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042253888	summary	The invention relates to a crack detection cell for checking the leakproofness of fuel element units in the deactivation swimming-pool of a nuclear reactor, and the corresponding process of checking. When the core of a nuclear reactor is recharged with fuel element units which have been in the reactor for a certain period, it is necessary to check the leakproofness of these units in order to avoid introducing into the reactor units suffering from leakages due to cracks in the sheathing material of the fuel elements. Equally, if used units are transferred to the retreatment factory, it is necessary to know the level of .gamma.-activity released by these units so as to take the necessary precautions to avoid contamination. It is thus necessary to carry out a rapid detection of crack faults, on irradiated fuel units, on site at the time of discharging, and at the time of charging, the core of the reactor. One of the processes most used to carry out this rapid detection is a method of crack detection which consists of measuring the activity of fluids which have been in contact with the fuel unit during a rise in temperature of the fuel unit. During a rise in temperature of the fuel unit, the fission gases contained in the fuel elements expand and their speed of release through cracks in the sheaths increases so that volatile or water-soluble fission products increasingly contaminate the medium surrounding the unit. In this process, crack detection cells which are generally located in the deactivation swimming pool of the reactor are used. The fuel unit is located in a pocket, containing water, inside the crack detection cell. When the temperature of the fuel unit rises, the fission products are released into the water in which the fuel unit is immersed. Taking samples and measuring the .gamma.-activity of this water thus makes it possible to check the leakproofness of the fuel units. In order to achieve a rise in temperature of the fuel unit, a heat barrier is generally created between the pocket in which the fuel unit is located, and the water of the deactivation swimming pool, which is then no longer able to absorb the heat evolved by the fuel unit. To produce this heat barrier, the pocket intended to receive the fuel unit is generally located inside a double-walled chamber which rests vertically on the bottom of the deactivation swimming pool, and which has, in its external wall, orifices by which the gap between the walls of the chamber is connected with the swimming pool. This gap between the two walls of the chamber is connected to a feed line for inert gas under pressure from which gas under pressure passes into the gap and forces out the water of the deactivation swimming pool, to create the heat barrier. This system is particularly suitable in the case of fuel units which have to be checked shortly after discharge from the reactor core, the latter having functioned for a normal period of time, that is to say it is suitable for units which have a high residual power. In contrast, in the case of units of low residual power, that is to say fuel units which are checked long after discharge from the reactor, or following premature stoppage of the reactor, the rise in temperature of the unit becomes very slow so that the utilisation times increase considerably, which can make the system completely inapplicable. The above described process is in effect based on a comparison of the measurements of the .gamma.-activity in the water surrounding the fuel unit before and after the rise in temperature of the unit. The difference in activity is thus a function of the extent of the temperature rise of the unit. In order to treat a larger number of fuel units within a given time, it has been proposed to use a large number of crack detection cells simultaneously, but it is then necessary to provide the corresponding space in the deactivation swimming pool. It has also been proposed to empty the water from the pocket so as to allow a rapid rise in temperature of the fuel unit. However, this technique is difficult to master and suffers from certain dangers. On the other hand, in all the known processes the activity is measured on samples of fluids treated manually, which further increases the duration of use. According to one aspect of the invention there is provided a crack detection cell for checking the leakproofness of a fuel element unit in the deactivation swimming pool of a nuclear reactor, by measuring the .gamma.-activity of fluids in contact with the fuel unit, said cell comprising: a double-walled chamber for resting vertically on the bottom of the deactivation swimming pool, the external wall of said chamber defining orifices which connect the gap between said walls of said chamber with the swimming pool; means for connecting said gap to a source of gas under pressure for creating a heat barrier between the interior of said chamber and the deactivation swimming pool; a leakproof pocket for receiving a fuel unit in a vertical position, said pocket being located inside said chamber and comprising an upper part for serving as a zone for introducing said unit and a leakproof cover for closing said pocket; means for introducing water into, and for removing water from, said pocket; and heating means located at the lower part of said pocket and forming part of the internal surface of said pocket and for heating water contained in said pocket by direct contact to cause a rise in temperature of a fuel element when in said pocket. Preferably, the crack detection cell according to the invention includes means for circulating an inert gas in the pocket, and means, included in said circulation means, for continuously measuring the .gamma.-activity of the gas which has passed through the water contained in said pocket, in which the fuel element is immersed. According to another aspect of the invention there is provided a process for checking the leakproofness of a fuel element unit, using a crack detection cell as described above, comprising: causing circulation of inert gas in said pocket containing water in which said fuel unit is immersed, so as to entrain any fission products released by said fuel unit; and continuously measuring the activity of said circulating gas; thermally isolating said pocket containing said unit and energising said heating plates surrounding said pocket for a part of the time for which said gas is circulated; and checking the leakproofness of said unit by comparing the measurements of the activity of said circulating gas when said unit is not thermally isolated and heated and when said unit is both thermally isolated and heated.
	summary
	claims	1. Multi-leaf collimator with leaf drives, with two sets of displaceable leaves arranged side by side of each other and facing each other in order to impress a high-energy beam with the shape of an irregularly formed treatment object by enabling each of the leaves to assume a position oriented along the shape of the treatment object by means of the leaf drives, with the leaf drives being designed in such a way that the leaves are each equipped with a gear rod-like drive engagement in a direction of displacement,wherein a pivotable leaf-side gear segment located, together with a motor-side gear segment on a segment disk, engages with the gear rod-like drive engagement, with a pinion drivable by a motor engaging with the motor-side gear segment, wherein the segment disks are arranged side by side for each set of leaves as a package on one axle, and wherein the motor-side gear segments of two segment disks located next to each other are staggered in such a way that they will not abut each other. 2. Multi-leaf collimator in accordance with claim 1,characterized inthat the pinions are wider than the motor-side gear segments. 3. Multi-leaf collimator in accordance with claim 1,characterized inthat the motors for each package of segment discs in an engagement range of the respective motor-side gear segments are arranged in series in the shape of an arch. 4. Multi-leaf collimator in accordance with claim 3,characterized inthat the motors are mounted on a bearing block which encompasses in each case a package of segment discs in their circumferential range. 5. Multi-leaf collimator in accordance with claim 4,characterized inthat the bearing block is equipped on both sides with motors. 6. Multi-leaf collimator in accordance with claim 4,characterized inthat the bearing block positions the pinions indirectly or directly by means of positioning agents in their engagement position opposite the motor-side gear segments. 7. Multi-leaf collimator in accordance with claim 6,characterized inthat the pinions are mounted on axles supported by motor bearings and the latter are mounted on the bearing block. 8. Multi-leaf collimator in accordance with claim 7,characterized inthat the motor bearings each comprise a motor holder for mounting the motor and an axle bearing for bearing the axle. 9. Multi-leaf collimator in accordance with claim 8,characterized inthat the motor holders are made of aluminum and/or titanium and the axle bearings are made of bronze and/or brass. 10. Multi-leaf collimator in accordance with claim 3,characterized inthat a step-like gradation of an arrangement of the pinions driven by the motors is provided for their engagement with the various motor-side gear segments. 11. Multi-leaf collimator in accordance with claim 10,characterized inthat at least two step-like gradations are provided, with segment discs located next to each other being driven by motors with pinions assigned to various ones of these step-like gradations and the motor-side gear segments of segment discs lying next to each other being located in different areas of the circumference of the package of segment discs. 12. Multi-leaf collimator in accordance with claim 3,characterized inthat due to an arrangement of motor-side gear segments on corresponding varying radii of the segment discs, the motors are arranged in arch-shaped sequences lying on top of each other. 13. Multi-leaf collimator in accordance with claim 1,characterized inthat, relative to its width, the gear rod-like drive engagement of the leaves is designed differently from the width of the leaf-side gear segment. 14. Multi-leaf collimator in accordance with claim 1,characterized inthat the leaf drives are mounted adjustably such that a position of the leaf drives relative to the leaves may be adjusted. 