cell of the first group and the second group comprising four fuel assemblies including at least one reload fuel assembly and a plurality of high-enrichment initial loading fuel assemblies for a period longer than half of a period of an operation cycle. 3. The method of operating a nuclear reactor according to claim 1 or 2 , wherein the control cells in which the control rods are inserted include new fuel assemblies loaded into the reactor core for the operation cycle. claim 1 2 4. The method of operating a nuclear reactor according to claim 1 or 2 , wherein the control rods are inserted in the control cells of each group at time intervals corresponding to a burn-up of several gigawatts day per ton. claim 1 2 5. The method of operating a nuclear reactor according to claim 4 , wherein the fuel assemblies other than those forming the control cells in which the control rods are inserted are removed from the reactor core after the completion of the operation cycle. claim 4 6. The method of operating a nuclear reactor according to claim 4 , wherein most of the fuel assemblies arranged in an outer peripheral region of the reactor core are those having small infinite multiplication factors. claim 4 7. The method of operating a nuclear reactor according to claim 1 , wherein the fuel assemblies other than those forming the control cells in which the control rods are inserted are removed from the reactor core after the completion of the operation cycle. claim 1 8. The method of operating a nuclear rector according to claim 2 , wherein the fuel assemblies other than those forming the control cells in which the control rods are inserted are removed from the reactor core after the completion of the operation cycle. claim 2 9. The method of operating a nuclear reactor according to claim 3 , wherein the fuel assemblies other than those forming the control cells in which the control rods are inserted are removed from the reactor core after the completion of the operation cycle. claim 3 10. The method of operating a nuclear reactor according to claim 1 , wherein most of the fuel assemblies arranged in an outer peripheral region of the reactor core are those having small infinite multiplication factors. claim 1 11. The method of operating a nuclear reactor according to claim 2 , wherein most of the fuel assemblies arranged in an outer peripheral region of the reactor core are those having small infinite multiplication factors. claim 2 12. The method of operating a nuclear reactor according to claim 3 , wherein most of the fuel assemblies arranged in an outer peripheral region of the reactor core are those having small infinite multiplication factors. claim 3 13. A method of operating a nuclear reactor having a reactor core in which a plurality of reload fuel assemblies comprising small-number cycle reload fuel assemblies and large-number cycle reload fuel assemblies having been operated for a larger number of operation cycles than a number of operation cycles of the small-number cycle reload fuel assemblies are arranged, said method comprising: inserting control rods alternately in control cells of a first group and of a second group with respect to time, each control cell of the first group and the second group comprising four reload fuel assemblies of the small-number cycle reload fuel assemblies for a period longer than half of a period of an operation cycle. 14. A method of operating a nuclear reactor having a reactor core in which a plurality of reload fuel assemblies and a plurality of initial loading fuel assemblies comprising high-enrichment initial loading fuel assemblies and low-enrichment initial loading fuel assemblies with a lower enrichment factor than an enrichment factor of the high-enrichment initial loading fuel assemblies are arranged, said method comprising: inserting control rods alternately in control cells of a first group and those of a second group with respect to time, each control cell of the first group and the second group comprising four fuel assemblies formed of at least one of the reload fuel assemblies and a plurality of the high-enrichment initial loading fuel assemblies for a period longer than half of a period of an operation cycle. 15. The method of operating a nuclear reactor according to claim 1 or 13 , wherein the number of operation cycles of the small-number cycle reload fuel assemblies is not greater than three and the number of operation cycles of the large-number cycle reload fuel assemblies is greater than three. claim 1 13 16. The method of operating a nuclear reactor according to claim 2 or 14 , wherein each control cell of the first group and the second group comprise no more than two of the reload fuel assemblies. claim 2 14
055925227	summary	TECHNICAL FIELD The present invention relates to a control rod for a nuclear reactor and particularly to a cruciform-shaped control rod having each wing divided into discrete compartments formed by structural stiffeners and which compartments contain neutron-absorbing material. BACKGROUND As well known, control rods in a nuclear reactor perform the functions of power distribution, shaping and reactivity control. This is accomplished generally by providing a plurality of control rods containing material for neutron absorption and manipulating the control rods within the reactor. Generally, a control rod used in a boiling water reactor (BWR) is provided in a cruciform shape and extends in complementary cruciform-shaped interstices between fuel channels. While there are a number of different control rod designs, two basic control rod designs have been typically employed previously in boiling water reactors. In one design, each control rod consists of a plurality, e.g., either 48 or 84, boron carbide absorber or hafnium rods, or a combination of boron carbide and hafnium rods, separated into four cruciform-shaped wings of either 12 or 21 tubes each. The tubes are enclosed in a perforated outer sheath which is welded to a full-length central tie rod which has a handle and connector/velocity limiter at the respective opposite ends of the rod. The perforated sheath affords a cooling medium to the tubes. If a tube fails in this design, the boron carbide powder is exposed to the reactor coolant. In another type of control rod, essentially square tubes are welded to one another to form the four wings of the cruciform-shaped control rods. Each wing contains 12 to 15 square tubes welded to one another and to a handle at one end and a connector/velocity limiter at the opposite end. The welds between the tubes and the tubes themselves provide the structural support for the control rod. The tubes serve as individual pressure vessels, as well as structural members subjected to all reactor induced loadings. In this design, the boron carbide powder is contained in sealed capsules inserted into the individual square tubes. One of the difficulties with the latter type of control rod construction is the magnitude of the welding required to weld each of the individual square tubes to an adjacent tube to provide the structural support necessary for the four wings of the control rod. While that construction has served well, it is quite expensive to manufacture. Another problem associated with the design of control rods is that the reaction of the neutron-absorbing material, e.g., boron carbide, with neutrons produces a helium gas. Thus, the tubes of each of the previously noted designs must have structural integrity to maintain the generated helium gas within the tube at increasing pressure throughout the lifetime of the control rod. It is known that the mechanical lifetime of a control rod is limited by the reactor burn-up and the corresponding helium pressure build-up in the tube containing the boron carbide exposed to the highest neutron flux. The magnitude and pressure of helium gas generated depends on the location of the tube within the control rod. That is, the boron carbide containing tube which has the highest exposure to neutrons in the control rod determines the mechanical life of the control rod. However, in each prior design, the neutron-absorbing material was contained in discrete tubes individually subjected to helium gas pressure build-up and without any relief. Consequently, the life of the control rod is dependent upon the structural integrity of only one of a large number of absorber tubes. DISCLOSURE OF THE INVENTION In accordance with the present invention, there is provided a control rod construction which requires considerably reduced manufacturing time and expense as compared with prior control rod designs, principally by reducing the amount of welding required to form the control rod. Additionally, the invention extends the mechanical life of the control rod by equalization of the pressures within the absorber tubes across each wing of the control rod, hence limiting the helium gas pressure within the control rod to an average pressure of the individual absorber tubes. Particularly, the present invention provides a control rod in cruciform shape wherein each wing of the cruciform-shaped control rod includes discrete compartments sealed from the surrounding environment. For example, each wing of the cruciform-shaped control rod includes laterally spaced, vertically extending structural or stiffening members interconnected at their opposite ends by generally horizontally extending structural or stiffening members defining a plurality of compartments. Neutron-absorbing material is disposed in each of the compartments. For example, hafnium rods or capsules containing boron carbide, or a combination of hafnium rods and boron carbide capsules, may be disposed in each of the compartments. The compartments are sealed on opposite sides by overlays of sheet metal secured to the vertical and horizontally extending stiffening members, thus sealing the compartments from the surrounding environment. In another aspect of the present invention, the compartments in each wing of the cruciform-shaped control rod communicate one with the other to afford an equalization of pressure within the entire wing of the control rod. More particularly, the capsules containing the boron carbide are designed to leak the helium generated by the boron-neutron reaction into the compartments. By providing openings or passages through the vertical members or stiffeners defining adjacent compartments, the helium leakage migrates between compartments, equalizing the pressure. As a consequence, variations in the generation of helium gas in the individual boron carbide capsules can be accommodated, thus extending the mechanical life of the control rod by providing an average pressure for all of the boron carbide capsules. In a preferred embodiment according to the present invention, there is provided a control rod for a nuclear reactor, comprising a cruciform control rod body having four elongated, substantially vertically extending wings arranged generally 90.degree. relative to one another, each wing including a plurality of elongated, generally vertically extending structural members spaced from one another and horizontally extending structural members adjacent opposite ends of the vertically extending members defining a plurality of vertically extending, side-by-side compartments, a sheath disposed along opposite sides of the vertically and horizontally extending members whereby the members and the sheath seal the compartments externally of the wing and a neutron-absorbing material disposed in each sealed compartment. In a further preferred embodiment according to the present invention, there is provided a control rod for a nuclear reactor comprising a control rod body having a plurality of elongated, laterally spaced, generally parallel structural members in part defining a plurality of side-by-side sealed compartments each containing neutron-absorbing material, and an opening in at least one of the members in part defining adjacent compartments enabling passage of gas under pressure generated by reaction of the neutron-absorbing material with neutrons from one of the adjacent compartments through the opening to another of the adjacent compartments to maintain the adjacent compartments under substantially equal pressure. Accordingly, it is a primary object of the present invention to provide a novel and improved control rod which can be readily, easily and inexpensively manufactured and which may extend the mechanical life of the control rod.
049842587	description	FIG. 1A is a diagrammatic cross-sectional view of a slit diaphragm 1 having a slit S through which passes a substantially planar, fan-shaped X-ray beam B originating from an X-ray source 2 in operation in the manner as conventional in slit radiography. Extending into the X-ray beam B, in operation, to a greater or lesser extent, is the free end of an elongate attenuation element 3 associated with a plurality of juxtaposed similar attenuation elements. Each attenuation element is adapted, in operation, to swivel relatively to a fixed point 5 according to an arrow 4 under the influence of control means not shown. Suitable control means are described in the copending Dutch patent application 8400845. Said control means do not form part of the present invention and will therefore not be further described herein. In the embodiment shown in FIG. 1, each attenuation element comprises a connecting rod 6 extending approximately transversely to the longitudinal direction of the attenuation element, the end of said rod distal from the attenuation element being connected to a piston 7 placed within a cylinder 8. Said cylinder is filled with a fluid in known manner, e.g. oil, braking the movement of the piston in the cylinder to some extent, so that oscillation or slipping of the attenuation element is prevented. FIG. 1B shows a similar damping member used with a slide-like attenuation element 3'. The embodiments of the present invention to be described hereinafter can be employed accordingly both for slide-like and for elongate swivelling attenuation elements. In the embodiment of the present invention shown in FIGS. 2 and 3, each attenuation element comprises a vane 9 extending into a vessel 11 fitted with a suitable liquid 10. As damping liquid may be used e.g. oil, alcohol or water. The vane is made of rigid material and may be corrugated, as indicated by lines 9a. The vane may be connected to the attenuation element by gluing and is preferably made of synthetic plastics material. The vane may also be connected to the attenuation element by means of a connecting rod. Such an embodiment is particularly suitable for slide-like attenuation elements. Vessel 11 may be made of any suitable material. FIG. 3 shows the arrangement of FIG. 2 in front view, with the front wall of the vessel 11 being omitted. The attenuation elements 3a-3f are each provided with a vane 9a-9f extending into the liquid 10 in the vessel 11. In order to prevent the vanes from influencing each other during movements of the associated attenuation elements, preferably parallel partitions 12-16 are disposed in the vessel, so that each vane extends into its own compartment of the vessel. Partitions 12-16, moreover, increase the damping effect. The extent of damping may be influenced by the choice of the viscosity of the liquid in the vessel, the height of the liquid level in the vessel and the form of the vane. If desired, the vessel can be filled with a plurality of layers of various poorly miscible liquids. FIG. 4 shows a variant of FIGS. 2 and 3, wherein each vane is made of synthetic plastics material and is provided on both sides partly with a layer of electrically conductive material 17a-17d. For the sake of clearness, only two attenuation elements with vanes are shown. Adjacent the layers of conductive material, there is generated an inhomogeneous magnetic field oriented transversely to said vanes by means of magnet poles N and S provided on both sides of the vessel 11. The damping is obtained by eddy current losses occurring in the layers 17a-17d, as soon as said layers 17a-17d move relatively to the magnetic field (Foucault effect). The vessel 11 in this case is preferably made of copper. The damping method by means of eddy current losses can be employed separately or, as shown in FIG. 4, in combination with a vessel containing liquid, also having a damping effect. It is observed that various modifications will readily occur to one skilled in the art after the foregoing. For instance, it is possible to provide the attenuation elements with electrically conductive metal vanes wherein eddy current losses occur under the influence of an inhomogeneous magnetic field. In that case, the layers 17 can be omitted from the embodiment shown in FIG. 4. Use could also be made of a vane of electrically non-conductive material provided with a conductive layer on one side only. Such modifications are deemed not to depart from the scope of the present invention.
046631186	summary	BACKGROUND Nuclear power reactors are well known and are discussed, for example, by M. M. El-Wakil in "Nuclear Power Engineering", McGraw-Hill Book Company, Inc., 1962. In a known type of nuclear power reactor, for example, as used in the Dresden Nuclear Power Station near Chicago, Ill., the reactor core is of the heterogenous type. In such reactors the nuclear fuel comprises elongated rods formed of sealed cladding tubes of suitable material, such as a zirconium alloy, containing uranium oxide and/or plutonium oxide as the nuclear fuel, for example, as shown in U.S. Pat. No. 3,365,371. A number of such fuel rods are grouped together and contained in an open-ended tubular flow channel to form a separately removable fuel assembly or bundle. A sufficient number of fuel assemblies are arranged in a matrix, approximating a right circular cylinder, to form the nuclear reactor core capable of self-sustained fission reaction. The core is submerged in a fluid, such as light water, which serves both as a coolant and as a neutron moderator. A well-known and widely used type of fuel assembly is shown by C. R. Mefford et al in U.S. Pat. No. 3,697,376. Such a fuel assembly is formed by an array of spaced fuel rods supported between upper and lower tie plates, the rods being several feet in length, on the order of one-half inch in diameter and spaced from one another by a fraction of an inch. As shown in U.S. Pat. No. 3,697,376, the lower tie plate is formed integrally with a tapered nose piece or nozzle which engages a fuel assembly support socket of the core support structure of the nuclear reactor. The nose piece is formed with openings for receiving pressurized fluid coolant and directing it upward past the fuel rods. To contain the coolant flow, the fuel assembly is surrounded by an open ended coolant flow channel. The lower open end of the flow channel is an unrestrained slip fit over the lower tie plate/nozzle which results in the problem of control of coolant leakage addressed in U.S. Pat. No. 3,697,376. Other fuel assembly arrangements are known, such as shown by Suvanto et al in U.S. Pat. No. 3,697,375, wherein the lower tie plate and the nozzle are formed as separate pieces and the lower end of the flow channel is permanently attached to the nozzle. This arrangement solves the coolant leakage problem and can provide other advantages as discussed in U.S. Pat. No. 3,697,375. The flow channel typically is formed of a zirconium alloy (to minimize neutron absorption) while the nozzle typically is formed of stainless steel. Thus welding of the flow channel to the nozzle as a method of attachment is not practical. Therefore, attachment of the flow channel to the nozzle has been accomplished by the use of rivets or screws as shown and described in U.S. Pat. No. 3,697,375. However, such attachment by rivets or screws has not been entirely satisfactory. This is because differential thermal expansion of the parts can result in over-stressing and consequent failure (or at least loosening) of the rivets or screws. Furthermore, if the flow channel is firmly attached to the nozzle (as with screws or rivets), the greater expansion and contraction of the stainless steel nozzle with changes in temperatures, as compared to the zirconium alloy flow channel, can cause bending and stressing of the lower end of the flow channel. Therefore, an object of the invention is a channel-to-nozzle attachment which substantially avoids stressing of the parts with differential thermal expansion. SUMMARY This and other objects of the invention are achieved by a channel-to-nozzle attachment in the form of tapered attachment bars or blocks (formed of material similar to that of the channel, e.g. zirconium alloy) and secured to the lower inside surfaces of the channels. These attachment bars are fitted into mating tapered grooves in the outside surfaces of the nozzle which the lower end of the channel surrounds, the nozzle being formed of material (e.g. stainless steel) having a different (e.g. higher) thermal coefficient of expansion. The angle of the mating taper between the attachment bars and the grooves in the nozzle is selected such that the fit between the bars and the grooves is maintained, for temperature changes from room temperature to operating temperature in a nuclear reactor, without bending or stressing of the lower end of the channel. As the channel, the attachment bars and the nozzle expand with increased temperature, the greater expansion of the nozzle is accommodated simply by movement of the attachment blocks further into the grooves, an initial clearance between the faces of the blocks and the bottoms of the grooves being provided to allow such movement.
	summary
	abstract	A method and apparatus for modifying an existing nuclear reactor moveable in-core detector system to insert and withdraw target specimens from a reactor core during reactor operation without practically impeding the moveable in-core detector system's ability to obtain flux maps of the core throughout the reactor's operation. The apparatus provides a separate drive unit and delivery cable that is independent of the detector drive system, but uses most of the same core delivery conduits to access the core. A specimen holder is remotely detachable from the delivery cable when appropriately positioned and can be remotely reattached for withdrawal after a scheduled period of radiation.