15. Multi-leaf collimator in accordance with claim 14,characterized inthat the position of the leaf drives is adjustable by at least one excenter. 16. Multi-leaf collimator in accordance with claim 1,characterized inthat spacers are provided between adjoining segment discs that reduce mutual friction to the largest extent. 17. Multi-leaf collimator in accordance with claim 1,characterized inthat the leaves have a trapezoid cross section to the effect that they taper in the direction of a radiation source corresponding approximately to a divergence of the high-energy beam. 18. Multi-leaf collimator in accordance with claim 17,characterized inthat the sets of leaves are tilted relative to an optical path to the effect that no rays can pass through a gap between the leaves. 19. Multi-leaf collimator in accordance with claim 1,characterized inthat the leaf drives are designed in such a way that the leaves of the two sets of leaves can come in contact with each other with their front faces only outside of a center plane of the multi-leaf collimator. 20. Multi-leaf collimator in accordance with claim 1,characterized inthat the pinions and/or the motors are mounted adjustably such that a relative position of the pinions with regard to the motor-side gear segments may be adjusted. 21. A multi-leaf collimator (MLC) for controlling a shape of a high-energy radiation beam emanating from a radiation source and propagating in a direction of propagation, comprising:a plurality of leaves individually displaceable in a direction of displacement that is generally transverse to the direction of propagation, said plurality of leaves having a predefined range of displacement in said direction of displacement, each said leaf including a rack gear extending along the direction of displacement;a plurality of individually rotatable segment disks positioned side by side along a common axis of rotation that is generally transverse to said direction of propagation and to said direction of displacement, each said segment disk corresponding to a respective one of said leaves, each said segment disk including a leaf-side gear segment formed along a first peripheral portion thereof that is engaged with said rack gear of the corresponding leaf to displace that leaf along said direction of displacement according to a motor-controlled rotation of said segment disk around said common axis of rotation; anda plurality of motor-driven pinions, each said motor-driven pinion being engaged with a respective one of said segment disks along a motor-side gear segment formed along a second peripheral portion thereof to provide said motor-controlled rotation thereof;wherein the motor-side gear segments of any two adjacent segment disks are staggered in such a way that they will not abut each other throughout the range of displacement of their corresponding leaves. 22. The MLC of claim 21, each said motor-driven pinion being coupled to a distinct electrical motor to form a respective plurality of motor-pinion assemblies, wherein said plurality of motor-pinion assemblies are arranged in an arch-like pattern relative to said common axis of rotation of said plurality of segment disks. 23. The MLC of claim 22, wherein said motor-pinion assemblies are mounted on a common bearing block extending peripherally around said plurality of segment disks in an arch-like shape relative to said common axis of rotation, said motor-pinion assemblies being mounted on respective step-like gradations formed in said bearing block along the direction of said common axis of rotation for achieving respective engagement of said motor-driven pinions with said motor-side gear segments of said segment disks. 24. The MLC of claim 21, wherein said motor-driven pinions are wider than their associated motor-side gear segments in a direction of said common axis of rotation. 25. The MLC of claim 24, further comprising a spacer agent disposed between each adjacent pair of said segment disks for reducing mutual friction therebetween. 26. The MLC of claim 21, said plurality of leaves, said plurality of segment disks, and said plurality of motor-driven pinions collectively forming a first leaf/drive assembly, wherein the MLC further comprises a second leaf/drive assembly generally similar to said first leaf-drive assembly and disposed on an opposing side of a center plane of the MLC. 27. The MLC of claim 21, said plurality of leaves collectively having a radiation source-facing side and a patient-facing side opposite said radiation source-facing side, wherein said plurality of segment disks are disposed on said radiation source-facing side of said plurality of leaves, and wherein each of said plurality of segment disks has a radius along said first and second peripheral portions thereof that is sufficiently comparable to said predefined range of displacement of said leaves such that each said leaf can be fully displaced through its range of displacement in less than one full turn of said segment disk, whereby structural compactness of the MLC is facilitated. 28. The MLC of claim 27, said plurality of leaves in conjunction with said predefined range of displacement defining an overall lateral range (L) in said direction of displacement, wherein said plurality of segment disks and said plurality of motor-pinion assemblies are configured and dimensioned to be entirely confined within said overall lateral range on said radiation source-facing side of said plurality of leaves. 29. The MLC of claim 28, wherein said segment disk radius is greater than one-half of said predefined range of displacement of said leaves. 30. The MLC of claim 29, wherein said segment disk radius is greater than said predefined range of displacement of said leaves. 31. The MLC of claim 21, wherein, for each of said segment disks, said first peripheral portion thereof containing said leaf-side gear segment is non-overlapping with said second peripheral portion thereof containing said motor-side gear segment. 32. A multi-leaf collimator (MLC) for controlling a shape of a high-energy radiation beam emanating from a radiation source and propagating in a direction of propagation, comprising:a plurality of leaves individually displaceable in a direction of displacement that is generally transverse to the direction of propagation, said plurality of leaves having a predefined range of displacement in said direction of displacement, said plurality of leaves collectively having a radiation source-facing side and a patient-facing side opposite said radiation source-facing side, each said leaf including a rack gear extending along the direction of displacement;a plurality of individually rotatable segment disks disposed on said radiation source-facing side of said plurality of leaves, said plurality of segment disks being positioned side by side along a common axis of rotation that is generally transverse to said direction of propagation and to said direction of displacement, each said segment disk corresponding to a respective one of said leaves, each said segment disk including a leaf-side gear segment formed along a first peripheral portion thereof that is engaged with said rack gear of the corresponding leaf to displace the corresponding leaf according to a motor-controlled rotation of said segment disk around said common axis of rotation; anda plurality of motor-driven pinions, each said motor-driven pinion being engaged with a respective one of said segment disks along a motor-side gear segment formed along a second peripheral portion thereof to provide said motor-controlled rotation thereof;wherein each of said plurality of segment disks has a radius along said first and second peripheral portions thereof that is sufficiently comparable to said predefined range of displacement of said leaves such that each said leaf can be fully displaced through its range of displacement in less than one full turn of said segment disk;whereby structural compactness of the MLC is facilitated. 33. The MLC of claim 32, each said motor-driven pinion being coupled to a distinct electrical motor to form a respective plurality of motor-pinion assemblies, wherein said plurality of motor-pinion assemblies are arranged in an arch-like pattern relative to said common axis of rotation of said plurality of segment disks. 34. The MLC of claim 33, wherein said motor-pinion assemblies are mounted on a common bearing block extending peripherally around said plurality of segment disks in an arch-like shape relative to said common axis of rotation, said motor-pinion assemblies being mounted on respective step-like gradations formed in said bearing block along the direction of said common axis of rotation for achieving respective engagement of said motor-driven pinions with said motor-side gear segments of said segment disks. 35. The MLC of claim 33, said plurality of leaves in conjunction with said predefined range of displacement defining an overall lateral range in said direction of displacement, wherein said plurality of segment disks and said plurality of motor-pinion assemblies are configured and dimensioned to be entirely confined within said overall lateral range on said radiation source-facing side of said plurality of leaves. 36. The MLC of claim 32, wherein said segment disk radius is greater than one-half of said predefined range of displacement of said leaves. 37. The MLC of claim 36, wherein said segment disk radius is greater than said predefined range of displacement of said leaves. 38. The MLC of claim 32, wherein the motor-side gear segments of any two adjacent segment disks are staggered in such a way that there will be no angular overlap therebetween throughout the range of displacement of their corresponding leaves. 39. The MLC of claim 38, wherein said motor-driven pinions are wider than their associated motor-side gear segments in a direction of said common axis of rotation. 40. The MLC of claim 39, further comprising a spacer agent disposed between each adjacent pair of said segment disks for reducing mutual friction therebetween. 41. The MLC of claim 32, said plurality of leaves, said plurality of segment disks, and said plurality of motor-driven pinions collectively forming a first leaf/drive assembly, wherein the MLC further comprises a second leaf/drive assembly generally similar to said first leaf-drive assembly and disposed on an opposing side of a center plane of the MLC. 42. The MLC of claim 32, wherein, for each of said segment disks, said first peripheral portion thereof containing said leaf-side gear segment is non-overlapping with said second peripheral portion thereof containing said motor-side gear segment.
048448591	claims	1. Removable and lockable guide ring in an orifice passing through a plate, an internal bore of the guide ring and the orifice in the plate being coaxial relative to one another, and comprising an integral tubular body a first, lower part of which enters the orifice in the plate and a second, upper part of which rests on one of the faces of the plate around the orifice, means of attaching the ring to the edges of the orifice on the face of the plate opposite the face on which the tubular body rests, and an elastic means of returning the attachment means into an attachment position for locking the ring in the orifice in the plate, wherein the attachment means comprising at least two claws each having the form of a bent lever having a first arm an end of which comprises an attachment end, and a second arm forming a predetermined angle relative to the first arm, the claws being mounted freely relative to the tubular body and being held in the tubular body by the elastic means, comprising at least one spring inserted between one part of the tubular body and the second arm of the claws, so as to push the attachment ends of the claws in the radial direction towards the ouside of the tubular body, and wherein said at least one spring is located entirely below said second arm. 2. Guide ring according to the claim 1, wherein the arms of the claws form an angle of approximately 90 degrees. 3. Guide ring according to claim 1, wherein the plate is an end plate of a guide tube of a control rod of a pressurized-water nuclear reactor, the ring having an internal bore for guiding a follower attached to the control rod. 4. Guide ring according to claim 1, wherein the wall of the tubular body has passing through it, in its first part, at least two apertures each communicating with a groove extending in the axial direction of the ring towards the second part of the tubular body, a lateral wall of which has at least two ports passing through said lateral wall in circumferential positions in line with the grooves, a claw being received in a housing comprised of an aperture, a groove and a port placed in line with one another in said axial direction, the first arm of the claw being received in the groove, and the second arm passing through the port to project in the radial direction outside the tubular body. 5. Guide ring according to claim 4, wherein the grooves open into the second part of the tubular body in an annular recess containing the spring. 6. Guide ring according to claim 5, further comprising a piece fastened rigidly to the tubular body to close off the annular recess containing the spring. 7. Guide ring according to claim 4, wherein the aperture passing through the wall of the first part of the tubular body has the form of a cylinder whose axis is inclined relative to the horizontal cross-section of the guide ring, and, on the same side as the outer surface of the first part of the tubular body, is continuous with an axially extending rectangular slot whose width is equal to the diameter of the cylindrical aperture, the end part of the claw having a cylindrical form and dimensions corresponding to those of the aperture, so that, under the action of the spring, the end attachment part of the claw is introduced into the slot, on which the end of the attachment part bears radially when the guide ring is in its locking position. 8. Guide ring according to claim 7, wherein the end attachment part has a bearing and attachment surface of convex form corresponding to the shape of a concave part of the profile of the lower surface of the plate around the orifice.
	claims	1. A channel-cut monochromator for X-rays manufactured by processing grooves on a single crystal block to have a plurality of reflecting surfaces, wherein: (a) said channel-cut monochromator comprises at least two kinds of reflecting surface pairs, which have different reflection Miller indices, processed on a common single crystal block, each of said reflecting surface pairs being composed of a first reflecting surface and a second reflecting surface between which X-rays are reflected an even-number times; (b) any one of said reflecting surface pairs has said first reflecting surface and said second reflecting surface both of which are perpendicular to a common reference plane; (c) said channel-cut monochromator is rotatable around an axis of rotation perpendicular to said reference plane; and (d) an X-ray beam incident on any one of said reflecting surface pairs or an extension line thereof is tangent to a common imaginary circle whose center coincides with said axis of rotation. 2. A channel-cut monochromator according to claim 1 , further comprising a direct path through which an X-ray beam passes without contact with any reflecting surface, and which is tangent to said imaginary circle. claim 1 3. A channel-cut monochromator according to claim 1 , wherein at least one of said reflecting surface pairs comprises at least one asymmetrical reflecting surface. claim 1 4. A channel-cut monochromator for X-rays manufactured by processing grooves on a single crystal block to have a plurality of reflecting surfaces, wherein: (a) said channel-cut monochromator comprises at least two kinds of reflecting surface pairs processed on a common single crystal block, each of said reflecting surface pairs being composed of a first reflecting surface and a second reflecting surface between which X-rays are reflected an even-number times; (b) any one of said reflecting surface pairs has said first reflecting surface and said second reflecting surface both of which are perpendicular to a common reference plane; (c) said channel-cut monochromator is rotatable around an axis of rotation perpendicular to said reference plane; and (d) an X-ray beam incident on any one of said reflecting surface pairs or an extension line thereof is tangent to a common imaginary circle whose center coincides with said axis of rotation; and wherein at least one of said reflecting surface pairs comprises at least one asymmetrical reflecting surface, and said asymmetrical reflecting surface is capable of condensing an X-ray beam width. 5. A channel-cut monochromator for X-rays manufactured by processing grooves on a single crystal block to have a plurality of reflecting surfaces, wherein: (a) said channel-cut monochromator comprises at least two kinds of reflecting surface pairs processed on a common single crystal block, each of said reflecting surface pairs being composed of a first reflecting surface and a second reflecting surface between which X-rays are reflected an even-number times; (b) any one of said reflecting surface pairs has said first reflecting surface and said second reflecting surface both of which are perpendicular to a common reference plane; (c) said channel-cut monochromator is rotatable around an axis of rotation perpendicular to said reference plane; and (d) an X-ray beam incident on any one of said reflecting surface pairs or an extension line thereof is tangent to a common imaginary circle whose center coincides with said axis of rotation; and wherein said channel-cut monochromator comprises at least five kinds of reflecting surface pairs. 6. A channel-cut monochromator according to claim 5 , wherein said block is made of silicon or germanium single crystal and comprises at least five kinds of reflecting surface pairs for reflection. claim 5 7. A channel-cut monochromator according to claim 6 , further comprising a direct path through which an X-ray beam passes without contact with any reflecting surface, and which is tangent to said imaginary circle. claim 6 8. A channel-cut monochromator according to claim 6 , wherein at least one of said reflecting surface pairs comprises at least one asymmetrical reflecting surface. claim 6 9. A channel-cut monochromator according to claim 8 , wherein said asymmetrical reflecting surface is capable of condensing an X-ray beam width. claim 8 10. A channel-cut monochromator according to claim 8 , wherein one of said reflecting surface pairs for reflection comprises at least one asymmetrical reflecting surfaces. claim 8 11. A channel-cut monochromator according to claim 10 , wherein said reflecting surface pair for reflection comprises one asymmetrical reflecting surface and one symmetrical reflecting surface. claim 10 12. A channel-cut monochromator according to claim 10 , wherein said reflecting surface pair for reflection comprises two asymmetrical reflecting surfaces. claim 10 13. A channel-cut monochromator for X-rays manufactured by processing grooves on a single crystal block to have a plurality of reflecting surfaces, wherein: (a) said channel-cut monochromator comprises at least two kinds of reflecting surface pairs processed on a common single crystal block, each of said reflecting surface pairs being composed of a first reflecting surface and a second reflecting surface between which X-rays are reflected an even-number times; (b) any one of said reflecting surface pairs has said first reflecting surface and said second reflecting surface both of which are perpendicular to a common reference plane; (c) said channel-cut monochromator is rotatable around an axis of rotation perpendicular to said reference plane; and (d) an X-ray beam incident on any one of said reflecting surface pairs or an extension line thereof is tangent to a common imaginary circle whose center coincides with said axis of rotation; and wherein a silicon single crystal block having at least two kinds of reflecting surface pairs and a germanium single crystal block having at least two kinds of reflecting surface pairs are fixed to each other, and each block has in common said axis of rotation and said imaginary circle. 14. A channel-cut monochromator according to claim 13 , wherein said silicon single crystal block and said germanium single crystal block each comprise at least three kinds of reflecting surface pairs for reflection. claim 13