059636114	summary	BACKGROUND OF THE INVENTION FIELD OF THE INVENTION The present invention relates to a steam separator, a nuclear power generation plant, and a boiler apparatus equipped with a separator/injector having a two-phase flow accelerator nozzle for guiding a two-phase flow of mixed liquid and vapor components in the interior thereof and accelerating the same, a liquid-phase capture means for capturing the liquid phase of the thus accelerated two-phase flow, and means for increasing the pressure of this liquid phase and imparting a recirculation drive force thereto. The description herein relates to nuclear power plants in general, taking a boiling-water reactor as an example with reference to the accompanying drawings. Referring to FIG. 27, a reactor container 106 of a boiling-water reactor (BWR) is configured of a reactor pressure vessel 102 accommodating a core 101, a drywell 103 that contains this reactor pressure vessel 102, and a wetwell 105 having a pressure suppression pool 104. In addition, this nuclear power plant comprises a turbine 107, a main steam line 108 that supplies steam to that turbine 107, a main condenser 109, a condensate pump 110, a feedwater pump 111 that supplies feedwater to the reactor pressure vessel 102, a feedwater heater 112, a feedwater pipeline 113, a reactor recirculation system 114 that causes changes in the quantity of core coolant that recirculates therethrough, a control rod drive system 115 that controls the output, a standby coolant system 116 that operates when the reactor has been isolated by valves, a residual heat removal system that removes residual heat when the reactor is halted, and an emergency core coolant system (ECCS) that operates during emergencies. Existing BWRs use a forced recirculation method by which coolant is sent through the core by the reactor recirculation system 114. This reactor recirculation system 114 is configured of a recirculation pump 117 and a jet pump 118. In an emergency, the recirculation pump 117 has a certain amount of inertia and takes about five seconds to stop, so the cooling efficiency of the coolant has to rely on a relatively weak natural circulatory system. The ECCS is configured of a high-pressure core spray system 119 and a low-pressure core spray system 120 that also acts as the residual heat removal system. These operate together with a containment spray 121. The emergency core coolant system uses a condensate storage tank 122 or the pressure control pool 104 as a water source and supplies water into the core 101 by the rotation of a centrifugal pump driven by power supplied from emergency diesel generators 123, or sprays the water into the reactor container 106. During an emergency, a boric acid solution is dumped in by an SLC (Standby Liquid Control System) pump 125 from a SLCS tank 124 into a lower plenum of the reactor pressure vessel. In an advanced boiling-water reactor (ABWR), which is an improvement on the BWR, the external recirculation piping for the reactor recirculation system of the above described BWR is removed and recirculation through the core is enabled by the provision of a jet pump that is smaller than an internal pump, instead. The employment of an internal pump has various effects, such as a reduction in the pressure losses with respect to the flow of reactor coolant, in comparison with a BWR. A cross-sectional view illustrating the concept of the systems of this ABWR is shown in FIG. 28. A core 52 that is provided with a large number of fuel rod assemblies is disposed slightly below the center of a reactor presure vessel 51. A large number of control rod guidance tubes 53 are provided below this core 52, and an upper aperture of a shroud 54 that shapes the core 52 is closed by a shroud head 55. Stand pipes 57 of steam separators 56 are erected above the shroud head 55, and flat, rectangular steam dryers 58 are disposed above the steam separators 56. A control rod drive mechanism 59 is provided in a lower portion of the presure vessel 51, to drive the cross-shaped control rods within the core 52, using the inner surfaces of the control rod guide tubes 53. A plurality of internal pumps 60 are disposed in a base portion between the inner side of the reactor presure vessel 51 and the outer side of the shroud 54. The core 52 is supported by a core support plate 61 that supports a lower portion of the large number of fuel rod assemblies, an upper portion thereof is supported by an upper lattice plate 62, and the entire core is surrounded by the shroud 54. A main steam line 108 that sends steam that has been dried by the steam dryers 58 to a turbine is connected to the reactor presure vessel 51. Coolant flowing into the reactor pressure 51 from a feedwater line 113 is recirculated by the internal pumps 60. The reactor presure vessel 51 is mounted on and fixed to a pedestal, with a support skirt 63 therebetween. An upper aperture of the reactor presure vessel 51 is hermetically sealed by an upper lid 64. A cross-sectional view of one of the steam separators 56 disposed within the reactor presure vessel 51 is shown in FIG. 29. This steam separator 56 comprises swirl vanes 41 provided above each of the stand pipes 57 to impart a swirling motion to a two-phase flow of steam-water mixtures, and steam separator stages 42a, 42b, and 42c provided above the swirl vanes in three consecutive stages in the axial direction as steam separator means for separating the steam from the two-phase liquid-vapor flow. Each of the steam separator stages 42a, 42b, or 42c has a double structure of a revolving tube 43a, 43b, or 43c with an outer tube 44a, 44b, or 44c positioned on the outer sides thereof. There is a hook-shaped pickoff ring 45a, 45b, or 45c formed on an upper portion of each of the outer tubes 44a, 44b, and 44c, respectively. The description now turns to the operation of the steam separator 56. Coolant that has been boiled off by the heat of the fission reaction forms a two-phase liquid-vapor flow in which ordinary water and steam are mixed. It is distributed between the steam separators 56, which normally number between 200 and 300, and rises to the stand pipes 57. As shown in FIG. 29, the coolant within the stand pipes 57 forms a fluidized state called an annular flow. In other words, a liquid layer 48 covers the inner wall surface of each stand pipe 57 and a mixture of water droplets 49 and steam 50 flows within this liquid layer 48. A centrifugal force is forcibly imparted to the two-phase flow rising through the stand pipe 57 by the swirl vanes 41 disposed directly above the stand pipe 57, to turn it into a rotating flow. At this point, the liquid-vapor density ratio of the coolant under normal operating conditions of the boiling-water reactor is 1:21, and thus a useful difference is generated in the centrifugal forces that are imparted by the rotational action to each of the vapor phase and the liquid phase. This ensures that the low-density steam is positioned towards the center of the lowermost steam separator stage 42a, the high-density liquid forms the liquid layer 48 along the inner wall surface of the revolving tube 43a of this steam separator stage 42a, and both rise while rotating. This liquid layer 48 is carried upward along the inner wall of the revolving tube 43a against its own weight by the shear forces of the high-speed rotating flow near the center and is captured by the pickoff ring 45a which is a slit having a width that is designed to be substantially equal to the thickness of this liquid layer 48. Then, a thin annular portion between the concentric tubes 43a and 44a falls under its own weight. A breakdown ring 47 is provided partway along this flow path to prevent the intermixing of a large quantity of vapor bubbles, and the flow is sent on at a slower speed to an upper downcomer where it mixes with the surrounding liquid. The larger part of the liquid phase that has not been captured by the lowermost steam separator stage 42a is captured by the pickoff rings 45b and 45c of the subsequent steam separator stages 42b and 42c. Note that the apparatus is designed in such a manner that approximately 90% of the moisture extracted by the steam separator 56 from the steam passing through the steam separator 56 is removed by the lowermost steam separator stage 42a, and the mass ratio of water amidst the two-phase flow at the outlet of the steam separator 56 is suppressed to no more than 10%. More of the moisture in the steam that has passed through the steam separator 56 is removed by the steam dryer 58 disposed above each steam separator 56. A steam injector has recently attracted attention as a static jet pump to be used instead of the prior-art rotary pump. This steam injector has a compact structure, requires no power source for operation, and can also be made to have a discharge pressure that is higher than the steam pressure at the inlet thereof. An objective of the present invention is to make full use of the above characteristics in the application of a steam injector to a steam separator, to provide a steam separator which achieves substantially the same liquid-vapor separation effect as that of the above prior-art steam separator, and, at the same time, provide a higher discharge pressure. The recirculation method currently used in BWRs and ABWRs necessitates components such as a large-scale pump, which is a rotating mechanism, and a large-capacity inverter power source for controlling that pump. From various viewpoints such as structural cost, material resources, and regularly scheduled maintenance, this method increases the cost of the plant and causes breakdowns in the rotating mechanisms. In contrast thereto, a movement has recently been seen to implement a simplified BWR with a modified natural recirculation method for the core that does not require jet pumps and internal pumps. Because the electrical output thereof is small in comparison to the size of the plant, the construction costs and unit-power costs tend to increase. In a similar manner, it has become possible to design smaller, simpler equipment in pressurized water reactors (PWRs) and fast breeder reactors (FBRs) as well, by reinforcing the natural recirculation forces within steam generators. This is not limited to reactors; there is also a large demand for smaller, simpler installations having processes that separate out a liquid phase comprised within steam, such as boilers. SUMMARY OF THE INVENTION The present invention was devised in the light of the above described problems of the prior art, and has as an objective the implementation of a steam separator wherein a discharge pressure is higher than the steam pressure at the inlet thereof, by using an injector type of steam separator instead of the steam separator of the prior art. Another objective of the present invention is to implement a great simplification of the equipment installed around the core of a reactor, by applying an injector type of steam separator, i.e., a separator/injector, to a nuclear power generation plant and by achieving both an increase in the performance of the steam separator for a two-phase flow at the core outlet and the establishment of a forced circulation in the core. A further objective of the present invention is to reduce the flow-rate of a recirculation pump of a boiler apparatus, and thus implement a simpler overall structure, by applying an injector type of steam separator to the boiler apparatus to convert the heat exchanger thereof from a natural circulatory system to a forced circulatory system. In order to achieve the above objectives, a first aspect of the present invention relates to a steam separator equipped with a separator/injector, wherein the separator/injector comprises: a two-phase flow accelerator nozzle having an inlet portion opening towards a source of a two-phase liquid-vapor flow and an outlet portion positioned higher than the inlet portion, for causing an acceleration of the two-phase liquid-vapor flow that flows into an interior portion thereof from the inlet portion and discharging the same from the outlet portion; a liquid-phase capture means connected to the outlet portion of the two-phase flow accelerator nozzle and having a guide wall formed as an inverted U-shape curve, in such a manner that the two-phase flow of steam-motor mixture that is discharged from the outlet portion of the two-phase flow accelerator nozzle is guided along a wall surface of the guide wall but is also capable of separating therefrom, wherein a difference in centrifugal forces that are imparted to a vapor-phase component and a liquid-phase component of the two-phase flow while the two-phase flow is being guided along the guide wall causes the liquid-phase component to be guided along the wall surface of the guide wall and be captured, whereas the vapor-phase component is allowed to separate from the guide wall; and a diffuser into which the liquid-phase component captured by the liquid-phase capture means is allowed to flow, which increases the pressure of the liquid-phase component as the liquid-phase component flows therethrough, and which discharges the liquid-phase component from an outlet side thereof. With this configuration, the two-phase flow of steam and water are accelerated by the two-phase flow accelerator nozzle, a strong centrifugal force is applied thereto, and thus the vapor and liquid are separated. In other words, the high-density water flows at substantially constant speed along the flowpath formed by the liquid-phase capture means, but the low-density steam separates from this flowpath, is released into a vapor space, and moves above the separator/injector. The water (liquid phase) that has flowed along the flowpath then flows into the diffuser, is decelerated (in accordance with the Bernoulli principle) as the cross-sectional area of the flowpath increases within the diffuser, and is then discharged out of the separator/injector through a diffuser outlet. During this time, the pressure of the discharge water is increased by the diffuser, so that the outlet pressure of the separator/injector can be made higher than the inlet pressure thereof. The nuclear power generation plant of a second aspect of the present invention is further provided with a foundation portion such that the separator/injector is erected upon the foundation portion, wherein the foundation portion comprises an upper plate and a lower plate positioned below the upper plate to form a space therebetween; the configuration being such that the inlet portion of the two-phase flow accelerator nozzle communicates with a space positioned below the lower plate, and the outlet side of the diffuser communicates with the space formed between the upper and lower plates. This provides complete separation between the two-phase flow that enters the separator/injector and the discharge water from the separator/injector. In a third aspect of this invention, the wall surface of the guide wall of the liquid-phase capture means is preferably formed in an arch-shaped curve orientated upward and at least one portion thereof is in the shape of a circular or elliptical arc. In addition, a side edge portion of the guide wall of the liquid-phase capture means is preferably bent in a direction to enclose the two-phase flow, so that the liquid phase does not flow out of the separator/injector from the underside of the flowpath. In a fourth aspect of this invention, riblet grooves are formed in the flow direction of the two-phase flow or the liquid-phase flow along at least one portion of an inner wall surface of the two-phase flow accelerator nozzle, the wall surface of the guide wall of the liquid-phase capture means, and an inner wall surface of the diffuser. This makes it possible to reduce frictional losses in the fluid in the vicinity of the inner wall surfaces. In a fifth aspect of this invention, the steam separator preferably further comprises an outer tube having an axis in the vertical direction comprising the separator/injector within the outer tube and an inner tube having an axis in the vertical direction disposed within the outer tube; wherein: the wall surface of the guide wall of the liquid-phase capture means is formed by part of an inner wall surface of the inner tube; the diffuser is formed to be in contact with an inner wall of the inner tube; and a space is formed between an inner wall of the outer tube and an outer wall of the inner tube in such a manner that the liquid-phase component discharged from the outlet side of the diffuser is capable of flowing therethrough. In this case, the lower portion of the two-phase flow accelerator nozzle is formed along the axial line of the inner tube, the outlet portion of the two-phase flow accelerator nozzle is formed in the vicinity of an inner wall of the inner tube, and the diffuser is formed in a helical shape with respect to the axial line of the inner tube. This configuration ensures that a centrifugal force is imparted to the two-phase flow discharged from the two-phase flow accelerator nozzle as it flows along the helical flowpath in the vicinity of the inner wall surface of the inner tube, the high-density water flows into the diffuser while being pressed against the inner wall surface, and the low-density steam is separated out towards the center in the axial direction and rises. Thus the two-phase flow is separated into liquid and vapor. The present invention also provides a nuclear power generating system in which this separator/injector is mounted. In other words, the fifth aspect of the present invention relates to a nuclear power generation plant using a boiling-water reactor, wherein the nuclear power generation plant comprises: a reactor pressure vessel; a plurality of fuel rod assemblies disposed within the reactor pressure vessel and through which a coolant flows; a shroud surrounding the plurality of fuel rod assemblies, within which is comprised a two-phase liquid-vapor flow that is created as the coolant flows within the plurality of fuel rod assemblies, and which is sealed by a shroud head at an upper end thereof; and a separator/injector erected above the shroud head; wherein the separator/injector comprises: a two-phase flow accelerator nozzle having an inlet portion opening towards the interior of the shroud and an outlet portion positioned higher than the inlet portion, for causing an acceleration of the two-phase liquid-vapor flow generated in the shroud that flows into an interior portion thereof from the inlet portion and discharging the same from the outlet portion; a liquid-phase capture means connected to the outlet portion of the two-phase flow accelerator nozzle and having a guide wall formed as an inverted U-shape curve, in such a manner that the two-phase liquid-vapor flow that is discharged from the outlet portion of the two-phase flow accelerator nozzle is guided along a wall surface of the guide wall but is also capable of separating therefrom, wherein a difference in centrifugal forces that are imparted to a vapor-phase component and a liquid-phase component of the two-phase flow while the two-phase flow is being guided along the guide wall causes the liquid-phase component to be guided along the wall surface of the guide wall and be captured, whereas the vapor-phase component is allowed to separate from the guide wall; and a diffuser into which the liquid-phase component captured by the liquid-phase capture means is allowed to flow, which increases the pressure of the liquid-phase component as the liquid-phase component flows therethrough, and which discharges the liquid-phase component from an outlet side thereof. During this time, recirculation is performed within the reactor by returning increased-pressure coolant that flows out of the diffuser back into the shroud. The wall surface of the guide wall of the liquid-phase capture means is preferably formed to be a smooth curve. The above configuration ensures that the two-phase flow is accelerated by the two-phase flow accelerator nozzle, the liquid phase thereof is subjected to a strong centrifugal force and is separated thereby, the water (liquid phase) then flows into the diffuser, is decelerated (in accordance with the Bernoulli principle) as the cross-sectional area of the flowpath increases within the diffuser, and the pressure thereof is increased, creating a recirculation drive force. In a sixth aspect of the present invention, the shroud head is formed as a double structure having an upper shroud head and a lower shroud head which is positioned below the upper shroud head to form a space between the upper and lower shroud heads; the inlet portion of the two-phase flow accelerator nozzle communicates with a space within the shroud that is positioned below the lower shroud head; and the outlet portion of the diffuser communicates with the space formed between the upper and lower shroud heads. The nuclear power generation plant of a seventh aspect of the present invention is further provided with a jet pump drive nozzle disposed in an upper portion of a downcomer portion surrounding the shroud; and a jet pump provided below the jet pump drive nozzle; wherein the configuration could be set in such a manner that, after the coolant discharged from the diffuser has passed through the space formed between the upper and lower shroud heads, the coolant is guided into the jet pump through the jet pump drive nozzle. The recirculatory force of the discharge water from the separator/injector makes it possible to reduce the number of jet pumps used in the prior art. In an eighth aspect of the present invention, a pipeline that branches off from a feedwater pipeline that links the reactor pressure vessel to a feedwater pump is connected to the jet pump drive nozzle. This makes it possible to supply water from the feedwater pump through the jet pump drive nozzle to the jet pump. If necessary, the nuclear power generation plant also comprises a feedwater pump for supplying water to the reactor pressure vessel; a feedwater pipeline linking the reactor pressure vessel to the feedwater pump; and a branch pipeline branching off from the feedwater pipeline and communicating with the jet pump drive nozzle, wherein the feedwater pump supplies water to the jet pump through the branch pipeline and the jet pump drive nozzle. In a nuclear power generation plant of a ninth aspect of this invention, pressure is increased to control the circulation flow-rate in sequence from the interior of the downcomer portion, the interior of the shroud, the inlet portion of the separator/injector, to the outlet portion of the separator/injector by controlling the flow-rate and discharge pressure of water supplied from the feedwater pump to the jet pump, thereby controlling the thermal output generated within the reactor pressure vessel. This makes it possible to increase the pressure in sequence from the downcomer portion, a lower plenum, the core, an upper plenum, the inlet portion of the separator/injector, to the outlet portion of the separator/injector by, for example, increasing the flow-rate and discharge pressure of water supplied from the feedwater pump to the jet pump, to increase the core circulation flow-rate, thereby controlling the thermal output of the core. In a tenth aspect of the present invention, the jet pump could be driven at the start-up of the plant by mixing a flow of feedwater supplied from at least one of a pump in a residual heat removal system and a pump in a reactor water clean-up system with a flow of feedwater supplied from the feedwater pump to the jet pump. The nuclear power generation plant of an eleventh aspect of this invention further comprises a recirculation flow-rate control valve disposed in the outlet portion or the inlet portion of the jet pump; and a flow-rate control means that uses at least one of an electrical generator output signal, a main steam flow-rate signal, a neutron flux output signal, and a jet pump pressure difference signal as an input signal, calculates a suitable recirculation flow-rate and corresponding degree of opening of the recirculation flow-rate control valve therefrom, and outputs a valve-opening signal. This adjusts the recirculation flow-rate of the core in accordance with the setting and adjustment of the degree of opening of the valve as appropriate in accordance with the electrical output required of the nuclear power generation plant. The nuclear power generation plant of a twelfth aspect of this invention further comprises a feedwater pump for supplying water to the reactor pressure vessel; a feedwater line connected to the reactor pressure vessel to the feedwater pump; a flow-rate adjustment valve provided in the feedwater pipeline; and a branch pipeline branching off from the feedwater pipeline and communicating with the jet pump drive nozzle, wherein the flow-rate control means controls a water level of the reactor by calculating a suitable degree of opening of the flow-rate adjustment valve and outputting a valve-opening signal to the flow-rate adjustment valve. This monitoring and control of the water level of the nuclear reactor makes it possible to maintain the core water level to be substantially constant. Alternatively, the flow-rate control means could control the flow-rate of water supplied to the reactor by calculating a suitable feedwater flow-rate and outputting a rotational frequency signal to the feedwater pump. The nuclear power generation plant of a thirteenth aspect of the present invention further comprises a jet pump drive nozzle disposed in an upper portion of a downcomer portion surrounding the