	description	This invention relates to monitoring the reactor core in a nuclear power plant and, particularly, to arranging gamma thermometers in the core of a boiling water nuclear reactors (BWR). A typical BWR nuclear power plant includes nuclear instruments that monitor the condition of the reactor core. The signals generated by these instruments are used to maintain the reactor core within allowable operating conditions. The instrument signals may be processed by a core monitor software that determines the 3-Dimensional (3D) nodal powers and the 2-Dimensional (2D) bundle flows. The 3D nodal powers and 2D bundle flows may then be used to determine thermal margins within the reactor core. The operators may use the determined thermal margins to make adjustments to the core operating conditions so not to exceed the allowable operating conditions. Further, the 3D nodal powers and 2D bundle flows may be used by plant operators to confirm that the reactor core is operating within allowable operating conditions. The instruments for a typical BWR nuclear reactor include a Transverse In-core Probe (TIP) system and a Local Power Range Monitor (LPRM) and/or Start-Up Range Neutron Monitor (SRNM) systems. The available types of TIP instruments comprise instruments to measure neutron thermal flux and instruments to measure gamma flux. LPRM instruments generally measure neutron thermal flux. TIP and LPRM instruments are arranged in a core to take axial measurements at fixed radial locations in the core. Conventionally, the TIPs are mechanically moved in and out of the core to calibrate the LPRMs and, particularly, to calibrate individual detectors in each LPRM. During the calibration process, a TIP is positioned next to a detector of a LPRM and the LPRM gain electronics are adjusted to cause the LPRM detector to generate an output signal equivalent to an output signal from the adjacent TIP. In addition, the TIPs may provide processed output signals indicating the neutron thermal flux and gamma flux at various elevations in the reactor core, such as at elevations at six inch (15 centimeters) intervals. The flux measurements taken at various elevations of the core provide axial information regarding the power shape in the core at non-LPRM core locations. Maintaining and operating the TIP mechanical system to raise and lower the TIPs is expensive. Gamma Thermometers (GT) sensors have been used instead of TIPs. Unlike the TIPs that were moved in and out of the core, the GT sensors are positioned at fixed axial locations in the core. Similar to TIPs, the GT sensors are used to calibrate the LPRMs. Because the GT sensors are at fixed axial locations, the expense of a mechanical movement system to raise and lower the TIPs has been eliminated for the stationary GT sensors. In a conventional application, seven or more GT sensors are arranged as a linear array, such as on a vertical string. These vertical arrays of seven GT sensors are positioned at various fixed elevations in the reactor core. The fixed elevations for the GT sensors are manufacturing determined and correspond to fixed axial positions on the string supporting the GT sensors. Fabricating the GT sensors on the strings is problematic due to the narrow tolerances for axial placement of the GT sensors in each string. Each GT of a string must be positioned precisely on the string to be positioned in the core at the elevations to which they are assigned. The GT sensors in each vertical string are each positioned within narrow vertical tolerances to assure that each GT sensor is positioned at its assigned axial position, e.g., adjacent a LPRM when the GT sensor is placed in the core. The narrow vertical tolerances for the GT strings are necessary so that the LPRMs can be accurately calibrated. Each string of GT sensors is permanently fixed in the core after the array is properly positioned and vertically aligned with the LPRMs. To expand the vertical tolerances for the GT strings would, in combination with the inherent uncertainty in any nuclear measurement systems, e.g., the LPRMs, create uncertainties in the determination of the operating conditions of a reactor core. An increase in the uncertainties in the determination of core operating conditions will likely lead to a narrowing of the reactor core operating limits as the operating margins are increased to compensate for the increased uncertainties. The increase margins can result in additional reactor fuel cost as the acceptable operating conditions are narrowed. The narrow axial tolerances applied to the strings of GT sensors are problematic with respect to the manufacturing of these arrays. The tolerances reduce the number of GT sensors that can be accurately positioned on a GT string to, for example, seven GT sensors. The limited number of GT sensors that can be manufactured on each string reduces the amount of core information that can be sensed by the GT sensors. The amount of information regarding the axial power shape of a core that can be sensed by a GT string is dependent on the number of GT sensors vertically arranged on the string. Each GT sensor collects data at a particular axial position on the string, which corresponds to an elevation in the core. Limiting the number of GT sensors on each GT string limits the core elevations for which there is data from GT sensors. The amount of information regarding the axial power shape increases as the number of GT sensors on a string increases. For example, seven GT sensors on a string provides less information regarding the axial power shape at various core elevations than would twenty GT sensors on a string. Increasing the information that is sensed by the GT sensors regarding the axial power shape in a core reduces the uncertainty of that power shape. A reduction in the uncertainty of the power shape allows for a corresponding reduction in the margins applied to the core operation limits. Reducing the uncertainty margins, allows for the core to be operated at conditions that are safe and more efficient respect to fuel consumption. It is conventional for the axial locations of each GT/sensor in a string to be specified prior to the manufacture of the string. In addition, each GT string for a core is manufactured such that the GT sensors are arranged at the same axial locations on each string. Thus, GT sensors are arranged at the same core elevations for every axial location of the GT arrays. Because these GT elevation locations are fixed, the GT elevations are typically hard coded into the core monitoring software. The core monitoring software does not allow for GT sensors to be arranged at core elevations outside the assign axial positions for each sensor and the narrow tolerances predefined for the GT sensors. There is a long felt need for an arrangement of GT sensors that can be readily manufactured and provide an increased amount of information regarding the axial power shape of a core. A method has been developed for collecting data regarding the operating condition of a nuclear reactor core including: positioning a first linear array of gamma thermometer (GT) sensors in a first instrument housing, wherein the GT sensors are arranged asymmetrically along the first linear array; positioning a second linear array of GT sensors in a second instrument housing, wherein the GT sensors are arranged asymmetrically along the second linear array and wherein the second linear array of GT sensors is asymmetrical with respect to the first linear array of GT sensors; positioning the first instrument housing in the reactor core at a first core location and positioning the second instrument housing at a second core location symmetrical with respect to the first core location; collecting core condition data from at least one of the GT sensors in the first linear array of GT sensors, and applying the collected core condition data as data collected from the second linear array. A pair of linear arrays of gamma thermometer (GT) sensors have been developed that are arranged in a nuclear reactor core, the pair comprising: a first linear array of GT sensors, wherein the GT sensors are arranged asymmetrically along a length of the first linear array; a second linear array of GT sensors, wherein the GT sensors are arranged asymmetrically along the second linear array and wherein the second linear array of GT sensors is asymmetrical with respect to the first linear array of GT sensors; the first linear array positioned in the reactor core at a first core location and the second instrument housing positioned at a second core location symmetrical with respect to the first core location. A method has been developed to collect and present data from gamma thermometer (GT) sensors indicative of a nuclear reactor core, the method comprising: forming a plurality linear GT arrays of GT sensors, wherein an axial positions of the GT sensors in each array is not predetermined prior to forming the array; determining the axial position of each of a plurality of GT sensors arranged in each of the linear GT arrays; storing the axial positions for each GT sensors in a data file associated with the linear GT array; loading the data file for each of the linear GT arrays into a core monitor software positioning; for each of the linear GT arrays, the core monitor software determines the elevation in the core of each of the GT sensors based on the axial positions in the data file; positioning the linear GT arrays in the core; collect data regarding an operating condition of the core from the GT sensors for each linear GT array, and the core monitoring software using the collected data to generate a presentation of a core condition at various core elevations. The method may further include: positioning each of the linear GT arrays in a separate instrument tube; positioning detectors for a Local Power Range Monitor (LPRM) in each of the instrument tubes; for each instrument tube, identifying one of the GT sensors of the array adjacent to each of the detectors, and calibrating each of the detectors by either using the identified adjacent GT sensor or a non-adjacent GT sensor that is closest to the detector. FIG. 1 is a schematic illustration of a reactor core 10 contained in a reactor vessel 12 of a BWR. The core 10 comprises, for example, hundreds of fuel bundles 14 (identified in FIG. 1 by white squares with black borders). Each fuel bundle 14 may include an array of fuel rods, water rods and channels. Control rod blades (not shown) are inserting between groups of bundles, e.g., between 2×2 arrays of bundles. Water in the reactor vessel 12 surrounds the rods. Heat generated by nuclear reactions is transferred from the rods to the water circulating through the core, boiling some of the water. The heat generated in the core is carefully controlled to maintain safe and efficient reactor operations in the core. The vessel 12 is typically a cylindrical water filled chamber sealed within the BWR. The fuel bundles are submerged in the water in the vessel 12, wherein the vessel may have a water filled depth of approximately 12 feet (4 meters) or