shroud; a bell mouth opening towards the downcomer portion in close proximity to the jet pump drive nozzle; a straight pipe having one end connected to a leading edge of the bell mouth and another end connected to an inlet side of the jet pump; and a jet pump provided below the jet pump drive nozzle; wherein after the coolant, which has accumulated in the upper shroud head without passing through the diffuser, has passed through the bell mouth and the straight pipe via the jet pump drive nozzle, the coolant is guided into the jet pump. This configuration imparts a recirculatory drive force of the coolant flowing through the downcomer portion. A fourteenth aspect of the present invention relates to an ABWR characterized in that the shroud head is formed as a double structure having an upper shroud head and a lower shroud head which is positioned below the upper shroud head to form a space between the upper and lower shroud heads; and the ABWR is further provided with: a downcomer pipe having an upper end portion opening upwards in an upper surface of the upper shroud head, for guiding downwards a liquid phase of coolant that has accumulated on the upper surface of the upper shroud head, without being captured by the liquid-phase capture means; an impeller provided in a lower portion of the downcomer pipe for rotating in such a manner that coolant flowing downward within the downcomer pipe is discharged towards a lower portion of the shroud; and an internal pump for driving the impeller; wherein the outlet side of the diffuser communicates with a space formed between the upper and lower shroud heads, and coolant that is discharged from the outlet side of the diffuser flows down through the space formed between the upper and lower shroud heads and into a downcomer portion on an outer side of the downcomer pipe, then is discharged to a lower portion of the shroud. The nuclear power generation plant of a fifteenth aspect of this present invention further comprises a recirculation flow-rate control means that uses at least one of an electrical generator output signal, a main steam flow-rate signal, a neutron flux output signal, and a core support plate pressure difference signal as an input signal, calculates a suitable recirculation flow-rate and a suitable rotational frequency of the impeller therefrom, and outputs a suitable rotational-frequency signal with respect to the internal pump. This configuration adjusts the recirculation flow-rate in the core by setting and regulating the rotational frequency of the pump as appropriate in accordance with the electrical output required of the nuclear power generation plant. The recirculation flow-rate control means controls the rotational frequency of the internal pump by performing calculations in accordance with overall proportional integral differential (PID) control relating to a difference from a predetermined water level based on an input reactor core water-level signal. This makes it possible to control the rotational frequency of the internal pump to maintain the core water level to be substantially constant. The nuclear power generation plant of a sixteenth aspect of the present invention is further characterized in that each of a shroud casing forming a side portion of the shroud and a core support plate forming a lower portion of the shroud is formed as a double structure; and the plant further comprises a first coolant circulation pathway formed so as to mutually communicate the space between the shroud heads, a space formed between the double shroud casings, and a space formed between the double core support plates, through which flows coolant discharged from the outlet portion of the diffuser; a water rod provided in the interior of the fuel rod assemblies, within which coolant flows; a first coolant guide pipe formed so as to communicate the first coolant circulation pathway with an outlet portion at a lower end of the water rod, for guiding increased-pressure coolant, which has been discharged from the outlet side of the diffuser and which is flowing through the first coolant circulation pathway, to the outlet portion at the lower end of the water rod; and a hole formed in a side surface of the water rod, for ejecting coolant that is flowing within the water rod to the exterior of the water rod. This configuration makes it possible to cause an increase in the liquid-phase flow-rate of the two-phase flow in the core, by allowing the high-pressure discharge water from the separator/injector to flow into the fuel rod assemblies. In the nuclear power generation plant of a seventeenth aspect of the present invention, the coefficient of thermal expansion of the material configuring the first coolant guide pipe and the coefficient of thermal expansion of the material configuring the water rod are set to be different in the vicinity of a connective portion between the first coolant guide pipe and the water rod. In the nuclear power generation plant of an eighteenth aspect of the present invention, labyrinth grooves are provided in the first coolant guide pipe and the water rod in the vicinity of a connective portion between the first coolant guide pipe and the water rod. This prevents leakage of the coolant by increasing the resistance of the flowpath of leaking coolant. The nuclear power generation plant of a nineteenth aspect of the present invention further comprises a second coolant guide pipe provided within a control rod tube positioned below the fuel rod assemblies, for guiding coolant that is outside the shroud into a lower tie plate of the fuel rod assemblies; and an orifice provided in a second coolant circulation path that is formed by the second coolant guide pipe, for locally constricting the flowpath thereof. This suppresses any increase in the pressure losses in the coolant flowpath. In a nuclear power generation plant of a twentieth aspect of the present invention, holes are provided in a side surface of an inner shroud casing of the double shroud casings, and a side surface of a channel box of the fuel rod assemblies. This makes it possible to even out the coolant density within the fuel rod assemblies. A twenty-first aspect of the present invention relates to a nuclear power generation plant using a boiling-water reactor, wherein the nuclear power generation plant comprises: a steam generator comprising a lower casing surrounding a heat exchanger formed of heat-exchange pipes having inlets and outlets for a primary coolant, and an upper casing provided connected to the lower casing and having a steam outlet for supplying steam to a turbine; a reactor container communicating with the steam generator and surrounding the primary coolant and a fuel rod assembly; and a separator/injector provided above the lower casing; wherein the separator/injector comprises a two-phase flow accelerator nozzle having an inlet portion opening towards the interior of the lower casing and an outlet portion positioned higher than the inlet portion, for causing an acceleration of a two-phase flow of the primary coolant that flows into an interior portion thereof from the inlet portion and discharging the same from the outlet portion; a liquid-phase capture means connected to the outlet portion of the two-phase flow accelerator nozzle and having a guide wall formed as a inverted U-shape curve, in such a manner that the two-phase flow that is discharged from the outlet portion of the two-phase flow accelerator nozzle is guided along a wall surface of the guide wall but is also capable of separating therefrom, wherein a difference in centrifugal forces that are imparted to a vapor-phase component and a liquid-phase component of the two-phase flow while the two-phase flow is being guided along the guide wall causes the liquid-phase component to be guided along the wall surface of the guide wall and be captured, whereas the vapor-phase component is allowed to separate from the guide wall; and a diffuser into which the liquid-phase component captured by the liquid-phase capture means is allowed to flow, which increases the pressure of the liquid-phase component as the liquid-phase component flows therethrough, and which discharges the liquid-phase component from an outlet side thereof. In this case, it is preferable that the nuclear power generation plant further comprises an inner casing surrounding the heat exchanger within the lower casing; wherein discharge water exhausted from the diffuser is guided into the heat exchanger through a space formed between the lower casing and the inner casing. This makes it possible to improve the heat-transfer characteristics by creating a forced circulation of fluid on a secondary side in the steam generator of this PWR. A twenty-second aspect of the present invention relates to a boiler apparatus comprising: a pressure vessel surrounding heat-transfer tubes that configure a heat exchanger and a combustor for heating the heat-transfer tubes; a recirculation pump for recirculating a fluid that flows through the pressure vessel; and a separator/injector provided above the heat-transfer tubes; wherein the separator/injector comprises a two-phase flow accelerator nozzle having an inlet portion opening towards the interior of the pressure vessel and an outlet portion positioned higher than the inlet portion, for causing an acceleration of a two-phase liquid-vapor flow that flows into an interior portion thereof from the inlet portion and flows through the heat exchanger, and discharging the same from the outlet portion; a liquid-phase capture means connected to the outlet portion of the two-phase flow accelerator nozzle and having a guide wall formed as a inverted U-shape curve, in such a manner that the two-phase flow that is discharged from the outlet portion of the two-phase flow accelerator nozzle is guided along a wall surface of the guide wall but is also capable of separating therefrom, whereby a difference in centrifugal forces that are imparted to a vapor-phase component and a liquid-phase component of the two-phase flow while the two-phase flow is being guided along the guide wall causes the liquid-phase component to be guided along the wall surface of the guide wall and be captured, whereas the vapor-phase component is allowed to separate from the guide wall; and a diffuser into which the liquid-phase component captured by the liquid-phase capture means is allowed to flow, which increases the pressure of the liquid-phase component as the liquid-phase component flows therethrough, and which discharges the liquid-phase component from an outlet side thereof. This makes it possible to cause a reduction in the flow-rate of the recirculation pump by creating a forced circulation of fluid within the boiler. The configuration of the present invention enables the following effects. In other words, it makes it possible to implement a steam separator that is equipped with a separator/injector that enables an outlet pressure that is higher than the inlet pressure thereof, in addition to the steam separation capability of the prior art. Installing this separator/injector in a nuclear power generation plant or boiler apparatus makes it possible to separate steam and water from a two-phase flow and achieve a forced circulation in the core, without requiring the complicated configuration of the prior art. This means that the number of items of dynamic recirculation equipment required in the art, such as recirculation pumps or internal pumps, can be reduced, which leads to a huge reduction in the equipment and structural resources of the entire apparatus, rationalization and simplification of the apparatus, and also a reduction in the time and costs involved in construction and maintenance.
	claims	1. A mass transfer apparatus, comprising:a housing and, provided therein, a flow reaction chamber filled with an oxidation agent, provided with an adjustable heating system, and systems for inlet and outlet of oxidizable material, wherein the housing is equipped with a repository for reserves of the oxidation agent. 2. The mass transfer apparatus according to claim 1, wherein an electric heater is used as an adjustable heating system. 3. The mass transfer apparatus according to claim 2, wherein a high resistance wire made of nichrome or fechral is used as a heating element. 4. The mass transfer apparatus according to claim 1, wherein the repository for reserves of the oxidation agent includes a bottom and a side wall, formed by a lower part of the housing. 5. The mass transfer apparatus according to claim 4, wherein openings are made in an upper part of the side wall of the repository for reserves of the oxidation agent, adjacent to the flow reaction chamber. 6. The mass transfer apparatus according to claim 4, wherein openings are made in a lower part of the side wall of the repository for reserves of the oxidation agent. 7. The mass transfer apparatus according to claim 1, wherein the repository for reserves of the oxidation agent is located below the flow reaction chamber. 8. The mass transfer apparatus according to claim 2, wherein the repository for reserves of the oxidation agent is located below a lower end of the electric heater. 9. The mass transfer apparatus according to claim 2, wherein, in an initial state, a volume of the repository for reserves of the oxidation agent is filled with the oxidation agent. 10. The mass transfer apparatus according to claim 1, wherein the flow reaction chamber is formed by a middle part of the housing, defined from below by an upper part of the repository for reserves of the oxidation agent, and from above, by a restrictive grille. 11. The mass transfer apparatus according to claim 10, wherein in the restrictive grille, there are openings provided. 12. The mass transfer apparatus according to claim 2, wherein the system for inlet of the oxidizable material is formed by an upper part of a side wall of the repository for reserves of the oxidation agent. 13. The mass transfer apparatus according to claim 1, wherein the system for outlet of the oxidizable material is formed by a restrictive grille of the flow reaction chamber and openings in a wall of the housing. 14. The mass transfer apparatus according to claim 13, wherein the system for outlet of the oxidizable material is located below the flow reaction chamber. 15. The mass transfer apparatus according to claim 1, wherein the oxidation agent is comprised of a solid-phase agent. 16. The mass transfer apparatus according to claim 1, wherein the produced oxidation agent is comprised of separate particles. 17. The mass transfer apparatus according to claim 1, wherein a granulated lead oxide is used as a solid-phase oxidation agent. 18. The mass transfer apparatus according to claim 5, wherein the openings are made in the form of a series of slits having a width lesser than a size of particles of a solid-phase oxidation agent. 19. The mass transfer apparatus according to claim 1, characterized in that it is located horizontally in the oxidizable material reservoir.
042971705	claims	1. A device for transversely holding the fuel rods of a bundle of fuel rods for a nuclear reactor assembly, which rods are arranged parallel to one another, in a uniform arrangement, having a plurality of support means for exerting forces on the rods in transverse directions to restrict vibrations thereof, while permitting axial lengthening of the rods, each of said support means comprising three sub-units comprising grids arranged adjacent one another in the longitudinal direction of the assembly, said sub-units comprising: two end grids each formed by small metal plates which are assembled in a uniform network delimiting cells inside which the rods pass, and whose side walls, parallel to the longitudinal axis of the assembly, are formed by said small metal plates arranged transversely, relative to the assembly, so that their width constitutes the thickness of said grid in the longitudinal direction, said end grids being identical and each grid having the same network as the other, each wall of each cell of one grid being located in the same longitudinal plane as the corresponding wall of the other grid, said grids being made integral with one another with a longitudinal spacing along the fuel rods, and an intermediate grid comprising small, elastic metal plates, the width of which is less than the width of said plates of said end grids, and which are assembled in a uniform network delimiting cells inside which the rods pass and arranged between said end grids in said longitudinal space provided therebetween, in a staggered manner in a transverse direction relative to said end grids, so that two adjacent elastic walls of the cells of said intermediate grid exert on each fuel rods transverse forces which press the rods against two walls of said cells of each end grid, said intermediate grid resting on at least one of said end grids. 2. The device according to claim 1, wherein at least said one of said end grids is provided with slots, in that part thereof which is in contact with said intermediate grid, for holding said intermediate grid in place relative to said end grids, with the possibility of transverse adjustment. 3. The device according to claim 1, wherein said intermediate grid has two axes of symmetry, said transverse staggering of said intermediate grid in each of its parts delimited by said axes of symmetry being symmetrical with said staggering in the other parts of said grid, relative to the axes of said grid. 4. The device according to claim 1, wherein said intermediate grid comprises a set of individual grids which are juxtaposed so as to cover the entire transverse surface of said assembly. 5. The device according to claim 1, wherein the walls of said cells of said end grids in contact with the fuel rods are provided with projections for engaging the rods. 6. The device according to claim 1, wherein the edges of said small metal plates of said intermediate grid are provided with projections or chamfeers for facilitating the guiding of the rods when they are introduced. 7. The device according to claim 1, largely consisting of small plates made of a metal with low neutron absorption. 8. The device according to claim 1, wherein the metal of said small metal plates of said intermediate grid has a higher coefficient of expansion than the metal of said small metal plates of said end grids.
	claims	1. A replacement thermal sleeve system comprising:a control rod drive mechanism (CRDM) comprising a head penetration adaptor, wherein the head penetration adaptor defines an entrance region, an exit region, and a shaft extending between the entrance region and the exit region, wherein the shaft comprises an inner diameter;a replacement thermal sleeve configured to replace an existing thermal sleeve situated in the head penetration adapter, the replacement thermal sleeve comprising:a base, wherein a central longitudinal axis extends through the base; andflanges extending from the base, wherein the flanges are disposed about the central longitudinal axis, wherein gaps are defined between the flanges, wherein each flange comprises a lug extending therefrom and a curved surface, and wherein the flanges are configurable in:a first configuration, wherein the lugs define a first diameter; anda second configuration, wherein the flanges are flexed toward the central longitudinal axis and the lugs define a second diameter, wherein the second diameter is less than first diameter and the inner diameter of the shaft; anda retention member comprising portions, wherein each portion is configured to engage the curved surface of one of the flanges to maintain the flanges in the second configuration prior to the replacement thermal sleeve entering the head penetration adaptor, and wherein the retention member is configured to release the flanges and allow the flanges to transition toward the first configuration, wherein the retention member is configured to maintain the flanges in the second configuration while the replacement thermal sleeve is moved inside the head penetration adaptor to an operational position, and wherein the retention member is configured to allow release of the flanges with the replacement thermal sleeve at the operational position. 2. The replacement thermal sleeve system of claim 1, wherein the flanges are configured to traverse the shaft from the entrance region toward the exit region in the second configuration. 3. The replacement thermal sleeve system of claim 2, wherein the flanges are configured to transition to the first configuration, based on the lugs exiting the shaft into the exit region. 4. The replacement thermal sleeve system of claim 3, wherein the exit region comprises a shoulder, and wherein the lugs are configured to engage the shoulder. 5. The replacement thermal sleeve system of claim 1, wherein each lug comprises a curved surface and a tapered surface, and wherein each tapered surface tapers toward the central longitudinal axis. 6. A system, comprising:a control rod drive mechanism (CRDM) comprising a head penetration adaptor, wherein the head penetration adaptor, defines:an entrance zone;an exit zone; anda shaft extending between the entrance zone and the exit zone, wherein the shaft comprises an inner diameter; anda replacement thermal sleeve configured to replace a thermal sleeve previously positioned in the head penetration adapter, comprising:a base, wherein a central longitudinal axis extends through the base; andflanges extending from the base, wherein flanges are disposed about the central longitudinal axis, wherein gaps are defined between flanges, wherein each flange comprises a head extending therefrom and a curved surface, and wherein the flanges are configurable in:a relaxed configuration, wherein the heads of the flanges define a first diameter; anda flexed configuration, wherein the flanges are flexed toward the central longitudinal axis and the heads define a second diameter, wherein the second diameter is less than first diameter and the inner diameter of the shaft; anda retention member comprising portions, wherein each portion is configured to engage the curved surface of one the heads to maintain the flanges in the flexed configuration prior to the replacement thermal sleeve entering the head penetration adaptor, wherein the retention member is configured to release the heads and allow the flanges to transition toward the relaxed configuration, wherein the retention member is configured to maintain the flanges in the flexed configuration while the replacement thermal sleeve is moved inside the head penetration adaptor to an operational position, and wherein the retention member is configured to allow release of the flanges with the replacement thermal sleeve at the operational position. 7. The system of claim 6, wherein flanges are configured to traverse the shaft from the entrance zone toward the exit zone in the flexed configuration. 8. The system of claim 7, wherein flanges are configured to transition to the relaxed configuration, based on the heads exiting the shaft into the exit zone. 9. The system of claim 8, wherein the exit zone comprises a shoulder, and wherein the heads are configured to engage the shoulder. 10. The system of claim 6, wherein each head comprises a curved surface and a tapered surface, and wherein each tapered surface tapers toward the central longitudinal axis. 11. A comprising:a control rod drive mechanism (CRDM) comprising a head penetration adaptor, wherein the head penetration adaptor defines an entrance region, an exit region comprising a shoulder, and a shaft extending between the entrance region and the exit region, wherein the shaft comprises an inner diameter;a replacement thermal sleeve configured to replace an existing thermal sleeve situated in the head penetration adapter, comprising:a base, wherein a central longitudinal axis extends through the base; andflanges extending from the base, wherein flanges are disposed about the central longitudinal axis, wherein gaps are defined between flanges, wherein each flange comprises a lug extending therefrom and a curved surface, wherein each lug comprises a curved surface and a tapered surface that tapers toward the central longitudinal axis, and wherein the flanges are configurable in:a relaxed configuration, wherein the curved surfaces of the lugs define a first diameter; anda flexed configuration, wherein the flanges are flexed toward the central longitudinal axis and the curved surfaces of the lugs define a second diameter, wherein the second diameter is less than first diameter and the inner diameter of the shaft;wherein flanges are configured to traverse the shaft from the entrance region toward the exit region in the flexed configuration;wherein flanges are configured to transition to the relaxed configuration, based on the lugs exiting the shaft into the exit region; anda retention member comprising engagement surfaces, wherein each engagement surface is configured to engage the curved surface of one the flanges and maintain flanges in the flexed configuration prior to the replacement thermal sleeve entering the head penetration adaptor, wherein the retention member is configured to release flanges and allow flanges to transition toward the relaxed configuration, wherein the retention member is configured to maintain the flanges in the flexed configuration while the replacement thermal sleeve is moved inside the head penetration adaptor to an operational position, and wherein the retention member is configured to allow release of the flanges with the replacement thermal sleeve at the operational position.
	description	This application claims priority of Provisional Application Ser. No. 60/350,071 filed Jan. 23, 2002, which is herein incorporated by reference. The United States Government has rights in this invention pursuant to Contract No. DE-AC03-76SF00098 between the United States Department of Energy and the University of California. The invention relates to plasma ion generators and neutron sources based on plasma ion generators, and more particularly to the production of ultra-short pulses from these ion generators and neutron sources. In many applications, such as time of flight measurements, ultra-short neutron pulses (pulse width<1 μs) with fast rise times or fall times are desired. These neutrons can be high energy, epithermal, thermal, or cold neutrons, and they are normally produced by a fission reactor or an accelerator-based neutron generator. When ultra-short pulses are needed, the neutron output flux can be chopped by means of a rotating mechanical chopper. There are some disadvantages when these mechanical chopper schemes are used to form ultra-short neutron pulses. First, a large percentage of neutrons will be discarded and activation of material may occur. Second, when pulsed accelerator systems are employed, the mechanical chopper and the ion beam acceleration have to be properly synchronized. Ultra-short pulses cannot be formed by manipulating the plasma discharge because the rise time due to plasma buildup is typically on the order of a few μs. Other neutron sources are based on ion generators. Conventional neutron tubes employ a Penning ion source and a single gap extractor. The target is a deuterium or tritium chemical embedded in a molybdenum or tungsten substrate. University of California, Lawrence Berkeley National Laboratory has produced a number of compact neutron sources with a relatively high flux, particularly sources which generate neutrons using the D—D reaction instead of the D–T reaction. These sources have a variety of different geometries, including tubular, cylindrical, and spherical, and are based on plasma ion sources, particularly multicusp plasma ion sources, with single or preferably multiple beamlet extraction. These neutron sources are illustrated by copending U.S. patent applications Ser. Nos. 10/100,956; 10/100,962; and 10/100,955. The invention is an ion source with an extraction system configured to produce ultra-short ion pulses, i.e. pulses with pulse width of about 1 μs or less and fast rise times or fall times or both, and a neutron generator based on the ion source which produces correspondingly ultra-short neutron pulses. A deuterium ion (or mixed deuterium and tritium ion or even a tritium ion) plasma is produced by RF excitation in a plasma ion generator using an RF antenna. The ion generator is preferably a multicusp plasma ion source. The single or multi-aperture extraction system of the ion source has two spaced electrodes—a plasma electrode and an extraction electrode. Although a single aperture extraction system can be used, a multi-aperture extraction system is preferred for higher ion extraction current and neutron flux. The plasma and extraction electrodes of a multiple beamlet system are typically spherical or cylindrical in shape. To form a neutron generator, a neutron generating target is positioned to receive the extracted ion beam from the ion generator. The extracted ions are accelerated to energies in excess of 100 keV before impinging on the target, which becomes loaded with neutral deuterium and/or tritium atoms. Very short pulses of 2.45 MeV D—D neutrons or 14.1 MeV D-T neutrons will be produced by striking the target with ultra-short ion beam bursts. To produce the ultra-short ion or neutron pulses, the apertures in the extraction system are suitably sized to prevent ion leakage, the electrodes are suitably spaced, and the extraction voltage is controlled. The ion beam current leaving the source is regulated by applying short voltage pulses of a suitable voltage on the extraction electrode. As shown in FIG. 1, compact high flux neutron generator 10 has a plasma ion source or generator 12, which typically is formed of a cylindrical shaped chamber. The principles of plasma ion sources are well known in the art. Preferably, ion source 12 is a magnetic cusp plasma ion source. Permanent magnets 14 are arranged in a spaced apart relationship, running longitudinally along plasma ion generator 12, to form a magnetic cusp plasma ion source. The principles of magnetic cusp plasma ion sources are well known in the art. Conventional multicusp ion sources are illustrated by U.S. Pat. Nos. 4,793,961; 4,447,732; 5,198,677; 6,094,012, which are herein incorporated by reference. Ion source 12 includes an RF antenna (induction coil) 16 for producing an ion plasma 18 from a gas which is introduced into ion source 12. RF antenna 16 is connected to RF power supply 20 through matching network 22. Ion source 12 may also include a filament 24 for startup. For neutron generation the plasma is preferably a deuterium ion plasma but may also be a deuterium and tritium plasma (or even a tritium plasma). Ion source 12 also includes a pair of spaced electrodes, plasma electrode 26 and extraction electrode 28, at one end thereof. Electrodes 26, 28 electrostatically control the passage of ions from plasma 18 out of ion source 12. Electrodes 26, 28 are substantially spherical or curved in shape (e.g. they are a portion of a sphere, e.g. a hemisphere) and contain many aligned holes 30 (shown in FIG. 2) over their surfaces so that ions radiate out of ion source 12. (In the simplest embodiment, there would only be a single extraction hole 30 in electrodes 26, 28.) Suitable extraction voltages are applied to electrodes 26, 28, e.g. plasma electrode 26 is at 0 kV and extraction electrode 28 is at −7 kV, so that positive ions are extracted from ion source 12. The extraction system of ion source 12 includes a third electrode, suppressor electrode 32 which contains a central aperture 34 therein. Suppressor electrode 32 is at a relatively high negative voltage, e.g. −160 kV, to accelerate the extracted ion beam. The three electrode extraction/accelerator system is used to expand a high current ion beam in a relatively short distance. The spherical shapes of the plasma and extraction electrodes 26, 28 are such that the ion beams (or beamlets) passing through all the holes 30 in electrodes 26, 28 are focused close to the suppressor electrode 32, pass through aperture 34, cross over, and expand or diverge on the other side of suppressor electrode 32. The diverging beam expands to a large area in a relatively short distance. Details of the extraction and acceleration system are shown in FIGS. 2, 3. The plasma density on the ion source side of the plasma electrode 26 must be uniform over the entire extraction area to ensure good ion beam extraction. Plasma uniformity is obtained by positioning a spherically curved magnetic filter 36 inside ion source 12 in front of plasma electrode 26. A spherically curved target 38 is positioned so that the expanding ion beam from ion source 12 passing through electrodes 26, 28, 32 is incident thereon. Target 38 forms a portion of a spherical surface of relatively large area at a relatively short distance from ion source 12. Target 38 is the neutron generating element, and may be water cooled. Target 38 is at a positive voltage relative to the suppressor electrode 32, e.g. at −150 kV. Ions from plasma source 12 pass through holes 30 in electrodes 26, 28, and through aperture 34 in electrode 32, and impinge on target 38, typically with energy of 120 keV to 150 keV, producing neutrons as the result of ion induced reactions. The target 38 is loaded with D (or D/T) atoms by the beam. Titanium is not required, but is preferred for target 38 since it improves the absorption of these atoms. Target 38 may be a titanium shell or a titanium coating on another chamber wall 40, e.g. a quartz tube. Ion source 12 is positioned at one end of a sealed tube 42, which also contains suppressor electrode 32, and neutron generating target 38, to form neutron generator 10. The entire neutron generator is very compact, e.g. about 30 cm in length. Because of the relatively large target area of target 38, and the high ion current from ion source 12, neutron flux can be generated from D—D reactions in this neutron generator as well as from D–T reactions as in a conventional neutron tube, eliminating the need for radioactive tritium. The neutrons produced, 2.45 MeV for D—D or 14.1 MeV for D–T, will go out from the end of tube 42. The neutron generator of the invention has a unique combination of high neutron production and compact size. The small size of the neutron generator is due mainly to the configuration of the extraction system, which allows one to extract a large ion beam current from a small ion source and to expand it onto a large area target. The large ion beam current is necessary for the high neutron output, because the neutron output is directly proportional to the ion beam current striking the target. The large area ion beam at the target is required to decrease the ion beam power density on the target, which would otherwise overheat the target and reduce neutron production. Compactness and high neutron output are achieved with the innovative extraction system and magnetic filter design. While the invention has been described with respect to a spherical electrode geometry, an alternate embodiment can be implemented with a cylindrical geometry, i.e. electrodes 26, 28 are cylindrical in shape (i.e. portions of cylinders), with aligned slots 30; suppressor electrode 32 is cylindrical, with central slot 34; and target 38 is cylindrical. The ion beam then focuses down to a line and expands to impinge on the target. The neutron generator of FIG. 1 has a tubular configuration, as shown in U.S. application Ser. No. 10/100,956. Other neutron generator configurations include cylindrical, as shown in Ser. No. 10/100,962, and spherical, as shown in Ser. No. 10/100,955. All these applications are herein incorporated by reference. The principles of the invention for ultra-short pulse production apply to any configuration. B. Ultra-short Pulse Production Ultra-short pulses of ions or neutrons, having pulse widths of about 1 μs or less with fast rise times or fall times or both, are produced by the design of the extraction system of the ion source and by controlling the extraction voltage. The ion beam current extracted from the ion source has an ultra-short pulse width by applying corresponding ultra-short voltage pulses on the extraction electrode. The pulse width is also controlled by designing the aperture(s) in the extraction system with a diameter that is not much greater than the plasma sheath thickness in the ion source, and by spacing the electrodes of the extraction system a distance about equal to the aperture diameter. To produce ultra-short neutron pulses, a neutron generating target is struck by accelerated ultra-short ion beam bursts of suitable ions, such as D, T, or D and T. In a typical ion source beam extraction system, the plasma potential is usually at a few volts above the plasma chamber potential (local ground) and the plasma electrode (the first or beam-forming electrode) is on the order of 10 volts below the local ground potential. The potential drop from the plasma potential to the plasma electrode potential occurs within a sheath region that has a thickness of about 10λD. The Debye shielding length λD is given by λ D = kT 4 ⁢ π ⁢ ⁢ ne 2 where T is the electron temperature and n is the plasma density. For a typical plasma with electron temperature T up to 10 eV and plasma density n at about 5×1011 cm3, 10λD is about 30 μm. Ions are accelerated from the plasma into the sheath while electrons are rejected by the sheath. However, if an aperture, on the plasma electrode is much larger than the sheath thickness, the sheath will “wrap around” the aperture, allowing the plasma to flow through the aperture without rejecting the electrons, i.e. the plasma simply leaks out of the aperture, preventing sharp narrow pulses from being formed. This situation is shown in FIG. 4A. The extraction system has a plasma electrode 50 and a spaced extraction electrode 52. A bias supply 54 is connected between electrodes 50, 52. A forward bias (electrode 52 is negative with respect to electrode 50) is applied for (positive) ion extraction and a reverse bias (electrode 52 is positive with respect to electrode 50) is applied to stop positive ions and for electron (and negative ion) extraction. Electrodes 50, 52 include one (or more) aligned apertures 56, 58 respectively. Plasma sheath 60 is adjacent to plasma electrode 50 and has a thickness t of about 30 μm. When the diameter d of aperture 56 in plasma electrode 50 is much greater than the plasma sheath thickness, i.e. d>>t, plasma leaks through aperture 56 around electrode 50. When a forward biased voltage is applied to extraction electrode 52, ions are accelerated and electrons are repelled, as shown in FIG. 4A. When a reverse biased voltage is applied to electrode 52, ions are repelled and electrons are accelerated, as shown in FIG. 4B. An electrode cloud 62 can build up between electrodes 50, 52 which can short out the electrodes. If the diameter of aperture 56 (and 58) is made smaller than the sheath thickness t, then the sheath 60 can cover the aperture, even in the reverse biased condition, as shown in FIG. 4C. Thus for micron sized apertures, most electrons cannot escape, even for a reverse bias voltage. Therefore, because of the ability to control ion extraction, micron sized apertures are preferred in the extractor system electrodes for producing ultra-short pulse widths. A multiple aperture multiple beamlet extraction system is thus preferred for the ion sources. To control the ion flow to produce good beam optics, the distance x between the plasma electrode 50 and the extraction electrode 52 must have approximately the same dimension as the aperture diameter d, i.e. an aspect ratio x/d of about 1. The potential required to repel ions at the extraction electrode is slightly above the plasma potential. Thus the voltage difference between the electrodes is about 20 V. The minimum required voltage gradient is 0.6 MV/m. In the forward bias case, the extraction electrode can be biased at local ground potential or some negative potential depending on the current density and beam optics design. This biasing effect has been experimentally demonstrated, using a single aperture setup as shown in FIG. 5. Experiments showed that ion as well as electron beams can be switched on and off using a biasing electrode 73 that stops the charged particles from exiting ion source 70. Biasing electrode 73 is part of a switchable extraction aperture system 77 that has two conducting electrodes 71, 73 separated by insulator layer 72. Electrode 71 is the plasma electrode and electrode 73 is the extraction electrode. System 77 is followed by insulator layer 74 and faraday cup 75. An aperture 76 is formed in the electrode and insulator layers. Electrode 71 is biased negatively (about 30 V) with respect to the chamber wall. Electrode 73 is used to stop the flow of ions by applying a positive bias with respect to the ion source chamber. Using argon as the working gas, a plasma discharge was produced with a discharge power of 40 W. The gas pressure inside the source was 2 mTorr. The source is biased at 30 V to allow the ions to be extracted, and the current is measured with the Faraday cup at ground potential. Electrode 71 is also biased with respect to the source to prevent back streaming electrons when the beam is switched on, and to avoid electron extraction when the beam is switched off. The beam energy at the Faraday cup is equal to the source potential plus the plasma potential. Because the discharge power is so low, the plasma potential is almost negligible. Thus, to read ion beam current at the Faraday cup, electrode 73 has to be biased equal to or less than the source. Experimentally, electrode 73 is first set at ground potential, which allows the ions to be extracted. The Faraday cup reads 23 nA. When electrode 73 is biased at 31 V, i.e. 1 V more positive than the source potential, the Faraday cup reading drops down to zero. Thus, by providing a micro-channel biasing system with a fast voltage switch, the invention enables one to generate ion and neutron beams with very short duration, about 1 μs or less and fast rise time and/or fall time. These ultra-short ion and neutron pulses can be used for a variety of applications, including neutron interrogation of nuclear materials and induction linacs. Changes and modifications in the specifically described embodiments can be carried out without departing from the scope of the invention which is intended to be limited only by the scope of the appended claims.
	description	The stimulable phosphor sheet of the invention is used in the radiation image recording and reproducing method described above, and is characterized by basically comprising a partition two-dimensionally dividing the plane of the stimulable phosphor sheet into small sections and stimulable phosphor-incorporated areas sectioned with that partition. The stimulable phosphor-incorporated areas have a reflection property (reflectance) differing from that of the partition at the wavelength of the stimulating rays. The partition is in the form of a grid to divide the plane into rectangular sections (i.e., rectangular stimulable phosphor-incorporated areas), and contains a stimulable phosphor or a UV-emitting phosphor (which absorbs the radiation and emits light in a UV or visible wavelength region). The partition containing a stimulable phosphor and the partition containing a UV-emitting phosphor are often referred to as xe2x80x9cstimulable phosphor-containing partitionxe2x80x9d and xe2x80x9cUV-emitting phosphor-containing partitionxe2x80x9d, respectively. Since the partition containing each phosphor prevents the simulating rays from diffusion or scattering in the phosphor sheet without impairing the sensitivity, the stimulable phosphor sheet of the invention gives an image of high sharpness with high sensitivity. From the viewpoint of fundamental performance, it is not necessary for the stimulable phosphor sheet of the invention to have a protective film and a support. The stimulable phosphor sheet, however, is preferably provided with them for ensuring safety in transportation and for avoiding deterioration, and hence a typical embodiment of the stimulable phosphor sheet of the invention comprises a support and a protective film on the bottom and on the top surface, respectively. By taking an example of the stimulable phosphor sheet having that structure (which is often referred to as xe2x80x9cradiation image storage panelxe2x80x9d), the invention is described below. In the following description, the stimulable phosphor sheet in the radiation image storage panel is often referred to as xe2x80x9cstimulable phosphor layerxe2x80x9d or simply xe2x80x9cphosphor layerxe2x80x9d. As the support, a sheet or a film of flexible resin material having a thickness of 50 xcexcm to 1 mm is usually employed. The support may be transparent or may contain light-reflecting material (e.g., titanium dioxide particles, barium sulfate particles) or voids for reflecting the stimulating rays or the stimulated emission. Further, it may contain light-absorbing material (e.g., carbon black) for absorbing the stimulating rays or the stimulated emission. Examples of the resin materials include polyethylene terephthalate, polyethylene naphthalate, aramid resin and polyimide resin. The support may be a sheet of other material such as metal, ceramics and glass, if needed. On the phosphor sheet-side surface of the support, auxiliary layers (e.g., light-reflecting layer, light-absorbing layer, adhesive layer, electro-conductive layer) or many hollows may be provided. On the other side surface, a friction-reducing layer or an anti-scratch layer may be formed. The stimulable phosphor layer (sheet) is provided on the support. The phosphor sheet according to the invention comprises the stimulable phosphor or UV-emitting phosphor-containing grid partition which divides two-dimensionally the plane of the sheet into small rectangular sections, and the stimulable phosphor-incorporated areas which are rectangularly sectioned with the grid partition and which have a reflectance differing from that of the grid partition at the wavelength of the stimulating rays. By referring to the attached drawings, the stimulable phosphor layer (sheet) having the aforementioned structure is described below. FIGS. 1-(1), -(2) and -(3) are sketchs showing a stimulable phosphor sheet 10 of the invention, a partial enlarged view of (1) and a partial sectional view of (2) sectioned with Ixe2x80x94I line, respectively. The shadowed portion in (2) and (3) indicates the stimulable or UV-emitting phosphor-containing grid partition 11, and the parts surrounded with the shadowed portion are the stimulable phosphor-incorporated areas 12. The thickness of the phosphor sheet (layer) is generally in the range of 20 xcexcm to 1 mm, preferably 50 xcexcm to 500 xcexcm. Preferably, the width of the partition is in the range of 5 xcexcm to 50 xcexcm, and each stimulable phosphor-incorporated area 12 has a width (in plane direction) of 20 xcexcm to 200 xcexcm on average. The top and the bottom of the partition in FIG. 1 appear on the surfaces of the sheet, but both or one of them may be buried under the phosphor sheet. Preferably, the height of the grid partition is in the range of 1/3 to 1/1 of the thickness of the phosphor sheet. FIG. 2-(1) shows a sectional view of another stimulable phosphor sheet of the invention in which the top of the partition is buried under the phosphor sheet. FIG. 2-(2) is a sectional view of the sheet of (1) provided with a support 13 and a protective film 14 on the bottom and the top surfaces, respectively. As the stimulable phosphor incorporated in the stimulable phosphor-containing grid partition and the stimulable phosphor-incorporated areas, a phosphor giving a stimulated emission of a wavelength in the range of 300 to 500 nm when it is irradiated with stimulating rays of a wavelength in the range of 400 to 900 nm is preferably employed. In Japanese Patent Provisional Publications No. 2-193100 and No. 4-310900, some examples of the stimulable phosphor are described in detail. Examples of the preferred phosphors include divalent europium or cerium activated alkaline earth metal halide phosphors (e.g., BaFBr:Eu, BaFBrI:Eu), and cerium activated oxyhalide phosphors. The stimulable phosphor in the partition and that in the phosphor-incorporated areas may be the same or different from each other, and usually are the same. Further, they may differ in composition but emit light in the same wavelength region when they are excited with stimulating rays of the same wavelength. As the UV-emitting phosphor incorporated in the UV-emitting phosphor-containing partition, a phosphor giving an emission peak in a UV or visible (preferably, UV) wavelength region is employed. In more detail, a phosphor absorbing a radiation of a wavelength not longer than 250 nm and immediately emitting light in the wavelength region of 250 to 400 nm (UV wavelength region) is preferably employed. Examples of the U-emitting phosphors include YTaO4, YTaO4:Gd, LnOX:Ac (in which Ln is Y, La, Gd and/or Lu; X is Cl, Br and/or I; Ac is Bi and/or Gd), LnF3:Ce (in which Ln is Y, La, Gd and/or Lu), GdF3, and BaF2. Further, UV light-emitting phosphors described in Japanese Patent Provisional Publication No. 2-176600 are also employable. The stimulable or UV light-emitting phosphor is usually used in the form of particles. The phosphor particles and a binder are well mixed in an appropriate solvent to prepare a coating dispersion. In the coating dispersion, the binder and the phosphor are introduced generally at a ratio of 1:1 to 1:100 (binder:phosphor, by weight), preferably 1:8 to 1:40 (by weight). As the binder material, various known resins are employable for forming the stimulable phosphor-incorporated areas, the stimulable phosphor-containing grid partition, or the UV light-emitting phosphor-containing grid partition. The stimulable phosphor sheet of the invention comprises the stimulable or UV light-emitting phosphor-containing grid partition which two-dimensionally divides the plane of the sheet into small sections and stimulable phosphor-incorporated areas which are rectangularly sectioned with the grid partition and which have a reflectance differing from that of the grid partition at the wavelength of the stimulating rays. With respect to the difference of reflectance, there are two cases: (1) the phosphor-incorporated area has a reflectance lower than that of the partition at the wavelength of the stimulating rays, and (2) the phosphor-incorporated area has a reflectance higher than that of the partition at the wavelength of the stimulating rays. The stimulable phosphor sheet of the embodiment (1) is preferred when a radiation image of particularly high sharpness is required. The reflectance of the phosphor-incorporated areas and that of the grid partition at the wavelength of the stimulating rays can be made higher by various methods. For example, the reflectance can be increased by relatively increasing the weight ratio of phosphor/binder, by relatively reducing the size (i.e., mean particle size) of the