more. FIG. 1 is a top down view of a vessel that shows the two-dimensional (2D) array of hundreds of fuel bundles 14 arranged in the core. The 2D array of fuel bundles may be arranged in rows and columns and have a generally circular perimeter. The center 16 of the array of fuel bundles is referred to as the radial center of the array. Fuel bundles 14 near the center 16 of the array tend to experience more severe operating conditions than do fuel bundles near the outer periphery 18 of the array. Fuel bundles at a common radii 20 in the array will typically experience similar core operating conditions. Various instruments are positioned in the core vessel to monitor conditions in the core. These instruments may include flux measuring devices that measure neutron and gamma flux in the core and provide data indicative of the power of the reactor. For example, startup range neutron monitor (SRNM) channels 22 (black squares in FIG. 1) and local power range monitor (LPRM) systems 28 (black circles). The SRNM channels measure low power ranges of neutron flux at their respective channel location in the core. The core may have other power monitors, such as local power range monitor (LPRM) channels that measure neutron and gamma flux in the core. In addition, TIP or GT instruments may be arranged side-by-side with the LPRMs and SRNMs. The instruments 22 and 28 are arranged at predetermined positions in the core, such at predetermined radial positions with respect to the center 16 of the core. The instruments may be arranged symmetrically in the core such, as is shown in FIG. 1. Each type of instrument, e.g., LPRM and SRNM, may be arranged symmetrically in the core. The monitoring instruments, e.g., LPRM, SRNM, TIPs and GT sensors, measure neutron and/or gamma flux at various radial positions and elevations in the core. The core has a line of symmetry 31 extending through the center 16 of the core 10. The core is symmetrical about the line of symmetry. The line of symmetry is typically a line extending at 45 degrees from the 1:30 position to the 7:30 position on the core, as shown in FIG. 1. In particular, the conditions in the arrangement and condition of the core mirrored on opposite sides of the line of symmetry. With respect to the instruments 33 at the same (common) distance from the line of symmetry and along a line perpendicular to the line of symmetry receive a uniform amount of flux from the core, provided that they are at the same elevation in the core. By way of example, instruments 33, e.g., a pair of GT string, are the same distance from the line of symmetry 31 along a line perpendicular to the line of symmetry. GT sensors in this pair of GT strings at the same elevation in the core will receive the same amount of flux from the core. FIG. 2 shows a conventional arrangement of fuel bundles 14, surrounding a flux measuring instrument 30, and adjacent a control rod blade 32. The flux measuring instrument 30 may be housed in a sealed hollow instrument tube 34 containing a linear GT string array 36 of individual sensors, including Gamma Thermometers (GT sensors) 38 and neutron flux detectors 40 for a LPRM or SRNM. The instruments 30 are typically arranged between a group of four adjacent fuel bundles. Each instrument 30 extends substantially the length of the fuel bundles, preferably at least the length of the active portion of the fuel rods in the bundle. The instruments are at the same elevation in the core as the fuel bundles and are preferably fixed in the core during core nuclear operation. The instruments 30 are arranged in a core to collect data regarding neutron and gamma flux levels at various radial positions in the core. At each radial position in the core, the instruments collect flux level data at various elevations in the core. Within each instrument tube 34, individual measuring sensors 38, 40 are arranged at various axial locations along the length of the tube. When the tube is in the core, these axial locations in the sensors correspond to various elevations in the core. The sensors provide data on flux levels for each of the axial locations of the sensors in the instrument tube 34 and at the core elevations corresponding to the axial locations. In each instrument tube 34, the sensors may include, for example, GT sensors, LPRMs and SRNMs. The sensors may also include TIPs, that are arranged adjacent the instrument tube. However, TIPs are typically not fixed in the core, and are moved mechanically in and out of the core by the operators of the plant. TIPs are preferably not used because they require complex and expensive mechanical conveyors to move the TIPs within the core. A benefit of the GT sensors over the TIPs is that GT sensors may be fixed in a tube housing 34 with the LPRMs or SRNMs. Each GT sensor 38 generates output signals providing data of gamma flux levels at the radial location on the string for the sensor which corresponds to an elevation in the core. The GT sensors in each tube housing 34 may be used to calibrate the LPRMs 40 in the same tube. To calibrate the LPRMs, the string 36 of GT sensors is positioned in the tube such that a GT sensor is axially aligned next to a LPRM in the tube. To calibrate a LPRM, the output signal of the LPRM is adjusted to match the output signal of the GT sensor adjacent to the LPRM. The output signal of the LPRM may be adjusted by changing the gain of the electronics in the LPRM that processes the signal from the LPRM. In addition, the GT sensors may be used to acquire flux data from the core. The arrangement of GT sensors 38 in the tube 34 shown in FIG. 2 are arranged at predetermined locations along the GT string 36. The GT string and its GT sensors are conventional. These predetermined locations are specified prior to the manufacturing process for the GT string. The string is manufactured such that the GT sensors 38 are positioned at the predetermined axial locations along the string. Typically, a relatively small number of GT sensors, e.g., seven, are positioned on a string at specified axial locations corresponding to the locations of the sensors for the LPRM. The predetermined axial string positions for the GT sensors are subject to narrow tolerances to ensure that the GT sensors are positioned adjacent the LPRM sensors. Positioning the GT sensors at the same axial position as the LPRM sensors is needed to accurately calibrate each LPRM sensor with an adjacent GT sensor. The greater the axial distance between the GT sensor and LPRM sensor, the greater the uncertainty in the calibration of the LPRM sensor. Due to restrictions of the manufacturing process for strings of GT and the tight axial tolerances for positioning GT sensors on the string, conventional strings about seven GT sensors is the maximum number of sensors that can be economically manufactured using conventional string manufacturing processes. In addition to calibrating the LPRM sensors, flux data from the GT sensors may be used to determine the power shape of the core. The limited number of GT sensors reduces the flux data that can be acquired at different core elevations by the GT sensors. The resolution of the determined power shape of the core is dependent, in part, on the number of GT sensors on each string. With only seven GT sensors on each string, the resolution of the power shape (as determined by the GT sensors) in an axial direction of the core, e.g., at different core elevations, is relatively coarse. The resolution of the determined power shape of the core may be increased by increasing the number of GT sensors on each GT string. A new GT string structure and method for fabricating the strings has been developed that eliminates the prefabrication axial positioning restriction on GT sensors in a string. By eliminating this positioning restriction, a relatively large number, e.g., sixteen or greater, of GT sensors may be arranged in a string. FIG. 3 is a schematic diagram of an instrument tube housing sensors 34 for an LPRM and a GT string 42 with a relatively large number of GT sensors 38. The tube housing, LPRM and GT sensors may be conventional. The fabrication process for the GT string may also be conventional, with the exception that the fabrication process is not limited to placing the GT sensors at predetermined axial positions on the string. The fabrication process may include positioning a relatively large number of GT sensors, e.g., sixteen or more, in a metallic rod of, for example, having a diameter of two inches (5 cm) and length of one foot (30 cm). FIG. 4 is a flow chart of a fabrication method and registration method for the GT strings. In step 60, the number of GT sensors placed in the rod may be as many as can fit in the pre-extruded rod. In step 62, the rod with the GT sensors is stretched through a conventional extrusion process to a length of, for example, 15 feet (4.5 m) and a diameter of 0.25 inches (6 mm). During the extrusion process, the GT sensors are somewhat arbitrarily dispersed along the axial length of the rod, as is shown in the stretched rod 42 shown in FIG. 3. After the extrusion process, the axial position 44 of each GT sensor 38 in the string is measured using conventional GT sensor detection instruments and methods, in step 64. The axial position 44 of each GT sensor is determined to a high degree of accuracy, e.g., to within 1 to 5 mm. The axial position of each GT sensor is registered and stored for future reference when using the particular GT string, in step 66. The axial position of each GT sensor is stored in a computer data file that is associated with the corresponding GT string. The data file of the axial positions of each GT sensor in a GT string is loaded into a computer system for the reactor core that includes a core monitoring software program, in step 68. This program may be a conventional system for monitoring reactor core operations, analyzing data from instruments and sensors monitoring the core and generating maps showing the power shape of the core in three dimensions (3D), including in radial directions and elevations in the core. Conventional core monitoring programs have hard coded axial positions as to known GT sensor positions in each GT string. For present purposes, the program is modified to accept an input file of the GT sensor positions for each of the GT strings in instrument tubes arranged in the core. In particular, the software program reads the data file of the axial positions of GT sensor for each GT string and determines the elevation in the core for each GT sensor, in