phosphor particles (i.e., by using fine particles of the phosphor), or by incorporating white pigments (e.g., titanium dioxide, barium sulfate) or stimulating ray-reflecting particles (e.g., phosphor particles other than the stimulable or UV light-emitting phosphor). These methods can be adopted singly or in combination. Further, a stimulating ray-reflecting thin film may be provided between the partition and the phosphor-incorporated areas. The reflectance at the wavelength of the stimulating rays can be made lower by, for example, relatively reducing the weight ratio of phosphor/binder, by relatively increasing the size (mean particle size) of the phosphor particles, or by incorporating dyes absorbing the stimulating rays. These methods can be adopted singly or in combination. Further, a stimulating ray-absorbing thin film may be provided between the partition and the phosphor-incorporated areas. The stimulable phosphor sheet of the invention can be produced, for example, in the following manner. First, a large number of stimulable phosphor films for forming the stimulable phosphor-incorporated areas and a large number of stimulable or UV light-emitting phosphor films for forming the grid partition are independently produced. From thus produced films, the phosphor sheet of the invention can be produced by the layered composition-slicing method illustrated in FIGS. 3 to 9 described below. FIG. 3 shows the stimulable phosphor film for forming the stimulable phosphor-incorporated areas (hereinafter, referred to as xe2x80x9cphosphor film Axe2x80x9d) and the stimulable or UV light-emitting phosphor film for forming the grid partition (hereinafter, referred to as xe2x80x9cphosphor film Bxe2x80x9d). The film A comprises a stimulable phosphor and a binder, and the film B comprises a stimulable or UV light-emitting phosphor and a binder. As shown in FIG. 4, a large number of the films A and B are alternately piled up to give a multi-layered composition. The composition is then heated under pressure in the manner shown in FIG. 5, so that the neighboring A and B films would be placed in close contact with each other to form a multi-layered composition block. The composition block is sliced repeatedly in the direction perpendicular to the film plane as shown in FIG. 6, to prepare a large number of striped phosphor films shown in FIG. 7 in which strips of the films A and B are alternately placed. The striped phosphor films shown in FIG. 7 and the aforementioned films B (for forming the grid partition) are alternately piled up to form a multi-layered composition shown in FIG. 8. The composition is then heated under pressure in the same manner as shown in FIG. 5, to form a multi-layered composition block. Finally, the composition block of FIG. 8 is sliced repeatedly in the direction perpendicular to the film plane as shown in FIG. 9, so that the appearing face would comprise the end faces of the strips of the films. Thus, the stimulable phosphor sheet of the invention is produced. On one surface of the sheet of the invention, a layer for reflecting the stimulating rays and/or the stimulated emission may be provided. However, if the phosphor sheet of the invention is to be used in a double-side reading system, it is preferred not to provide the reflecting layer. The reflecting layer enhances the sensitivity of the stimulable phosphor sheet, and can comprise white pigments (e.g., titanium dioxide particles, barium sulfate particles) or non-stimulable phosphor particles (which exhibit no stimulated emission) dispersed in a binder. The stimulable phosphor sheet of the invention is preferably placed on a support. In this case, the reflecting layer is usually provided between the stimulable phosphor sheet and the support. In place of the reflecting layer, a stimulating ray-absorbing layer can be provided between them. The absorbing layer is effective for particularly improving the sharpness of the radiation image. On the surface not facing the support, the stimulable phosphor sheet of the invention preferably has a protective film. In order not to affect the simulating rays or the stimulated emission, the film is preferably transparent. Further, for efficiently protecting the stimulable phosphor sheet from chemical deterioration and physical damage, the protective film should be both chemically stable and physically strong. The protective film can be provided by fixing a beforehand prepared transparent plastic film on the stimulable phosphor sheet with adhesive, or by coating the phosphor sheet with absolution of protective film material and drying the coated solution. Into the protective film, fine particle filler may be incorporated so as to reduce blotches caused by interference and to improve the quality of the resultant image. Examples of preferable materials for the transparent plastic film include polyester resins (e.g., polyethylene terephthalate, polyethylene naphthalate), cellulose derivatives (e.g., cellulose triacetate), and various other resin materials such as polyolefin and polyamide. The thickness of the protective film is generally in the range of not more than 30 xcexcm, preferably 1 to 15 xcexcm, more preferably 5 to 12 xcexcm. For enhancing the resistance to staining, a fluororesin layer is preferably provided on the protective film. The fluororesin layer can be formed by coating the surface of the protective film with a solution of a fluororesin in an organic solvent, and drying the coated solution. The fluororesin may be used singly, but generally a mixture of the fluororesin and a film-forming resin is employed. In the mixture, an oligomer having polysiloxane structure or perfluoroalkyl group can be further added. The coating can be performed using known coating means such as doctor blade, roll coater, and knife coater. Into the fluororesin layer, fine particle filler may be incorporated so as to reduce blotches caused by interference and to improve the quality of the resultant image. The thickness of the fluororesin layer generally is in the range of 0.5 to 20 xcexcm, preferably 1 to 5 xcexcm. In the formation of the fluororesin layer, additives such as a crosslinking agent, a film-hardening agent and an anti-yellowing agent can be used. In particular, the crosslinking agent advantageously improves durability of the fluororesin layer. The aforementioned layered composition-slicing method can be also advantageously employed for producing the sheet comprising a light-reflecting material-containing grid partition containing no phosphor. The method can be used for producing a stimulable phosphor sheet used in the radiation image recording and reproducing method comprising the steps of recording a radiographic image as a latent image, irradiating the latent image with stimulating rays to release stimulated emission, and electrically processing the emission to reproduce the radiation image; the sheet comprising a light-reflecting material-made grid partition two-dimensionally dividing the plane of the sheet into small rectangular sections, and stimulable phosphor-incorporated areas rectangularly sectioned with the grid partition. The light-reflecting material-made grid partition surrounds the phosphor-incorporated areas and reflects the stimulating rays and/or the stimulated emission, and hence encloses the rays and/or the emission in the phosphor-incorporated areas. Consequently, the partition prevents the stimulating rays and/or the stimulated emission from diffusing or scattering horizontally, so as to improve the sharpness of the resultant image. When the stimulable phosphor sheet comprising a light-reflecting material-containing grid partition is produced by the slicing method, the procedure illustrated in FIGS. 3 to 9 is performed in the same manner except that light-reflecting material films are used in place of the phosphor films B. As the light-reflecting material film, a film comprising light-reflecting particles and a binder is usually employed. Examples of the light-reflecting particles include white pigments such as titanium dioxide particles and barium sulfate particles, but they by no means restrict the invention. As the binder, various known materials described above are usable. The stimulable phosphor sheet of the invention can be used either in a conventional single-side reading system (in which the stimulated emission is collected from the side having been exposed to the stimulating rays or from the other side) or in a known double-side reading system. The stimulable phosphor sheet of the invention is particularly effective in the double-side reading system. 1) Stimulable phosphor (BaFBr:Eu) particles (median of the particle sizes: 5 xcexcm) and a thermoplastic high molecular weight-polyester resin were dispersed in an organic solvent in a weight ratio of 20:1. The prepared phosphor dispersion was coated on a temporary support having a releasing surface, and dried to give a dry phosphor layer. The phosphor layer thus formed was then peeled from the temporary support to give a stimulable phosphor film (1) (thickness: approx. 100 xcexcm). 2) Stimulable phosphor (BaFBr:Eu) particles (median of the particle sizes: 5 xcexcm), alumina particles (median of the particle sizes: 1 xcexcm) and a thermoplastic high molecular weight-polyester resin were dispersed in an organic solvent in a weight ratio of 40:20:3. The prepared alumina-phosphor dispersion was coated onto a temporary support having a releasing surface, and dried to give a dry alumina-containing phosphor layer. The alumina-containing phosphor layer thus formed was then peeled from the temporary support to give an alumina-containing stimulable phosphor film (2) (thickness: approx. 30 xcexcm). 3) Each of the stimulable phosphor film (1) and the alumina-containing stimulable phosphor film (2) was cut to give 350 square pieces (40 mmxc3x9740 mm). The pieces of the films (1) and (2) were alternately piled up to form a multi-layered composition consisting of 700 layers. The composition was then heated under pressure (pressure: approx. 1 kg/cm2, temperature: 100xc2x0 C.) for 1 hour to produce a multi-layered composition block (1). 4) The multi-layered composition block (1) was repeatedly sliced vertically to the layer plane with a wide microtome, to produce 200 sheets of striped phosphor film (3) (thickness: 100 xcexcm). 50 200 sheets of the striped film (3) and 200 sheets of the alumina-containing film (2) were alternately piled up to form a multi-layered composition consisting of 400 layers. The composition was then heated under pressure (pressure: approx. 1 kg/cm2, temperature: 100xc2x0 C.) for 1 hour to produce a multi-layered composition block (2). 6) The multi-layered composition block (2) was sliced vertically to the layer plane with a wide microtome so that the appearing face would comprise the end faces of the strips, to produce a stimulable phosphor sheet (1) (thickness: approx. 215 xcexcm) having a grid structure on the surface. 1) Stimulable phosphor (BaFBr:Eu) particles (median of the particle sizes: 5 xcexcm), alumina particles (median of the particle sizes: 1 xcexcm) and a thermoplastic high molecular weight-polyester resin were dispersed in an organic solvent in a weight ratio of 20:40:3. The prepared alumina-phosphor dispersion was coated on a temporary support having a releasing surface, and dried to form a dry alumina-containing phosphor layer. The alumina-containing phosphor layer thus formed was then peeled from the temporary support to give an alumina-containing stimulable phosphor film (4) (thickness: approx. 30 xcexcm). 2) The procedure of Example 1 was repeated except for using the above-prepared alumina-containing stimulable phosphor film (4) in place of the alumina-containing phosphor film (2), to produce a stimulable phosphor sheet (2) (thickness: approx. 215 xcexcm) having a grid structure on the surface. 1) Stimulable phosphor (BaFBr:Eu) particles (median of the particle sizes: 1 xcexcm) and a thermoplastic high molecular weight-polyester resin were dispersed in an organic solvent in a weight ratio of 20:1. The prepared phosphor dispersion was coated on a temporary support having a releasing surface, and dried to form a dry phosphor layer. The layer thus formed was then peeled from the temporary support to give a stimulable phosphor film (5) (thickness: approx. 30 xcexcm). 2) The procedure of Example 1 was repeated except for using the above-prepared stimulable phosphor film (5) in place of the alumina-containing phosphor film (2), to produce a stimulable phosphor sheet (3) (thickness: approx. 215 xcexcm) having a grid structure on the surface. 1) Stimulable phosphor (BaFBr:Eu) particles (median of the particle sizes: 3 xcexcm) and a thermoplastic high molecular weight-polyester resin were dispersed in an organic solvent in a weight ratio of 20:1. The prepared phosphor dispersion was coated on a temporary support having a releasing surface, and dried to form a dry phosphor layer. The phosphor layer thus formed was then peeled from the temporary support to give a stimulable phosphor film (6) (thickness: approx. 30 xcexcm). 2) The procedure of Example 1 was repeated except for using the above-prepared stimulable phosphor film (6) in place of the alumina-containing phosphor film (2), to produce a stimulable phosphor sheet (4) (thickness: approx. 215 xcexcm) having a grid structure on the surface. 1) Stimulable phosphor (BaFBr:Eu) particles (median of the particle sizes: 5 xcexcm) and a thermoplastic high molecular weight-polyester resin were dispersed in an organic solvent in a weight ratio of 25:1. The prepared phosphor dispersion was coated on a temporary support having a releasing surface, and dried to form a dry phosphor layer. The phosphor layer thus formed was then peeled from the temporary support to give a stimulable phosphor film (7) (thickness: approx. 30 xcexcm). 2) The procedure of Example 1 was repeated except for using the above-prepared stimulable phosphor film (7) in place of the alumina-containing phosphor film (2), to produce a stimulable phosphor sheet (5) (thickness: approx. 215 xcexcm) having a grid structure on the surface. 1) Stimulable phosphor (BaFBr:Eu) particles (median of the particle sizes: 5 xcexcm) and a thermoplastic high molecular weight-polyester resin were dispersed in an organic solvent in a weight ratio of 30:1. The prepared phosphor dispersion was coated on a temporary support having a releasing surface, and dried to form a dry phosphor layer. The phosphor layer thus formed was then peeled from the temporary support to give a stimulable phosphor film (8) (thickness: approx. 30 xcexcm). 2) The procedure of Example 1 was repeated except for using the above-prepared stimulable phosphor film (8) in place of the alumina-containing phosphor film (2), to produce a stimulable phosphor sheet (6) (thickness: approx. 215 xcexcm) having a grid structure on the surface. 1) Stimulable phosphor (BaFBr:Eu) particles (median of the particle sizes: 5 xcexcm), ultramarine particles and a thermoplastic high molecular weight-polyester resin were dispersed in an organic solvent in a weight ratio of 40:0.2:3. The prepared dye-containing phosphor dispersion was coated on a temporary support having a releasing surface, and dried to form a dye/phosphor layer. The dye/phosphor layer thus formed was then peeled from the temporary support to give an ultramarine-containing stimulable phosphor film (9) (thickness: approx. 30 xcexcm). 2) The procedure of Example 1 was repeated except for using the above-prepared stimulable phosphor film (9) in place of the alumina-containing phosphor film (2), to produce a stimulable phosphor sheet (7) (thickness: approx. 215 xcexcm) having a grid structure on the surface. 1) UV light-emitting phosphor (YTaO4) particles (median of the particle sizes: 1 xcexcm) and a thermoplastic high molecular weight-polyester resin were dispersed in an organic solvent in a weight ratio of 20:1. The prepared phosphor dispersion was coated on a temporary support having a releasing surface, and dried to form a phosphor layer. The phosphor layer thus formed was then peeled from the temporary support to prepare a UV light-emitting phosphor film (10) (thickness: approx. 30 xcexcm). 2) The procedure of Example 1 was repeated except for using the above-prepared UV light-emitting phosphor film (10) in place of the alumina-containing phosphor film (2), to produce a stimulable phosphor sheet (8) (thickness: approx. 215 xcexcm) having a grid structure on the surface. 1) UV light-emitting phosphor (GdF3) particles (median of the particle sizes: 1 xcexcm) and a thermoplastic high molecular weight-polyester resin were dispersed in an organic solvent in a weight ratio of 20:1. The prepared phosphor dispersion was coated on a temporary support having a releasing surface, and dried to form a phosphor layer. The phosphor layer thus formed was then peeled from the temporary support to prepare a UV light-emitting phosphor film (11) (thickness: approx. 30 xcexcm). 2) The procedure of Example 1 was repeated except for using the above-prepared UV light-emitting phosphor film (11) in place of the alumina-containing phosphor film (2), to produce a stimulable phosphor sheet (9) (thickness: approx. 215 xcexcm) having a grid structure on the surface. 1) Stimulable phosphor (YLuSiO5:Ce,Zr) particles (median of the particle sizes: 5 xcexcm) and a thermoplastic high molecular weight-polyester resin were dispersed in an organic solvent in a weight ratio of 20:1. The prepared phosphor dispersion was coated on a temporary support having a releasing surface, and dried to form a phosphor layer. The phosphor layer thus formed was then peeled from the temporary support to give a stimulable phosphor film (12) (thickness: approx. 100 xcexcm). 2) (UV light-emitting phosphor (YTaO4) particles (median of the particle sizes: 1 xcexcm) and a thermoplastic high molecular weight-polyester resin were dispersed in an organic solvent in a weight ratio of 20:1. The prepared phosphor dispersion was coated on a temporary support having a releasing surface, and dried to form a phosphor layer. The phosphor layer thus formed was then peeled from the temporary support to prepare a uv light-emitting phosphor film (13) (thickness: approx. 30 xcexcm). 3) Each of the stimulable phosphor film (12) and the UV light-emitting phosphor film (13) was cut to give 350 square pieces (40 mmxc3x9740 mm). The pieces of the films (12) and (13) were alternately piled up to form a multi-layered composition consisting of 700 layers. The composition was then heated under pressure (pressure: approx. 1 kg/cm2, temperature: 100xc2x0 C.) for 1 hour to produce a multi-layered composition block (3). 4) The multi-layered composition block (3) was repeatedly sliced vertically to the layer plane with a wide microtome, to produce 200 sheets of striped phosphor film (14) (thickness: 100 xcexcm). 5) 200 sheets of the striped film (14) and 200 sheets of the UV light-emitting phosphor film (13) were alternately piled up to give a multi-layered composition consisting of 400 layers. The composition was then heated under pressure (pressure: approx. 1 kg/cm2, temperature: 100xc2x0 C.) for 1 hour to produce a multi-layered composition block (4). 6) The multi-layered composition block (4) was sliced vertically to the layer plane with a wide microtome so that the appearing face would comprise the end faces of the strips, to produce a stimulable phosphor sheet (10) (thickness: 215 xcexcm) having a grid structure on the surface. 1) UV-emitting phosphor (GdF3) particles (median of the particle sizes: 1 xcexcm) and a thermoplastic high molecular weight-polyester resin were dispersed in an organic solvent in a weight ratio of 20:1. The prepared phosphor dispersion was coated on a temporary support having a releasing surface, and dried to form a phosphor layer. The phosphor layer thus formed was then peeled from the temporary support to prepare a uv light-emitting phosphor film (15) (thickness: approx. 30 xcexcm). 2) The procedure of Example 10 was repeated except for using the above-prepared UV light-emitting phosphor film (15) in place of the UV light-emitting phosphor film (13), to produce a stimulable phosphor sheet (11) (thickness: approx. 215 xcexcm) having a grid structure on the surface. 1) Alumina particles (median of the particle sizes: 1 xcexcm) and an acrylic polymer resin were dispersed in an organic solvent in a weight ratio of 20:1. The prepared alumina dispersion was coated on a temporary support having a releasable surface, and dried to form a phosphor layer. The phosphor layer thus formed was then peeled from the temporary support to prepare an alumina film (thickness: approx. 30 xcexcm). 2) The procedure of Example 1 was repeated except for using the above-prepared alumina film in place of the alumina-containing phosphor film (2), to produce a stimulable phosphor sheet (12) (thickness: approx. 215 xcexcm) having a grid structure on the surface. The procedure of Example 10 was repeated except for using the alumina film produced in Example 12 in place of the UV emitting-emitting phosphor film (13), to produce a stimulable phosphor sheet (13) (thickness: approx. 215 xcexcm) having a grid structure on the surface. 1) Stimulable phosphor (SrS:Ce,Sm) particles (median of the particle sizes: 5 xcexcm) and a thermoplastic high molecular weight-polyester resin were dispersed in an organic solvent in a weight ratio of 20:1. The prepared phosphor dispersion was coated on a temporary support having a releasing surface, and dried to form a layer. The phosphor layer thus formed was then peeled from the temporary support to give a stimulable phosphor film (16) (thickness: approx. 100 xcexcm). 2) UV light-emitting phosphor (BrFBr:Eu) particles (median of the particle sizes: 1 xcexcm) and a thermoplastic high molecular weight-polyester resin were dispersed in an organic solvent in a weight ratio of 20:1. The prepared phosphor dispersion was coated on a temporary support having a releasing surface, and dried to form a phosphor layer. The phosphor layer thus formed was then peeled from the temporary support to give a UV light-emitting phosphor film (17) (thickness: approx. 30 xcexcm) 3) Each of the stimulable phosphor film (16) and the UV light-emitting phosphor film (17) was cut to give 350 square pieces (40 mmxc3x9740 mm). The pieces of the films (16) and (17) were alternately piled up to form a multi-layered composition consisting of 700 layers. The composition was then heated under pressure (pressure: approx. 1 kg/cm2, temperature: 100xc2x0 C.) for 1 hour to produce a multi-layered composition block (7). 4) The multi-layered composition block (7) was repeatedly sliced vertically to the layer plane with a wide microtome, to produce 200 sheets of striped phosphor film (18)(thickness: 100 xcexcm). 5) 200 sheets of the striped film (18) and 200 sheets of the UV-emitting phosphor film (17) were alternately piled up to form a multi-layered composition consisting of 400 layers. The composition was then heated under pressure (pressure: approx. 1 kg/cm2, temperature: 100xc2x0 C.) for 1 hour to produce a multi-layered composition block (8). 6) The multi-layered composition block (8) was sliced vertically to the layer plane with a wide microtome so that the appearing face would comprise the end faces of the strips, to produce a stimulable phosphor sheet (14) (thickness: 215 xcexcm) having a grid structure on the surface. 1) UV light-emitting phosphor (BrFBr:Eu) particles (median of the particle sizes: 1 xcexcm) and a thermoplastic high molecular weight-polyester resin were dispersed in an organic solvent in a weight ratio of 20:1. The prepared phosphor dispersion was coated on a temporary support having a releasing surface, and dried to form a phosphor layer. The phosphor layer thus formed was then peeled from the temporary support to prepare a UV light-emitting phosphor film (19) (thickness: approx. 30 xcexcm). 2) The procedure of Example 14 was repeated except for using the above-prepared UV light-emitting phosphor film (19) in place of the UV light-emitting phosphor film (17), to produce a stimulable phosphor sheet (15) (thickness: approx. 215 xcexcm) having a grid structure on the surface. 1) Alumina particles (median of the particle sizes: 1 xcexcm) and an acrylic polymer resin were dispersed in an organic solvent in the weight ratio of 20:1. The prepared alumina dispersion was applied onto a temporary support having the surface beforehand subjected to releasing surface treatment, and dried to form a layer. The layer thus formed was then peeled from the temporary support to prepare an alumina film (thickness: approx. 30 xcexcm). 2) The procedure of Example 14 was repeated except for using the above-prepared alumina film in place of the UV-emitting phosphor film (17), to produce a stimulable phosphor sheet (16) (thickness: approx. 215 xcexcm) having a grid structure on the surface. The procedure of Example 3 was repeated except that the ratio of phosphor/binder for the phosphor-incorporated area and that in the partition were set at 5/1 and 15/1, respectively, to produce a stimulable phosphor sheet (17) (thickness: approx. 215 xcexcm) having a grid structure on the surface. The procedures of Example 13 were repeated except that the ratio of phosphor/binder in the phosphor-incorporated area was set at 5/1 and the ratio of alumina/binder (acrylic resin) in the grid partition was set at 15/1, to produce a stimulable phosphor sheet (18) (thickness: approx. 215 xcexcm) having a grid structure on the surface.
051165660	claims	1. Apparatus for refurbishing the ends of a plurality of control rods where the rods are disposed in a grid of rows and columns with a known predetermined spacing therebetween, comprising: a. support block means including means for rapidly clamping and unclamping said support block to two adjacent rods of said control rods in a given column or row; b. motor mounting means coupled to said support block means and reciprocally movable relative to said support block means in a direction parallel to the longitudinal axis of said tubular rods, said motor mounting means including a bearing housing; c. a rotatable spindle member journalled for rotation in said bearing housing for supporting an end mill bit, said spindle having its axis of rotation generally coaxially aligned with the longitudinal axis of a given one of said control rods when said support block means is clamped to said two adjacent ones of said tubular rods; d. motor means mounted on said motor mounting means in driving relation to said spindle member for rotating said spindle about its axis of rotation; and e. means for translating said motor mounting means reciprocally relative to said support block means for selectively bringing said end mill bit into contact with the end of said given one of said control rods as said spindle member is rotated. a) first and second parallel slots formed in said support block means and spaced from one another by said known predetermined spacing, said slots being aligned with the longitudinal axis of said control rods, the first and second slots having a width greater than the outer diameter of said tubular rods; b) a plurality of clamping bars pivotally coupled at a first end to said support block for rotation in planes transverse to said parallel slots; and c) thumb screw means for securing the second end of said clamping bars relative to said support block with said clamping bars intersecting said parallel slots. a) A rack gear secured to one of said support block means and motor mounting means and extending parallel to the longitudinal axis of said control rods when said support block means is clamped to adjacent ones of said control rods; b) a pinion gear journalled for rotation in the other of said support block means and said motor mounting means and engaging said gear rack; and c) means for rotating said pinion gear. a) a pair of parallel track slots extending inwardly from opposed side surfaces of said support block; b) a rack gear receiving slot formed inwardly from the edge surface of said support block extending between said opposed side surfaces; and c) said motor mounting means including (i) a pair of rails for slidingly engaging said parallel track slots formed in said support block; and (ii) a pinion gear receiving bore extending transverse to said rack gear receiving slot. a) selecting a first pair of CRDs adjacent to a given CRD to be refurbished; b) clamping a milling machine to the selected pair, said milling machine including: c) advancing the end mill into contact with said given CRD to be refurbished; d) driving the spindle until the CRD is milled down below the location of its welded end cap; e) disengaging the end mill from contact with the milled CRD; f) removing said milling machine from said first pair of CRDs; g) selecting another pair of CRD's adjacent to a CRD to be refurbished in accordance with steps (a) through (d). 2. The apparatus of claim 1 wherein said means for rapidly clamping and unclamping comprises: 3. The apparatus as in claim 1 wherein said means for translating comprises: 4. The apparatus as in claim 1 wherein said motor means is fluid driven. 5. The apparatus as in claim 3 wherein said support block includes: 6. A method of refurbishing welds on control rod drive penetrations in a boiling water nuclear reactor, said CRD's having end caps welded therein and being arranged in a grid of columns and rows comprising the steps:
	description	This application is the U.S. national phase of International Application No. PCT/RU2017/000817 filed Nov. 3, 2017, which claims priority to Russian Patent Application Nos. RU 2016151504 and RU 2016151505 filed Dec. 26, 2016, the disclosures of which are incorporated in their entirety by reference herein. The invention relates to fuel assembly designs for nuclear reactors and can be used in fast neutron reactor cores. A fuel assembly for nuclear reactors with rod-type fuel elements is a top nozzle and a bottom nozzle connected to one another by a structural frame providing the fuel assembly with the necessary strength and rigidity. Between the top nozzle and the bottom nozzle of the fuel assembly, a fuel element bundle allowing for free temperature and radiation expansion is installed with the aid of retaining and spacing elements along the fuel assembly axis in the reactor core. The top nozzle and the bottom nozzle of the fuel assembly are fitted with means for installing and retaining the fuel assembly in the reactor core, and also with means for removing the fuel assembly from the reactor. Different embodiments of the fuel assembly designs with rod-type fuel elements for thermal and fast neutron reactors have been developed; they differ from one another in the specific design of the aforementioned structural elements and in the materials they are made of. A fuel assembly for the WWER-type nuclear reactor is known comprising top nozzle, a bottom nozzle and a central pipe connected to one another by a jacket of hexagonal form, inside which a fuel element bundle in the form of fuel kernels in sealed cylindrical claddings is placed using retaining and spacing elements (RU2088982). Means for retaining fuel elements in a fuel assembly and for their transverse spacing in the bundle are provided in the form of top and bottom end grids and several spacing grids installed between the end grids. The fuel element butts are fixed in the end grids, wherein the top grid is configured to move along the axial direction in slots provided in the jacket corners. Spacing grids are installed on the central pipe and can move along the axial direction in slots provided in the central pipe. The use in the known solution of several grids for retaining and spacing of fuel elements results in increased hydraulic resistance for coolant flow and worse heat transfer conditions. Moreover, corrosion and radiation-induced creep of the grid material in the core section leads to a change in the geometry of cells and in the elastic properties of the grid materials, which increases the probability of vibration and fretting of the cladding of individual fuel elements. A design for fuel assemblies for power thermal and fast neutron nuclear reactors is known comprising a top nozzle, a bottom nozzle, a frame, top and bottom end grids, spacing elements and a bundle of rod-type fuel elements arranged in the fuel assembly with the aid of the end grids and spacing elements (RU 2340019). The frame connects the top nozzle and the bottom nozzle and is made in the form of a jacket, a central pipe with end grids installed thereon, and tie bars installed in end grids along the fuel assembly perimeter. The spacing elements are made in the form of longitudinal tubes installed in parallel between the fuel elements and fixed in end grids, and also in the form of perforated shell rings arranged inside the fuel assembly along the jacket perimeter. The spacing element tubes can be made with a longitudinal slot and cuts forming cylindrical spacing ribs connected with each other and arranged along the height of the fuel assembly at a specified pitch. The frame is fitted with several retaining elements, which wrap around the fuel element bundle and are installed along the height of the frame. The retaining elements are made of a material (for example, a molybdenum-based material) with a coefficient of linear expansion lower than the coefficient of the fuel element claddings (for example, steels EP-823). The invention is intended to ensure reliable spacing of fuel elements in the reactor core. One drawback of the known design of the fuel assembly is its increased metal intensity due to the presence of a frame in the form of a jacket, support elements, central pipe and perforated shell rings. The fuel assembly has a complex structure providing for installation of a fuel element bundle, tube-type spacing elements and a system of frame tie bars in the bottom and top end grids. This results in reduction of the flow section for passage of coolant in the grids area and increased hydraulic resistance of the fuel assembly. A fast neutron nuclear reactor fuel assembly for BN-type reactors with rod-type fuel elements is known including a top nozzle, a bottom nozzle and a jacket in the form of a hexagonal section pipe (Development, Production and Operation of Fuel Elements of Power Reactors. Moscow, Energoizdat, 1995, Book 2, page 158). A fuel element bundle with cladding of an external diameter within the range of about 6 mm and a thickness of cladding of about 0.3 mm is placed inside the fuel assembly with the aid of retaining and spacing elements. Stainless steel is used for fabrication of fuel assembly elements and fuel element cladding. The means for retaining and spacing of fuel element bundle in the fuel assembly are made in the form of a support grid and wires with a diameter of about 1 mm wrapped in a spiral with a pitch of about 100 mm around the fuel element cladding surface. In order to equalize the release of energy and consumption of coolant along the core section, the peripheral row of fuel elements (fuel elements installed directly at the jacket wall) in the fuel assembly is spaced with the aid of wires with an oval cross section of 1.3×0.6 mm. One drawback of this design for the fuel element is the rigidity of the spacing element in the transverse plane. As a result, swelling of fuel and increase in the diameter of the fuel element claddings in the process of fuel irradiation in the core lead to an increase in local stress in the thin-walled cladding of fuel elements, its deformation and the acceleration of pitting corrosion processes in the zone of contact between the cladding and the spacing element. Moreover, the use of a spacing element in the form of a wire with an oval cross section substantially changes the conditions for placing and spacing peripheral fuel elements inside a fuel assembly. Thus, along the height of the fuel assembly equal to the pitch of the wire wrapping, each fuel element placed inside the bundle with a triangular grid has contact with adjacent fuel elements at 12 points. Meanwhile, 6 contacts are formed between its spacing element and the claddings of adjacent fuel elements, and 6 more contacts are formed between the cladding of this fuel element and the spacing elements of the six adjacent fuel elements. Each fuel element in the peripheral row (except for the corner elements) is adjacent to two other peripheral fuel elements, two internal fuel elements in the bundle, and the jacket. Along the height of the fuel assembly equal to the pitch of the wire wrapping, each peripheral fuel element has three spacing contacts only: two contacts between its cladding and the round wires of two internal fuel elements in the bundle, and one contact between its wire with an oval cross section and the fuel assembly jacket. This is due to the fact that the oval-shaped spacing element is not in contact with the claddings of the four adjacent fuel elements and the gap between them is 0.45 mm (1.05-0.6). In this way, the peripheral fuel elements of the known solution have four times (12/3) fewer spacing contacts than the fuel elements arranged inside the bundle. When compensating for spacing loads in a fuel element bundle, this results in a significant increase in stress within the zones of contact between the claddings of peripheral fuel elements and the spacing elements of internal fuel elements and the fuel assembly jacket, which contributes to the acceleration of pitting corrosion processes acting on the cladding and increases the probability of destruction of its integrity. One drawback of the known solution is the complexity of the fabrication of fuel elements and fuel assemblies, in which fuel elements with an increased diameter of cladding (about 10 mm) and wire (about 3 mm) and a distance between the claddings of adjacent fuel elements of more than 3 mm are used. In this case, the use of the known design is related to the massive and rigid wire made of stainless steel wrapped around the thin-walled cladding of fuel elements, which results in deformation of the cladding and disruption of the geometrical form of the fuel element (for example, bending of the element). The objective of the invention is to improve the reliability of a fuel assembly in the core of a nuclear reactor, improve the metal intensity of a fuel assembly, and develop a fuel assembly design for the use in lead-cooled fast neutron reactors and reactors with nitride uranium-plutonium fuel. The technical result of the invention is to increase the reliability of a fuel element bundle spacing in a fuel assembly and reduce the local stress in the claddings of fuel elements within the region, in which said elements are in contact with spacing elements. The technical result is achieved by the fact that, in the fuel assembly (including a top nozzle and a bottom nozzle connected to one another with the aid of a jacket; a bundle of rod-type fuel elements arranged in the fuel assembly with the aid of a grid and spiral spacing elements wrapped around the cladding of each fuel element and fixed at their ends), at least the peripheral fuel elements in a bundle are provided with spacing elements in the form of thin-walled tubes with longitudinal through slots, wherein the elements have a substantially oval cross section in the regions where they are in contact with the jacket. Possible specific embodiments of fuel assemblies are characterized by the following parameters: all the fuel elements of a fuel assembly are provided with spacing elements (hereinafter elements) in the form of thin-walled round tubes with longitudinal through slots and the same external diameter; peripheral fuel elements are provided with elements in the form of thin-walled round tubes with longitudinal through slots, while the fuel elements inside the bundle are provided in the form of wire, wherein the elements and wires have the same external diameter, the width of the slot in the elements is set within 0.1 to 0.35 of its diameter the thickness of the element wall is set within 0.25 to 1 of the thickness of the fuel element cladding; parts of peripheral fuel elements have reduced resistance to deformation in the transverse plane as compared to the other fuel elements in the bundle; the width of the slots in peripheral fuel elements is set between 0.20 and 0.35, and in other fuel elements—between 0.1 and 0.30 of the element diameter; the wall thickness of peripheral fuel elements is set between 0.25 and 0.6, while that of the other fuel elements is between 0.4 and 1 of the fuel element cladding thickness. The technical result is also achieved by the fact that, in the method for producing a fuel assembly (including forming a fuel element bundle with spiral spacing elements, placing the fuel element bundle in a jacket, and connecting the jacket with the top and bottom nozzles of the fuel assembly), at least the peripheral fuel elements in a bundle are provided with spacing elements in the form of thin-walled round tubes with longitudinal through slots, and the fuel elements assume a substantially oval cross section in the regions where they are in contact with the jacket through transverse compression of fuel element bundles during fabrication of a fuel assembly. Possible specific options for implementing the fuel assembly production method are characterized by the following parameters: prior to the insertion of a fuel element bundle into a jacket, the bundle is compressed in the transverse plane with the aid of several hexagonal compressing rims, which are removed from the bundle surface one by one as it is inserted into the jacket; the fuel element bundle is compressed within the limits of the elastic deformation of the elements. The essence of the invention consists in provision of a fuel element bundle with spacing elements in the form of spiral thin-walled tubes with longitudinal through slots, as well as in establishment of an aggregate of features ensuring the reliability of spacing of both internal and peripheral fuel elements and the reduction of local stresses in fuel element claddings. The technical result is achieved due to the fact that the proposed solution allows the required cross profile of peripheral fuel elements to be formed: a round profile in the zone of contact with four adjacent fuel elements of the bundle, and an oval profile in the zone of contact with the jacket. The oval cross section of the element in its zone of contact with the jacket is formed directly from the round cross section during transverse compression and installation of the bundle of fuel elements in the jacket in the course of assembling the fuel assembly. Local deformation of peripheral fuel elements is achieved under the impact of two factors: a) concentration of expanding pressures of fuel element bundles in the zones of contact between the elements and the jacket, and b) the reduced resistance of these elements to deformation in the transverse direction compared to other fuel assembly elements. Reduced resistance to deformation is the result of a relative increase in the slot width and a decrease in the thickness of the walls of the peripheral fuel elements. At the first stage of compression of a bundle, technological gaps are selected and an oval cross section of elements is formed in their zones of contact with the jacket; next, the required spacing tension of the fuel element bundle in the jacket is achieved. Improved reliability of fuel element spacing is also achieved due to the fact that the proposed solution ensures, at each pitch of wrapping of the element, 9 points of contact for the dimensional spacing of each peripheral fuel element. Five contacts are formed by the fuel element and the cladding of four adjacent fuel elements and the jacket, and four contacts are formed by the cladding of this fuel element with adjacent fuel elements. As compared to the known fuel assembly described in the prototype, the invention 3 times increases the number of spacing contacts of peripheral fuel elements threefold (9/3). The solution makes it possible to substantially increase the uniformity and decrease the degree of local mechanical stresses both in the claddings of peripheral fuel elements and in the claddings of other fuel elements of the bundle occurring due to the impact of the elements upon the temperature and radiation swelling of fuel. The technical result is also achieved by the use of a method for producing a fuel assembly, according to which at least peripheral fuel elements in a bundle are provided with spacing elements in the form of thin-walled round tubes with longitudinal through slots, and the fuel elements assume a substantially oval cross section in the regions where they are in contact with the jacket through transverse compression of fuel element bundles during fabrication of a fuel assembly. Specific exemplary embodiments of the fuel assembly and variations on the method for producing it using the proposed solution are given in more detail below. The preferred embodiment is provision of all fuel assembly elements with the elements in the form of thin-walled round tubes with longitudinal through slots and the same external diameter. This design permits both substantial improvement in the reliability of spacing of peripheral fuel elements and a decrease in local stresses in the claddings of all fuel elements of a bundle. This is achieved by deformation of elements given a temperature and radiation increase in the cladding diameter. One possible embodiment of the invention is the provision of peripheral fuel elements with elements in the form of thin-walled round tubes with longitudinal through slots, and provision of the fuel elements inside the bundle with wire spacing elements. In this case, the exterior diameters of the elements and wires are equal, providing reliable spacing of all fuel elements of the bundle. Some reduction of local stresses in the claddings of all fuel elements of a bundle is provided due to additional deformation of peripheral fuel elements. However, this design can be implemented only for the fuel assemblies, in which spacing elements with a relatively small diameter (up to 1.5-2 mm) are used. The fuel assembly design ensures compensation for temperature and radiation swelling of fuel element claddings in the reactor core by making possible deformation of the element when spacing loads increase. Deformation of an element in the transverse direction within the limits of the slot width takes place due to bending of the element profile in the transverse plane, and does not result in a significant increase in contact stress in the cladding of the fuel element, which improves its reliability at high levels of fuel burnout. Moreover, this element design ensures input and output of coolant in the reactor core through the element slot both in the longitudinal and in the transverse direction. It decreases the probability of coolant impurities settling in local sites as well as the probability that areas of overheating and corrosion of the cladding of the fuel element will form. Given that the cladding burns out and swells along the height of the fuel element in a non-uniform manner, the width of the slot can vary. It can be smaller in the middle part of the element than in the peripheral part. In order to assure an optimal combination of longitudinal rigidity of spacing elements and reduction of loads during their transverse deformation, the wall thickness of the element is set between 0.25 and 1 of the fuel element cladding thickness. Said value for the element wall thickness and its fabrication from the same material as the material of the fuel element cladding (for example, from stainless steel) helps to create optimal conditions for ensuring the strength and reliability of weld joints between end sections of the element and the cladding of the fuel elements or its end plugs. The element can be made from a thin-walled tube, in which a through cut is made to create a longitudinal slot of a set width, or else by bending a thin-walled strip. The width of the slot in the element can be created both by milling of the tube and by cutting and bending inside the tube of the cut edges without removal of metal. Elements with edges bent inside along the line of the cut increase the rigidity and stability of the element shape during the process of fabrication of the fuel element and also when fuel burns out in the core. Reduced resistance to deformation in the transverse plane for the spacing elements of the peripheral fuel elements is achieved by the relative increase of their slot width from 0 to 0.35 of the element diameter, with the slot width of elements of other fuel elements ranging from 0.1 to 0.25 of the element diameter. To additionally reduce resistance to transverse deformation, the wall thickness of peripheral fuel elements is set between 0.25 and 0.6 of the thickness of the fuel element claddings, and the wall thickness of other fuel elements is set between 0.4 and 1 of the thickness of the fuel element claddings. In each specific case, varying these parameters makes it possible to ensure substantial deformation of the spacing elements of peripheral fuel elements in their zones of contact with the jacket, and to form in these zones an oval cross section of an element of given dimensions directly when assembling a fuel assembly. In accordance with the method for production of a fuel assembly, peripheral fuel elements assume a substantially oval cross section in the regions where they are in contact with the jacket by transverse compression of fuel element bundles during production of a fuel assembly. Transverse compression of fuel elements fixed in a grid is carried out either directly, due to pressure from the jacket sides when the bundle is inserted in the jacket, or with the aid of several hexagonal compressing rims. Rims are put on a fuel element bundle in an open state, and then their parts are pulled together until fitting gaps between fuel elements and elements are created and until assumption of an oval cross section by peripheral fuel elements at their points of contact with the rims. The fuel assembly jacket is put on the assembled fuel element bundle beginning with its butt; next, the compressing rims are removed from the bundle surface one by one as the bundle is inserted inside the jacket. As a result, a bundle with a cross section of the given form and an oval cross section of the peripheral fuel elements in their zones of contact with the jacket is formed inside the jacket. Since the fuel element bundle is compressed within the limits of elastic deformation of the elements, the bundle can be installed in the fuel assembly jacket with a certain tension in the zones of contact between the fuel elements in the bundle and between the bundle and the jacket. Selection of element parameters within said limits makes it possible to select technological gaps at the first stage of compression of the bundle and to form an oval cross section of elements in their zones of contact with the jacket, and then to ensure the required spacing tension of the fuel element bundle in the jacket and the jacket itself. FIG. 1 and FIG. 2 provide illustrations of the formation of a fuel assembly and the method for its production in accordance with the invention. FIG. 1 shows a fragment of the cross section of a fuel assembly with a triangular grid for packing fuel elements into a bundle. FIG. 2 shows fragments of the cross section of a fuel assembly in the zone of contact of the peripheral fuel element with the jacket in different conditions of deformation of the element cross section. In