step 70. The software associates the elevation of each GT sensor with the radial position of the GT string in the core (and optionally the angle of the radii associated with the sensor). Knowing the elevation and radial position of each GT sensor, the core monitoring software uses the data collected by the sensors to generate a power shape map of the core, in step 72. An exemplary 3D power shape graph 55 is shown in FIG. 5. The 3D power shape graph presents radial power distributions at different axial elevations (E) in the core corresponding to the elevation of the GT sensors on the GT strings. Core locations 57 may be color coded to represent the core power at each location. For example, a core location 57 shown in red may indicate a higher power level than another core location shown in blue. The core locations are shown in the 3D graph 55 at various core elevations 53 and at different core locations on each elevation. In addition, the manufacturer of the instrument tube 34 or the core monitoring software identifies the GT sensor 46 in a string adjacent each detector of the LPRM 40 in the tube, in step 74. The identified GT sensor adjacent each detector is used to calibrate the detector, in step 76. Because of the relatively large number of GT sensors arranged in the string, there will be a GT sensor adjacent each LPRM detector. The GT sensors adjacent each LPRM detector can be determined knowing the axial location of both the GT sensors and the LPRM detectors. If the adjacent GT sensor fails or is unavailable, the closest GT sensor to the detector will be used to provide the detector calibration. Using a GT sensor that is further away from the detector, increases the uncertainty of the calibration. Having the manufacture of the GT strings determine the as-built axial locations of the GT sensors after the string is fabricated allows for the removal of the requirement that a particular GT sensor be positioned at a specific axial location in the string. Without this requirement, many more GT sensors, e.g., two to three times more GT sensors, can be placed in a GT string than when the requirement is imposed. By increasing the number of GT sensors in a string increases the elevations in the core at which GT sensors collect flux data regarding the power of the core. FIG. 6 is a schematic diagram of a pair of instrument tubes 52, 54 at symmetrical positions (see instruments 33 in FIG. 1) in the core and having GT strings 56, 58 with GT sensors arranged asymmetrically on the string. The pair of instrument tubes 52, 54 that are arranged at a common distance from the line of symmetry 31 (FIG. 1). These tubes 52, 54 both have a GT string 56, 58 of GT sensors and a LPRM 40. The GT strings 56, 58 detect the same core conditions, e.g., gamma flux, because they are at symmetric positions in the core. Symmetric positions in the core include positions at the same core elevation, and same distance from the line of symmetry along a line core perpendicular to the line of symmetry. For symmetrically situated GT strings, it can be safely assumed that the flux data collected by the GT sensors on the strings should be the same, except for elevation variations of the GT sensors. The asymmetrical arrangement of GT sensors in the strings may be intentional and based on axial locations of GT sensors determined before the fabrication of the GT string. Because of the predetermined axial locations of the GT sensors, the number of sensors that can be arranged in the string is limited to, for example seven GT sensors. The GT string 56 on the left side of FIG. 6 has more GT sensors towards the top half of the string than in the bottom half. Similarly, the GT string on the right side of FIG. 6 has more GT sensors in the bottom half of the string than in the top half. Another form of asymmetric locations is to have the same number of GT sensors on two strings at symmetric and the axial locations of GT sensors on each string be different. The asymmetrical axial arrangement of GT sensors in each string results in at least some of the GT sensors in one string 56 being at elevations in the core at which there is no corresponding GT sensor in the other string 58. At those core elevations, only one GT string is acquiring data regarding the flux in the core. In addition, there are fewer GT sensors than detectors 40 for the LPRM. Calibrating the detectors without an associated GT sensor is problematic. The LPRM detector axial positions are identical for both instrument tubes 56, 58. The symmetrically positioned GT strings 56, 58, have GT sensors arranged such that collectively for both strings there is a GT sensor at each of the LPRM detector axial positions in the instrument tube. The core monitor software identifies GT sensors that are directly adjacent a LPRM detector, e.g., at the same elevation, or the GT sensors closest to the LPRM detector. If only one GT sensor is identified (for example a GT sensor is at the same elevation ad the detector), the detector is calibrated by adjusting the electronics associated with the LPRM detector until the detector outputs the same signal level as being sensed by the identified GT sensor. If multiple GT sensors are identified as being proximate a LPRM detector, the core monitor software may interpolate the GT sensor signals, such as by a weighted average based on the axial distance from each GT sensor to the same elevation of the detector. The weighted average of GT sensors is used by the software as a reference signal to which is matched the output of the LPRM detector by adjusting the electronics for the LPRM detector. The core monitor software infers that GT strings at symmetric locations, e.g., at the same distance from the radial symmetry axis, experience the same core conditions, including core flux. Based on this inference, the core monitor software applies the GT sensor data from one string (the “originating string”) as data that is collected a both the originating string and data collected at another string at a symmetrical core location with the originating string, unless the other string has a GT sensor at the same core location. Similarly, a LPRM detector 40 that does not have a directly adjacent GT sensor may be calibrated using the data from the GT sensor at the same or close elevation as the detector and in an instrument tube at a symmetrical core location. FIG. 7 is a chart showing the strong correlation of data from GT sensors (GT-1, GT-2) and data from detectors (LPRMs-string-1, LPRM-string-2), where the sensors and detectors are at almost the same axial locations in an instrument tube, and the tubes are symmetrically arranged in the core. In particular, a first instrument tube with GT sensors (GT-1 and detectors (LPRMs-string-1) is symmetrically arranged in the core with a second tube with GT sensors (GT-2) and detectors (LPRMs-string-2). The GT sensors at the same axial height in both tubes have substantially the same expected data signal output. Similarly, the detectors (LPRMs) in both tubes and at the same axial height have the same expected detector signal output, which is equal to the GT sensor signals for the corresponding elevation. Because of the similarity of signals, a GT sensor in one tube can be used to calibrate a detector in another tube, provided the detector is at the same elevation as the GT sensor and the tube with the detector is symmetrically arranged in the core with the tube having the GT sensor. The core monitor software identifies GT strings in the core at symmetrical core locations, in step 78 of FIG. 4. For symmetrical GT strings, the controller identifies core elevations at which one but not both of the symmetrical strings have a GT sensor, in step 80. For the one GT sensor at each of the identified core elevations, apply the sensor data as originating from both symmetrical strings in analyzing the condition of the core, such as in generating a power shape of the core. To calibrate the LPRM detectors, the core monitor software identifies detectors 40 that do not have an adjacent GT sensor in the same instrument tube, in step 84. To calibrate each of the identified detectors 40, use the data from the GT sensor at the same axial position as the detector but from an instrument tube symmetrically positioned in the core. The two schemes disclosed herein for increasing the effective number of GT sensors monitoring a core reduce the uncertainty in the nuclear measurement system relative to the current economical optimum of about only seven GT sensors per GT string. The first scheme increases the number of GT sensors in each string, but has a risk of not positioning a GT sensor directly adjacent a LPRM detector for calibration purposes. The second scheme uses a reduced number (albeit a conventional number) of GT sensors but relies on an asymmetrical arrangement of sensors and substitution of sensor data between symmetrically positioned instrument tubes in the core. Both schemes collect data at more core elevations than can be achieved by known methods of fabricating GT strings. Reducing the uncertainty in the measurement system favorably impacts the operating limits for the core and can result in better core performance, lower fuel costs and the elimination of complex TIPs. The monitoring software uses the asymmetrical arrangement of sensors to substitution sensor data between symmetrically positioned instrument tubes in the core, the monitoring and to calibrate the detectors, e.g., LPRM detectors in different tubes. This capability allows the software to calibrate a greater number of LPRM detectors than the number of fixed GT sensors. The calibration can be conducted frequently to ensure that the data generated by the detectors is accurately calibrated. Further, detectors that are not adjacent GT sensors may be calibrated by using data from GT sensors that are symmetrically positioned in another instrument tube. While the substitutability of sensor data from symmetrical locations in a core is conventional, this disclosure presents an application of the substitutability of sensor data that was previously unknown to the best of the knowledge of the inventors. While the invention has been described in connection with what is presently considered to be the most practical and preferred embodiment, it is to be understood that the invention is not to be limited to the disclosed embodiment, but on the contrary, is intended to cover various modifications and equivalent arrangements included within the spirit and scope of the appended claims.