accordance with the invention, the fuel assembly includes a top nozzle and a bottom nozzle (not shown in the illustrations), connected to one another with the aid of a jacket (1); and a bundle (2) of rod-type fuel elements arranged in a fuel assembly with the aid of a grid (not shown). The bundle consists of internal fuel elements (3) with spacing elements (4), along with peripheral fuel elements (5) with spacing elements (6), which are in contact with the jacket (1). The spacing elements (4) and (6) are provided with longitudinal through slots (7). The elements (6) of fuel elements (5) of the peripheral row in their zones of contact (8) with the jacket (1) have a substantially oval cross section (9) (FIG. 1 shows the oval shape (9) in the plane passing through the contacts (8)). The width of the slot (7) in the elements (4) and (6) is set between 0.1 and 0.35 of the element diameter. The wall thickness of elements (4) and (6) is set between 0.25 and 1 of the thickness of the fuel element cladding. The elements (6) of the peripheral fuel elements (5) have reduced resistance to deformation in the transverse direction as compared to the elements (4). For this purpose, the width of the slots (7) in the elements (6) is set between 0.20 and 0.35 of the element diameter, with corresponding values between 0.1 and 0.30 for the elements (4); the wall thickness of the peripheral fuel elements (5) is set between 0.25 and 0.6 of the fuel element cladding thickness, while the corresponding figures for the internal fuel elements (3) are between 0.4 and 1. Fuel assemblies are produced as follows. According to the known technology, a fuel element bundle with elements is formed by fastening their lower end parts in a grid (not shown in the figures). Transverse compression of the bundle (2) can be achieved, for example, through direct contact of the bundle with the internal sides of the jacket when it is inserted inside the jacket. Transverse compression of the bundle (2), consisting of fuel elements (3) and (5), can be also carried out with the aid of several two-piece hexagonal rims (10), which in contracted conditions have a cross section identical to the cross section of the fuel assembly jacket (1). The rims (10) are put on a bundle (2) in the open position; next, the halves of the rims are pulled towards one another and the bundle is compressed in the transverse plane until fitting gaps are created, and then until the deformation of elements (6) in their zones (8) of contact with the rims (11) with the oval cross section (9). Next, the free end of the contracted bundle (2) is inserted in the jacket (1) and the jacket is put on the bundle by removal of the rims (10) one by one as they are replaced by the jacket (1). As a result, a bundle (2) with the cross section of the given shape (for example, hexagonal) and an oval cross section (9) of the fuel elements (5) in their zone (8) of contact with the jacket (1) are formed in the jacket (1). FIG. 2a shows the initial condition of the cross section of the element (6) before the bundle is compressed, and FIG. 2b shows the oval cross section of the element (6) after the bundle (2) has been compressed by the rims (10); this cross section does not change even after insertion of the bundle (2) into the jacket (1). Compression of the bundle (2) in the rims (10) and the jacket (1) is carried out within the limits of elastic deformation of the elements (4) and (6), which ensures the controlled spacing tension between the fuel elements in the bundle, as well as between the peripheral fuel elements (5) and the jacket (1). In the process of irradiation of fuel in the core, radiation swelling of fuel and increase of the fuel element cladding diameter occur. In accordance with the invention, in a fuel assembly, these changes of dimension are compensated for in the transverse plane by substantial deformation of the elements (4) and (6). The shape of the cross section of the element (6) in conditions of additional deformation is shown in FIG. 2c), while the shape of the cross section of the element (4) changes as the slot (7) width decreases as far as the oval cross section shown in FIG. 2b). The reliability of peripheral fuel element spacing improves significantly, affecting nine points of contact in the invention (three in the known solution). This results in reduction of local stresses in the thin-walled fuel element cladding, as well as in a decrease in pitting corrosion processes in the zones of contact between the claddings and the elements. Said properties of the fuel assembly allow it to be used for planned lead-cooled fast neutron reactors and reactors with nitride uranium-plutonium fuel, in which the exterior diameter of the cladding can be more than 10 mm, and the minimum distance between the claddings of adjacent fuel elements can be more than 3 mm. The proposed design of the fuel assembly makes it possible to reduce the metal intensity of the assembly due to the fabrication of elements in the form of thin-walled tubes. A longitudinal through slot in the element allows for a significant reduction in the likelihood of its deformation in the transverse plane, and for a reduction in the local stresses occurring in the cladding of the fuel element when fuel in the reactor core burns out. Achievement of said technical results helps to improve the neutronic parameters of the fast neutron nuclear reactor core and improve the operational reliability of the fuel elements and fuel assemblies. The technical solution allows the rigidity of the spacing element in the longitudinal direction to be substantially reduced. This makes it possible to wrap the blanks of a spacing element with a specified tension around the surface of thin-walled cladding without noticeable disruption of the geometrical shape of the fuel element. Moreover, the proposed design makes the weld joints of the spacing element and cladding highly reliable due to its homogeneous composition, its structure and the geometrical shape of the welded pieces. Said properties of the technical solution allow for its practical implementation during fabrication of fuel assemblies for new-generation power fast neutron nuclear reactors. The new features of the design and method of fuel assembly arrangement, in conjunction with other features, allow simple and reliable means to be employed to ensure the required characteristics of the arrangement of fuel elements both inside a bundle and at its periphery. These characteristics include assurance of the specified level of local stresses in the zone of contact between the fuel element cladding and the spacing elements upon longitudinal and transverse changes in the dimensions of the fuel elements and the fuel assembly frame under the impact of temperature and radiation, and assurance of a minimum metal intensity of fuel assembly elements and simplicity of fabrication and arrangement of fuel assemblies. The proposed design can be used for fuel assemblies with various shapes of cross section, for example square or hexangular.
	abstract	A writing apparatus includes a writing unit configured to write a first pattern onto a first mask substrate and a second pattern being complementary to the first pattern onto a second mask substrate using a charged particle beam, and an addition unit configured to add a positional deviation amount of the first pattern having been written on the first mask substrate to a writing position of the second pattern, wherein the writing unit writes the second pattern at the writing position on the second mask substrate, where the positional deviation amount of the first pattern has been added.
	summary
	claims	1. A shipping container system for a first nuclear fuel product comprising: an elongated tubular container designed to receive and support the first nuclear fuel product therein, an exterior of the tubular container having at least two substantially flat walls which extend substantially over an axial dimension of the tubular container, with at least one circumferential end of at least one of the walls having a hinged interface to provide access to the interior thereof; an elongated, tubular overpack having an axial dimension at least as long as the tubular container, an internal cross-section larger than the tubular container and an interior tubular channel having an axially extending lower support section supporting a plurality of shock mounts, with at least one of said plurality of shock mounts positioned on either radial side of the lower support section, the shock mounts positioned on one radial side of the lower support section support one of the axially extending flat walls of the tubular container and the shock mounts positioned on the other radial side of the lower support section support another of the axially extending flat walls of the tubular container, in spaced relationship with the lower support section when the overpack is supported in a horizontal position, with at least one circumferential end of the lower support section having a clamped interface substantially along the axial dimension thereof to provide access to the interior of the overpack; and means for supporting the overpack in the horizontal position. 2. The shipping container system of claim 1 wherein the tubular container has a rectangular or hexagonal cross-section. claim 1 3. The shipping container system of claim 2 wherein at least two adjacent sides of the tubular container are hingably connected axially to abutting remaining sides of the tubular container and latchably connected axially at an abutting joint of the two adjacent sides so that when unlatched the two adjacent sides can swing open to reveal the inside of the tubular container. claim 2 4. The shipping container system of claim 1 wherein the elongated tubular container is constructed substantially out of aluminum. claim 1 5. The shipping container system of claim 1 for shipping a nuclear fuel assembly having a grid for supporting a plurality of fuel rods in an array, wherein the tubular container has an inside surface including an axially adjustable pad attachable to the inside surface at an axial elevation of a location of the grid to rest up against said grid when said fuel assembly is in said tubular container. claim 1 6. The shipping container system of claim 5 wherein a face of the pad resting against the grid is constructed from a material that is softer than the material the grid is constructed out of. claim 5 7. The shipping container system of claim 6 wherein the face of the pad is constructed from cork rubber. claim 6 8. The shipping container system of claim 3 wherein the latch is a kingpin of a hinge and the sides of the tubular container can be opened at anyone of a plurality of mating side interfaces at the hinged connections. claim 3 9. The shipping container system of claim 1 including neutron-absorbing material attached to one surface of the walls of the tubular container. claim 1 10. The shipping container system of claim 1 wherein the tubular overpack is constructed in at least two axially extending segments with one segment forming the lower support section. claim 1 11. The shipping container system of claim 10 wherein each segment is formed from a stainless steel hollow shell with the hollow area filled with a lightweight material. claim 10 12. The shipping container system of claim 11 wherein the lightweight material is Polyurethane. claim 11 13. The shipping container system of claim 11 including thermal insulation within the hollow area. claim 11 14. The shipping container system of claim 10 wherein the segments are keyed to align when fit together. claim 10 15. The shipping container system of claim 10 wherein the segments latch together. claim 10 16. The shipping container system of claim 1 wherein the shock absorbing mounts are axially spaced on either radial side of the lower support section and the shock absorbing mounts on one radial side of the lower support section are radially spaced from the shock absorbing mounts on the other radial side of the lower support section. claim 1 17. The shipping container system of claim 16 wherein the number of shock absorbing mounts is dependent on the weight of the first nuclear product. claim 16 18. The shipping container system of claim 1 wherein the shock absorbing mounts are rubber mounts. claim 1 19. The shipping container system of claim 1 wherein a wall of the interior tubular channel of the tubular overpack is lined with a neutron absorbing material. claim 1 20. The shipping container system of claim 1 wherein the at least two substantially flat walls of the elongated tubular container abut each other at a joint at an interior angle of less than 180 degrees to form xe2x80x9cv-shapexe2x80x9d and the xe2x80x9cv-shapexe2x80x9d is supported substantially at a point radially within the overpack that is lower than any other points on the exterior of the tubular container when the overpack is supported in the horizontal position. claim 1 21. A shipping container system for a first nuclear fuel assembly comprising: an elongated, tubular container having an axial dimension at least as long as the first nuclear fuel assembly, an internal cross-section larger than the first nuclear fuel assembly and an interior tubular channel that is sized to closely receive and is shaped to substantially conform to the peripheral shape of the first nuclear fuel assembly, the tubular container having at least two circumferentially spaced axial seams that when separated open a wall section of the tubular container to reveal the interior tubular channel and through which the first nuclear fuel assembly can be loaded into and unloaded from the tubular container, an exterior of the tubular container having at least two substantially flat walls which extend substantially over the axial dimension of the tubular container; an elongated, tubular overpack having an axial dimension at least as long as the tubular container, an internal cross-section larger than the tubular container and an interior tubular channel having an axially extending lower support section supporting a plurality of shock mounts, with at least one of said plurality of shock mounts positioned on either radial side of the lower support section, the shock mounts positioned on one radial side of the lower support section support one of the axially extending flat walls of the tubular container and the shock mounts positioned on the other radial side of the lower support section support another of the axially extending flat walls of the tubular container, in spaced relationship with the lower support section when the overpack is supported in a horizontal position; and means for supporting the overpack in the horizontal position so that the abutment joint of the flat walls is at the lowest point.
	abstract	A loop dissolution system specifically suited for dissolving uranium compounds in an acidic bath that continually circulates the acid over the uranium compound to be dissolved. The dissolution system includes an upper material feed dissolution plate on which the material to be dissolved is fed, a lower mixing and dissolution ring and a drop pipe system connecting and establishing fluid communication between the upper material feed dissolution plate and the lower mixing and dissolution ring. A pump for circulating the acidic fluid has an intake from the lower mixing and dissolution ring and an outlet that directs a first portion of a fluid to the upper material feed dissolution plate and a second portion of the fluid back into the lower mixing and dissolution ring to circulate the material suspended in the fluid within the lower mixing and dissolution ring to promote turbulence and facilitate dissolution.
	claims	1. A grid foil shadow removing method for a radiographic apparatus for obtaining fluoroscopic images and having a synchronous grid with grid foil strips arranged at regular intervals so that grid foil shadows fall on middles of pixels which detect radiation, the method comprising:an extracting step including a grouping step for dividing pixels forming a fluoroscopic image into groups each having a predetermined number of pixels within each row in a direction of row, a most influenced pixel selecting step for selecting a pixel most influenced by one of the grid foil shadows in each group as most influenced pixel, a voting step for casting, with the most influenced pixel in each group serving as a reference, a predetermined number of votes for other pixels spaced apart forward and backward in the direction of row, and an electing step for electing a pixel given a maximum number of votes within each group as an uninfluenced pixel which is free from influences of a foil shadow of the grid;an approximate fluoroscopic image calculating step for obtaining an approximate fluoroscopic image by carrying out an interpolation process based on detection signal values of the uninfluenced pixels;a grid foil shadow image calculating step for obtaining a grid foil shadow image based on a difference between the fluoroscopic image and the approximate fluoroscopic image;a foil shadow standard image calculating step for obtaining a foil shadow standard image by averaging the grid foil shadow image in a longitudinal direction of the grid foil shadows; anda foil shadow removing step for removing the grid foil shadows from the fluoroscopic image based on a difference between the fluoroscopic image and the foil shadow standard image, thereby to obtain a foil shadow removed image. 2. The method according to claim 1, wherein:when the predetermined number of pixels constituting each group is four, and the predetermined number of votes is 1;the voting step casts 1 vote for each of a pixel located next but one forward and a pixel located next but one backward in the direction of row; andthe electing step elects a pixel having obtained 2 votes as the uninfluenced pixel;the voting step and the electing step having, interposed therebetween adjusting steps including:a first adjusting step for adjusting the number of votes obtained to 2 for a pixel whose number of votes obtained is 1 when pixels located next but three to such pixel forward and backward in the direction of row have 2 votes, respectively;a second adjusting step for adjusting the number of votes obtained to 0 for a pixel whose number of votes obtained is 1 when a pixel located next to such pixel in the direction of row has 2 votes;a third adjusting step for comparing detection signal values of a pixel whose number of votes obtained is 1 and an adjoining pixel, adjusting the number of votes from 1 to 2 for the pixel having the larger detection signal value, and adjusting the number of votes to 0 for the pixel having the smaller detection signal value; anda fourth adjusting step for adjusting the number of votes obtained from 1 to 0 for a pixel whose number of votes obtained is 1 when one of pixels located next but one to such pixel forward and backward in the direction of row has 2 votes. 3. The method according to claim 2, further comprising a fifth adjusting step executed, when there remains a pixel whose number of votes obtained is 1 after the fourth adjusting step, for adjusting the number of votes obtained by such pixel to 2. 4. The method according to claim 2, comprising a forcible changing step executed after the extracting step, when a predetermined range includes an uninfluenced pixel skipping four pixels, and an uninfluenced pixel skipping two pixels, for forcibly changing the uninfluenced pixels so that each have three pixels at both sides. 5. The method according to claim 3, comprising a forcible changing step executed after the extracting step, when a predetermined range includes an uninfluenced pixel skipping four pixels, and an uninfluenced pixel skipping two pixels, for forcibly changing the uninfluenced pixels so that each have three pixels at both sides. 6. The method according to claim 1, wherein the most influenced pixel selecting step is executed for selecting a pixel having a minimum detection signal value as the most influenced pixel. 7. The method according to claim 2, wherein the most influenced pixel selecting step is executed for selecting a pixel having a minimum detection signal value as the most influenced pixel. 8. The method according to claim 3, wherein the most influenced pixel selecting step is executed for selecting a pixel having a minimum detection signal value as the most influenced pixel. 9. The method according to claim 4, wherein the most influenced pixel selecting step is executed for selecting a pixel having a minimum detection signal value as the most influenced pixel. 10. The method according to claim 5, wherein the most influenced pixel selecting step is executed for selecting a pixel having a minimum detection signal value as the most influenced pixel. 11. A radiographic apparatus for obtaining radiographs, comprising:a radiation emitting device for emitting radiation to a patient;a radiation detector having pixels arranged in a two-dimensional array for detecting radiation transmitted through the patient;a synchronous grid with foil strips arranged at regular intervals so that grid foil shadows fall on middles of the pixels of the radiation detector;an extracting unit including a grouping unit for dividing pixels forming a fluoroscopic image into groups each having a predetermined number of pixels within each row in a direction of row, a most influenced pixel selecting unit for selecting a pixel most influenced by one of the grid foil shadows in each group as most influenced pixel, a voting unit for casting, with the most influenced pixel in each group serving as a reference, a predetermined number of votes for other pixels spaced apart forward and backward in the direction of row, and an electing unit for electing a pixel given a maximum number of votes within each group as an uninfluenced pixel which is free from influences of a foil shadow of the grid;an approximate fluoroscopic image calculating unit for obtaining an approximate fluoroscopic image by carrying out an interpolation process based on detection signal values of the uninfluenced pixels;a grid foil shadow image calculating unit for obtaining a grid foil shadow image based on a difference between the fluoroscopic image and the approximate fluoroscopic image;a foil shadow standard image calculating unit for obtaining a foil shadow standard image by averaging the grid foil shadow image in a longitudinal direction of the grid foil shadows; anda foil shadow removing unit for removing the grid foil shadows from the fluoroscopic image based on a difference between the fluoroscopic image and the foil shadow standard image, thereby to obtain a foil shadow removed image. 12. The apparatus according to claim 11, wherein:when the predetermined number of pixels constituting each group is four, and the predetermined number of votes is 1;the voting unit casts 1 vote for each of a pixel located next but one forward and a pixel located next but one backward in the direction of row; andthe electing unit elects a pixel having obtained 2 votes as the uninfluenced pixel;the voting unit and the electing unit having, interposed therebetween, an adjusting unit including:a first adjusting unit for adjusting the number of votes obtained from 1 to 2 for a pixel whose number of votes obtained is 1 when pixels located next but three to such pixel forward and backward in the direction of row have 2 votes, respectively;a second adjusting unit for adjusting the number of votes obtained to 0 for a pixel whose number of votes obtained is 1 when a pixel located next to such pixel in the direction of row has 2 votes;a third adjusting unit for comparing detection signal values of a pixel whose number of votes obtained is 1 and an adjoining pixel, adjusting the number of votes from 1 to 2 for the pixel having the larger detection signal value, and adjusting the number of votes from 1 to 0 for the pixel having the smaller detection signal value; anda fourth adjusting unit for adjusting the number of votes obtained from 1 to 0 for a pixel whose number of votes obtained is 1 when one of pixels located next but one to such pixel forward and backward in the direction of row has 2 votes. 13. The apparatus according to claim 12, wherein the adjusting unit further includes a fifth adjusting unit arranged, when there remains a pixel whose number of votes obtained is 1 after the adjustment by the fourth adjusting unit, to adjust the number of votes obtained by such pixel to 2. 14. The apparatus according to claim 12, further comprising a forcible changing unit arranged, after the process by the extracting unit, when a predetermined range includes an uninfluenced pixel skipping four pixels, and an uninfluenced pixel skipping two pixels, to forcibly change the uninfluenced pixels so that each have three pixels at both sides. 15. The apparatus according to claim 13, comprising a forcible changing unit arranged, after the process by the extracting unit, when a predetermined range includes an uninfluenced pixel skipping four pixels, and an uninfluenced pixel skipping two pixels, to forcibly change the uninfluenced pixels so that each have three pixels at both sides. 16. The apparatus according to claim 11, wherein the most influenced pixel selecting unit is arranged to select a pixel having a minimum detection signal value as the most influenced pixel. 17. The apparatus according to claim 12, wherein the most influenced pixel selecting unit is arranged to select a pixel having a minimum detection signal value as the most influenced pixel. 18. The apparatus according to claim 13, wherein the most influenced pixel selecting unit is arranged to select a pixel having a minimum detection signal value as the most influenced pixel. 19. The apparatus according to claim 14, wherein the most influenced pixel selecting unit is arranged to select a pixel having a minimum detection signal value as the most influenced pixel. 20. The apparatus according to claim 15, wherein the most influenced pixel selecting unit is arranged to select a pixel having a minimum detection signal value as the most influenced pixel.
	abstract	Beads coded with phosphor particles and methods of making and using them are provided.
	abstract	A medical imaging system for detecting ionizing radiation. The system includes one or more pixilated imagers positioned to acquire patient image data and one or more position sensors positioned to acquire patient position data. Once the patient image data and patient position data are acquired, one or more processors operably connected to each of the one or more pixilated imagers and one or more position sensors calculate a three-dimensional mass distribution based on patient image data and patient position data.
052456412	description	DESCRIPTION OF THE PREFERRED EMBODIMENT Referring now to the drawings wherein like reference characters designate like or corresponding parts throughout the several views, there is shown in FIGS. 1-3 a fuel rack of modular design particularly adapted to receive fuel assemblies of the type used in boiling water reactors. The module includes a 0.50 inch thick base plate 14 designed to support multiple stainless steel containers or cells 16 each designed to hold a fuel assembly. Stability and rigidity is imparted to the modular unit by a base assembly and welds at the corners of each cell connecting all cells in a unitary structure. Adjustable leveling pads 26 located at the corners of the module and intermediate leveling pads 28 (FIGS. 2, 3) located at appropriate points beneath the base plate assures desired rigidity. Squareness and verticality in the module are also assured by leveling the base plate to a horizontal position. As more clearly shown in FIGS. 3, 7 and 10, the base plate 14 includes multiple openings 30 of a design and size compatible with the configuration of the bottom nozzle of a fuel assembly adapted to be placed in each cell. In addition to receiving and supporting the bottom end of a fuel assembly, openings 30 provide a natural circulation flow path upwardly to assure proper cooling of the stored fuel assemblies by water or other coolant medium. Openings 30 further provide access to the leveling pads 26 while the openings in the central part of plate 14 provide access to the interior leveling pads 28. As more fully described hereafter, other openings have cut-away sections 32 which are particularly designed to receive a lifting tool for lifting and transferring the plate and/or the module to different locations. As illustrated in FIGS. 4-6, the spent fuel rack module is substantially different from prior art designs inasmuch as no space exists between adjacent cells. The cells 16 shown are manufactured from a number of L-shaped sections 34 of a height sufficient to extend higher than the height of a fuel assembly and of a length and width just slightly in excess of those dimensions of a fuel assembly. To import strength and stability to the cells, the cell walls are made of 0.075 inch stainless steel although cell walls of different thicknesses may be used depending on the design criteria for the particular application. Neutron absorbing material 36, more fully described hereafter, is attached along the length and on the outside cell wall surfaces to help preclude the transfer of neutrons from one fuel assembly to another located in adjacent cells. All of the cells in the fuel rack module except those on the fuel rack periphery have neutron absorbing material on all four sides. Each inner cell is made up of two L-shaped sections 34, and to provide simplicity in the manufacturing operations, neutron absorbing material is attached to the outer side of the walls of both L-shaped sections. To form a complete cell, one of the two thusly formed manufacturing sections 34 is placed in contact with, or in substantial contact with the corresponding longitudinal edges on the other sections. When these abutting edges of the two L-shaped sections are joined together by welds 42 (FIG. 5), a complete cell is formed of a size sufficient to accept a fuel assembly. The cell is enclosed by neutron absorbing material 36 on all four walls (FIG. 1) and the material on these walls serves the same function in absorbing neutrons from fuel assemblies in the adjacent cells. When the cells of a complete module are constructed in this manner, examination of FIG. 4 will show that every alternate cell is formed by the L-shaped sections which surround it. Note for example that the cell designated 44 is formed by walls 46, 48, 50, and 52. Each of these walls also comprise one wall of each cell which surround the cell 44. Cells which are thusly formed by the L-shaped walls are designated C and those cells which share the walls of the C cell are designated O. The neutron absorbing material 36 prefereably comprises Boraflex which is an elastomeric silicone polymer matrix manufactured by Brand Industrial Services, Inc. of Parkridge, Ill. Other neutron absorbing materials may be used if desired. The Boraflex is approximately 0.045 inch thick and extends substantially the full length and width of the side wall on which it is mounted. A wrapper plate 54 (FIGS. 7, 10) of 0.035 inch stainless steel protects the Boraflex against physical damage and is welded along the edges to the L section wall surfaces. The wrapper plate 54 may terminate short of the end of each L-shaped section as shown in FIG. 10, or extend to the complete end. An inspection hole 55, FIG. 10, is used to visually verify proper placement of the neutron absorber material. Also, it will be noted that each alternate cell on the module periphery is closed by a panel 58, FIG. 4, which extends the cell complete length. To form a spent fuel rack, all the L-shaped sections are assembled into cells as described above, and with this construction, the interior of each cell, other than those cells located on the module periphery, is bounded by walls having neutron absorbing material located either on the inside or the outside walls of a particular cell. Since the corner cells of the module do not require neutron absorbing material on its outside walls, the L-shaped sections 56 located at the corners comprise only stainless steel plate having their longitudinal edges welded to adjacent L-shaped sections on the module. When thus welded, each corner cell is then of a size and configuration to accept a fuel assembly. FIGS. 4-6 illustrate how the L-shaped sections are welded together to form cells. Variations in the characteristics of the stainless steel material after being exposed to varying temperatures and stresses during the manufacturing process, produces slight distortion in the material such that the L section longitudinal edges do not always fall in a vertical plane. Usually the longitudinal edges can be welded along their complete length as shown in FIGS. 5 and 6 wherein weld 42 metallurgically joins adjacent L-shaped sections 34. In the event the adjacent L sections of adjacent cells to be joined are spaced a distance greater than that which can be bridged by a single weld, such as weld 63, 0.81 inch spacer wire 68 shown in FIG. 5, is welded in the gap formed at the intersection of the L sections of adjacent cells. Separate welds 63 are then made between each section and the wire spacer and along their complete height to provide stability to the fuel rack module. Since the space between adjacent sections may vary, it is evident that spacer wire, or other appropriate filler material, of different diameters or strips of different thicknesses and lengths may be welded to the adjacent L sections to hold the parts together. The base plate 14 illustrated in FIGS. 3, 7 and 10 includes openings 30 which receive the bottom nozzle 64 (FIG. 10) of a fuel assembly adapted to be placed in the opening 30 provided in each cell 16. As shown in FIG. 10, the plate opening 30 is of a size larger than the bottom nozzle 64 but smaller than an upper portion 66 thereof. The wall of opening 30 includes a bevel 64 complementary to the sides of the nozzle 64 which provides a surface area which supports a fuel assembly when positioned on the rack. The openings 30 are further modified, FIGS. 3, 8 and 10, to permit access to lifting apparatus for transferring the base plate 14 alone, or the base plate plus the stainless steel cells, from one location to another. To accomplish this, each of four openings 30 is provided with oppositely directed slots 32 (FIG. 7) of a size sufficient to accept lifting apparatus (not shown). This apparatus includes a base plate engaging mechanism which is of a length just sufficient to fit through slot 32 prior to being rotated 90.degree. to a position for engagement with the underside of plate 14 for lifting purposes. To impart strength to the base plate 14 in the areas where the engaging mechanism of the lifting apparatus will engage the plate, a one-inch stainless steel block 70, FIGS. 8 and 9, having a lifting tool opening 73 of the same width as slot 32 in the base plate 14, is welded to the underside of the base plate and at the four positions indicated in FIG. 3. Depending on the size of base plate and module being lifted, either a greater or lesser number of openings may be provided for lifting purposes. The stainless steel block 70 includes in the underside a tool retaining slot 71. The slot is located in the manner shown to provide an impediment to rotation of the lifting apparatus after it is inserted through the slots 32 and rotated 90.degree. to a plate lifting position. When the base plate is lifted, the retaining slot will prevent rotation of the lifting apparatus in the event it is jarred or vibrated during transfer of the base plate from one position to another. The openings 30 with the slots 32 and the blocks 70 constitute receiving and locking means for lifting apparatus which is wholly within the periphery of base 14. The leveling apparatus beneath base plate 14 is similar to the design set forth in U.S. Pat. No. 4,820,472 inasmuch as the leveling pads are designed to conform to the slope of the pool floor and still achieve true squareness in the construction relative to vertical and horizontal axes. The leveling apparatus is also described in the parent of this application U.S. Pat. No. 4,900,505. As illustrated in FIGS. 7 and 10, each corner leveling pad 26 is mounted on the pool floor 100 (FIGS. 1, 2) for leveling base plate 14. It will be apparent that many modifications and variations are possible in light of the above teachings. It therefore is to be understood that within the scope of the appended claims, the invention may be practiced other than as specifically described.
	claims	1. A method of producing radionuclides, the method comprisingcausing a liquid capture matrix to continuously flow into a mobile phase filter compartment inside an irradiation core and then into a collection reservoir outside the radiation core while simultaneously irradiating a target comprising a target nuclide trapped in the mobile phase filter compartment by a mobile phase filter,wherein the target is irradiated with radiation, ionizing radiation, particles, or a combination thereof to produce radionuclides that are ejected from the target and into the liquid capture matrix such that the radionuclides are collected in the collection reservoir. 2. The method of claim 1, wherein the radionuclides are produced from the target nuclide via a reaction selected from the group consisting of a (n, γ) reaction, a (γ, n) reaction, and a (n, 2n) reaction. 3. The method of claim 1, wherein the radiation is selected from the group consisting of neutron radiation, gamma radiation, and a combination thereof. 4. The method of claim 1, wherein the radiation comprises thermal neutron radiation, epithermal neutron radiation, or a neutron radiation having a neutron energy above 0.4 eV. 5. The method of claim 1, wherein the liquid capture matrix is water. 6. The method claim 1, wherein the liquid capture matrix has a pH of about 3 to 5. 7. The method of claim 1, wherein the target nuclide has one or more organic ligands attached thereto, and the radionuclides are ejected from the target by breaking one or more bonds with the ligands. 8. The method of claim 1, wherein the target nuclide is selected from the group consisting of elements having atomic number from 21 to 102. 9. The method of claim 1, wherein the target nuclide is selected from the group consisting of 23Na, 31P, 37Cl, 50Cr, 55Mn, 75As, 81Br, 89Y, 98Mo, 104Ru, 127I, 152Sm, 165Ho, 174Yb, 175/176Lu, 185Re, 187Re, 194Pt, 197Au and 237Np. 10. The method claim 1, wherein the target nuclide is selected from the group consisting of the lanthanides and the actinides. 11. The method of claim 1, wherein the target comprises one or more organic ligands selected from the group consisting of acetylacetonate, picolinate, 8-hydroxyquinolinate, dimethylglyoximate, oxalate, 4-aminobenzoate, glycinate, and derivatives thereof. 12. The method of claim 1, wherein the target comprises a support structure, and wherein the target nuclide is in a polymer coating on the support structure. 13. The method of claim 12, wherein the support structure is a mesoporous resin. 14. The method of claim 12, wherein the support structure is a copolymer of styrene and divinylbenzene. 15. The method of claim 12, wherein the polymer coating on the support structure has a thickness that is less than a recoil range of the target nuclide. 16. The method of claim 12, wherein the target nuclide is 237Np, andwherein the radionuclide is 238Np that decays to produce 238Pu. 17. The method of claim 12, wherein the target nuclide is 96Mo, and wherein the radionuclide is 99Mo that decays to produce 99mTc. 18. The method of claim 1, wherein the target has a solubility in the capture matrix of about 0.001 M or less at room temperature. 19. The method of claim 1, wherein the enrichment factor is from 3 to 30.
	claims	1. A configuration for obtaining quantitative x-ray images from a sample, the configuration comprising:an X-ray source outputting x-rays;a position-sensitive detector with spatially modulated detection sensitivity having a number of individual pixels;a recorder for recording the x-ray images of said position-sensitive detector;a processor for evaluating intensities in a single shot image for obtaining characteristics of the sample including absorption, differential phase contrast and directional, small-angle scattering contrast, for specified regions of the pixels;a phase-shift periodic structure being:i) a 2D periodic structure composed of unit cells, said unit cells being circular gratings, a period of said unit cells is P and that of said circular gratings is p, wherein periodic structures in a circular grating generate a considerable X-ray phase shift difference, which is of π/2 or odd multiples thereof, hereinafter referred to as π/2 shift; or π or π+N×2×π, hereinafter referred to as π shift, where N is an integer number; orii) a honeycomb structure composed of unit cells, each unit cell is a circular grating which allows a maximum filling factor of such a periodic structure, periodic structures in said circular grating generate a considerable X-ray phase shift difference, which is of π/2 or odd multiples thereof, hereinafter referred to as π/2 shift; or π or π+N×2×π, hereinafter referred to as π shift, where N is an integer number;said phase-shift periodic structure creates a periodic interference pattern with a repetition of each said unit cell being P′ and a period within each said unit cell is p′ at a known distance, Talbot effect, downstream on said position-sensitive detector; where P′ and p′ match a radius of curvature of an incident wavefront by relation p ′ = 1 η ⁢ p ⁢ d 1 + l 1 l 1 , P ′ = 1 η ⁢ P ⁢ d 1 + l 1 l 1 where:l1 is a distance between said X-ray source and said phase-shift periodic structure; andd1 is a distance between said phase-shift periodic structure and a created self-image η=1 for π/2 shift grating while η=2 for π shift grating. 2. The configuration according to claim 1, wherein said phase-shift periodic structure is made by:deep etching into silicon, a polymer or similar material, for low energy X-ray photons; ordepositing a heavy metal into gaps of a low-absorbing structure or grow the heavy metal on a low-absorbing substrate, in either case the heavy metal is used as a phase shift material for high energy X-ray photons. 3. The configuration according to claim 1, further comprising an absorption grating disposed in front of, or embedded into said X-ray source:wherein said l1 is a distance between said absorption grating and said phase-shift periodic structure; andwherein said absorption grating is a 2D chessboard/mesh-type grating with a pitch of p 0 = p × l 1 d 1 or integer multiples thereof. 4. The configuration according to claim 1, further comprising a mechanism being comprised to place the sample to be investigated between said X-ray source and said phase-shift periodic structure, or between said phase-shift periodic structure and said position-sensitive detector. 5. The configuration according to claim 1, wherein an analysis procedure is implemented for obtaining the absorption, the differential phase contrast and directional scattering contrasts of the sample that comprises the steps of recording two intensity images of an interference pattern, one sample image with the sample and one flat image without the sample on said position-sensitive detector. 6. The configuration according to claim 1, wherein said processor detects a location of individual ones of said unit cells on a recorded flat image by using a circular nature of said phase-shift periodic structure, that being said, an intensity maximum is observed in a center of said unit cell. 7. The configuration according to claim 6, wherein said processor calculates a shift between the recorded flat image and a sample image of each said unit cell. 8. The configuration according to claim 1, wherein said processor evaluates a radial visibility reduction for every angle in order to obtain omnidirectional scattering images. 9. The configuration according to claim 1, further comprising means for rotating the sample relatively to remaining components of the configuration to perform data collection for a tomographic scan. 10. The configuration according to claim 1, wherein said phase-shift periodic structure is an absorption grating. 11. The configuration according to claim 1, whereinthe x-rays are hard x-rays; andsaid X-ray source is a standard polychromatic X-ray source. 12. The configuration according to claim 6, wherein said processor calculates a shift between the recorded flat image and a sample image of each said unit cell, achieved either with Fourier based methods and/or Hilbert transform methods by calculating an analytical signal or spatial correlation methods. 13. The configuration according to claim 1, wherein said processor evaluates a radial visibility reduction for every angle in order to obtain omnidirectional scattering images accomplished by Fourier methods from the following formula ( n , m , O ) = R k s ⁢ R 0 f R k f ⁢ R 0 f . 14. The configuration according to claim 1, wherein said X-ray source has a 2D array of individual sources that may be mutually incoherent and whose lateral separation is p 0 = p × l 1 d 1 or integer multiples thereof. 15. The configuration according to claim 3, further comprising a mechanism being comprised to place the sample to be investigated between said absorption grating and said phase-shift periodic structure, or between said phase-shift periodic structure and said position-sensitive detector.

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