	claims	1. A drawing apparatus including a plurality of drawing units each of which is configured to perform drawing on a substrate with a charged particle beam, the apparatus comprising:a plurality of first processors configured to be selectively connectable to each of the plurality of drawing units;an information processor configured to determine, from the plurality of first processors, a first processor to be connected to a first drawing unit among the plurality of drawing units, based on drawing data; anda connection unit configured to connect the determined first processor to the first drawing unit. 2. The apparatus according to claim 1, wherein the determined first processor is configured to correct the drawing data. 3. The apparatus according to claim 1, wherein the information processor is configured to determine the first processor to be connected to the first drawing unit, based on a size of a drawing area corresponding to the drawing data. 4. The apparatus according to claim 3, further comprising:a plurality of second processors respectively connected to the plurality of drawing units,wherein the determination unit is configured to determine the first processor to be connected to the first drawing unit if the size of the drawing area exceeds a tolerance associated with the second processor. 5. The apparatus according to claim 1, whereineach of the plurality of drawing units includes a deflector configured to blank the charged particle beam, andthe drawing data includes control data for the deflector. 6. The apparatus according to claim 1, whereineach of the plurality of drawing units includes a deflector configured to scan the charged particle beam on the substrate, andthe drawing data includes control data for the deflector. 7. A method of manufacturing an article, the method comprising:performing drawing on a substrate using a drawing apparatus;developing the substrate on which the drawing has been performed; andprocessing the developed substrate to manufacture the article,wherein the drawing apparatus includes a plurality of drawing units each of which is configured to perform drawing on a substrate with a charged particle beam, the apparatus including:a plurality of first processors configured to be selectively connectable to each of the plurality of drawing units;an information processor configured to determine, from the plurality of first processors, a first processor to be connected to a first drawing unit among the plurality of drawing units, based on drawing data; anda connection unit configured to connect the determined first processor to the first drawing unit. 8. An information processing apparatus connected to a drawing apparatus including a plurality of drawing units each of which is configured to perform drawing on a substrate with a charged particle beam, the drawing apparatus including a plurality of first processors configured to be selectively connectable to each of the plurality of drawing units, and a connection unit configured to connect one of the plurality of first processors to one of the plurality of drawing units, wherein the information processing apparatus is configured todetermine, from the plurality of first processors, a first processor to be connected to a first drawing unit among the plurality of drawing units, based on drawing data, andcontrol the connection unit so that the determined first processor is connected to the first drawing unit.
	description	This application claims priority of European application No. 08017322.2 filed Oct. 1, 2008, which is incorporated by reference herein in its entirety. This invention relates to a method of condition monitoring of a wind turbine, and a wind turbine condition monitoring system therefore. In order to detect irregularities during the operation of a wind turbine and to trigger an alarm condition based upon the detected irregularity, wind turbine monitoring systems of the state of the art obtain measurement values of various kinds and evaluate the values thus obtained to determine if a predefined alarm condition is satisfied. The system proposed in European patent specification EP 1 531 376 B1 begins to capture input measurement values whenever predetermined operation variables of the wind turbine are within a certain predetermined operation range, said operation variables being received from the control system of the wind turbine. Such taking of data upon occurrence of predetermined conditions is often described as triggered capturing or triggered measurements. However, as data capturing commences after a pre-specified triggering condition has been met, the capturing of data may suffer from the incoming data becoming unstable during the intended capturing time. Such instability of the captured values may involve sudden changes of the operating parameters of the wind turbine. The system proposed in the aforementioned European patent specification detects if any of the measured operation variables varies during the entire signal capturing beyond a predefined upper or lower limit, and, if so, aborts the capturing and discards the captured data. A consequence of this is that the capturing procedure must be started again as soon as the triggering condition is met again and further that valuable information is lost by discarding the recorded data. United States patent application publication US 2003/0200014 A1 discusses the need for different alarm thresholds depending on the operating parameters of the monitored device. A similar observation is made regarding wind turbines in Orbit 2Q04, p. 21-27. It is thus an object of present invention to provide a method of condition monitoring of a wind turbine that avoids the loss of input data that might be used for valid condition monitoring even where temporary time intervals of unstable input data occur, and a wind turbine condition monitoring system therefore. This object is achieved by the subject-matter of the claims. The invention according to claim 1 is a method of condition monitoring of a wind turbine, comprising the steps of: capturing measurement data to be evaluated and operating condition parameters; selecting a subset of operating condition parameters from the captured operating condition parameters; calculating a set of n characterizing moments based on the selected subset of operating condition parameters; providing a finite n-dimensional space each of the n dimensions representing possible values for one of the characterising moments the n-dimensional space being sub-divided into a number of bins, where each bin represents a n-dimensional interval defined by n one-dimensional intervals each of the n one-dimensional intervals representing an interval in one of the n dimensions, and where each bin defines an acceptable range for the set of n characterizing moments; determining if the set of calculated characterising moments belongs to one of the bins; accepting the set of characterising moments and (typically processed) measurement data if it is detected that the set of characterising moments belongs to one of the bins. Moreover, at least one index for each bin may be provided and the captured measurement data and characterising moments may be tagged with the index, or indices, of the bin to which the set of accepted characterising moments belongs. By calculating characterizing moments of the subset of operating condition parameters and by then determining if the calculated characterizing moments match any one of the bins (that is, one single of the bins), it is established that even when operating parameters undergo rapid changes, data captured from the various sensors can still be used as long as they can be correlated with the characterizing moments and therefore assigned to their respective bin. In this way, discarding of captured data that might have been used for condition monitoring is avoided, and all data that match the correct bin are indeed used for condition monitoring with the present method. Another technical effect of this is that the captured measurement data do not necessarily have to correlate in time but can origin from different times of measuring. With the present method (and system), it is merely required that the characterising moments associated to the captured measurement data match the same bin. Thus, the present method enables to use any valid set of captured measurement data, which can be selected from a continuous capturing of measurement data. As soon as the set of captured measurement data values matches (any) one single of the bins, the data can be used for further evaluation and generation of alarms. Discarding of valuable data becomes unnecessary. The invention can be embodied as provided in the claims referring back to claim 1 and/or as given in the detailed embodiment. For example, in case it is determined that the characterizing moments do not belong to one of the bins, the step of determining if the set of calculated characterising moments belongs to one of the bins may be performed again at a later point of time, i.e. with characterizing moments which are based on a later selected subset of operating condition parameters. Thus, it is established that monitoring is always based on a sound data set, and that the system automatically retries obtaining a proper data set in case of temporary lack of valid data. Additionally or alternatively, in case it is determined that the characterizing moments do belong to one of the bins, it is determined whether measurement data which has undergone one or more processing steps (called processed measurement data in the following) and characterising moments are to be stored in a long-term storage. The step of determining whether the processed measurement data and characterising moments are to be stored in a long-terms storage may comprise a step of detecting whether an alarm status has changed with respect to the alarm status of the processed measurement data and characterising moments contained in the long-term storage and the step of storing the processed measurement data and characterising moments in the long-term storage if it is detected that the alarm status has changed. Additionally or alternatively, step of determining whether the processed measurement data and characterising moments are to be stored in a long-term storage may comprise the step of determining whether a given time has passed since the last storing of processed measurement data and characterising moments in the long-term storage and the step of storing the processed measurement data and characterising moments in the long-term storage if it is detected that the given time has passed. The step of determining whether a given time has passed since the last storing of processed measurement data and characterising moments in the long-term storage is, in a special embodiment, only performed if it is detected that the alarm status of the processed measurement data and characterising moments has not changed with respect to the alarm status of the processed measurement data and characterising moments contained in the long-term storage. Thus it is established that the stored association of processed measurement data to bins is always up-to-date. Embodiments may also be characterized in that measurement data and operating condition parameters are continuously captured (obtained and recorded). By such continuous capturing, that may be performed in parallel with the remaining steps of the method so as to run in the background, there is always a sufficient amount of current data available. In embodiments, the set of input captured measurement data and operating condition parameters may comprise measurement data from a condition monitoring system and/or a vibration measurement value and/or a strain gauge measurement value and/or a wind speed measurement value and/or a rotor rotational speed value and/or a generated power value and/or a temperature measurement value and/or a measurement value representative of the amount of metal particles detected in the lubricating oil of the wind turbine. In embodiments, an evaluation method for the captured measurement data and operating condition parameters can be selected based on evaluating a rule definition. In such embodiments, a rule definition repository may contain at least one rule expression and the rule expression may define the type of evaluation method and/or the time of evaluating the captured measurement data and/or the frequency of evaluating the captured measurement data and/or the type of data to be used for evaluating the captured measurement data and/or the amount of data to be used for evaluating the captured measurement data. Thus, the behaviour of the wind turbine condition monitoring system can be easily adjusted to facilitate different factory settings for different product lines, system parameterization during the initial installing of the wind turbine, or maintenance during the lifetime of the installed wind turbine. In embodiments, calculating a characterizing moment can comprise calculating a root mean square and/or a mean value of the selected subset of operating condition parameters. In embodiments of the method, evaluating the captured measurement data may comprises the steps of processing the captured measurement data and comparing the processed measurement data to a predetermined threshold for the bin the set of calculated characterising moments associated to the processed measurement data belongs to. Embodiments may further be characterized in that evaluating the captured measurement data comprises processing the captured measurement data and comparing the processed measurement data with previously processed measurement data associated to characterising moments which belong to the same bin. The invention according to claim 14 teaches a wind turbine monitoring system for the method of condition monitoring of a wind turbine according to any one of claims 1 to 14, comprising a memory or buffer and a processing unit, characterized in that the memory or buffer comprises captured and processed measurement data and captured operating parameters, and a number of bins; and the processing unit is configured to carry out the method steps of the method according to any one of claims 1 to 14. As the person skilled in the art will acknowledge, the wind turbine monitoring system can be embodied by having the processing unit be configured to execute any of the method steps specified in the claims referring back to claim 1 and/or as specified in the detailed description. The processing unit may be configured to execute any of the method steps by appropriate programming, or any combination of specific hardware and/or general-purpose processors and/or application specific integrated circuits and/or software. FIG. 1 shows a schematic overview of a first part of an embodiment of the method of present invention. The first part deals with evaluating captured measurement data. In step 100, the processing unit of a wind turbine control system continuously captures measurement data to be evaluated and operating condition parameters. These captured measurement data to be evaluated and operating condition parameters may comprise data output from the condition monitor sensors, from meteorological sensors, as well as operating parameters of the wind turbine. The parameters thus measured may, e.g., include the rotating speed of the wind turbine, the power generated by the wind turbine, the wind speed, the temperature of various parts of the wind turbine, vibration data from sensors placed on various parts of the wind turbine, and/or the number of metallic particles detected in the lubricating oil in the wind turbine. As the person skilled in the art will understand, the measurement data to be evaluated and operating condition parameters given herein are referred to by example only, and further kinds of measurement data and operating condition parameters may be used with the present invention as well. The signals containing measurement data to be evaluated and operating condition parameters are captured from various sources. Some signals are captured from sources that are strictly synchronous at high data rates (RPM sampling up to 100 MHz, dynamic signals at 41 kHz, etc.). Other signals are captured from external sources and are tagged with time at the origin. In this case the tagging is synchronised with the wind turbine condition monitoring system. Capturing measurement data is a real-time process that may involve some degree of ring buffering. The captured measurement data to be evaluated and operating condition parameters are either processed in real-time (Step 110) or stored in a memory or a buffer (Step 120) for subsequent offline processing (step 130). Real-time in the sense of the invention means “without loss of data” rather than “guaranteed response time”. Examples for real-time processing of captured signals are pulse-counting on oil-debris monitoring equipment, safety critical monitoring of general vibration levels, tower sway detection, over speed detection, etc. Offline processing is used for more sophisticated analysis of captured measurement data, i.e. analysis that is computationally more demanding than the real-time processing and leads, e.g., to autospectra, time series, envelopes, etc. It can take up to one hour to process buffered measurement data. Real-time processing as well as offline processing can lead to specific moments and/or analysis results. Specific moments are calculated from a subset of the captured operating condition parameters which describes the operational state of the turbine. The calculated specific moments allow for characterising the operating condition of the wind turbine. Hence, the specific moments are referred to as characterizing moments in the entirety of the present document. These characterizing moments may, e.g., be a mean value or a root mean square of the parameter values in the subset of operating condition parameters. Real-time processing results and offline processing results are tagged with the time of their original capture and stored in buffers that comprise memory depths measured in hours in step 140. Buffering is done to account for possible processing and communication delays. Actual processing schedules for real-time and offline processing, i.e. certain evaluation methods that are prioritised to apply to measurement data captured in a specified time interval, are selected by the processing unit. Note, that the specified time interval during which the measurement data was captured must match the time interval during which the operating condition parameters for calculating the characterizing moments were captured. Selecting the processing schedules can be implemented by the processing unit activating a certain prespecified data evaluation method, or by selecting a data evaluation method from a number of data evaluation methods implemented in the processing unit. The selection may be implemented based on a rule definition repository in the memory of the wind turbine monitoring system, the rule definition repository comprising a set of predetermined rules which define when and how often a given method is to be applied. In this way, the method and system are enabled to apply different evaluation methods at different times and for different time intervals of captured input measurement values. Thus, the processing unit may evaluate the rule definition repository to select a data evaluation method according to the rule definition comprised therein. Capturing of measurement data to be evaluated and operating condition parameters, real-time processing and offline processing, takes place in parallel with all of the remaining method steps that will be described, so that while the processing unit carries out the further steps as described in the following, measurement data to be evaluated and operating condition parameters are continuously captured and processed in the background. The characterizing moments span a finite n-dimensional space, with one dimension for each of the specific moments discussed above. This n-dimensional space is divided into n-dimensional intervals called bins, where each dimension of an interval represents the range of an acceptable value for one of the characterizing moments. In step 150, the processing unit compares the characterizing moments with a number of bins stored in the memory of the wind turbine monitoring system according to time tags and determines if the respective calculated characterizing moments fall into one of the bins. If the characterizing moments fall into one of the bins, they are recognised as being within acceptable ranges. This process runs in parallel to capture and processing (steps 100 to 140) and also cleans the buffer or buffers used in step 140 for buffering the real-time processing results and the offline processing results, i.e. the characterising moments and the analysis results. In case the characterizing moments do not fall into one of the bins, a wait state is executed in step 160. After a certain waiting time has passed, the system returns to step 150 (compare actual characterizing moments with bins). If, on the other hand, the characterizing moments fall into one of the bins, the captured measurement data is accepted by the system and the processing unit proceeds to step 170. Since evaluating the captured measurement data requires an amount of data that is specific to the selected data evaluation method, and further requires data captured over the time interval specified for the selected data evaluation method, the processing unit determines in step 170 if the required input values are available. If the required input variables are not available, a wait state is executed in step 180. Then, after a certain waiting time has passed, the system returns to step 150 (compare actual characterizing moments with bins). If the required input variables are available, the processing unit proceeds to step 190 evaluate as to whether the processed measurement data and characterising moments are to be stored in a long-term storage. This evaluation can, e.g., involve comparing the processed measurement data with predetermined thresholds for the bin that the data belongs to, i.e. the bin the characterizing moments which are associated with the measurement data falls into. In embodiments, evaluation may alternatively or at the same time involve comparing the processed measurement data with processed measurement data captured at an earlier point of time and which belongs to the same bin (trending). The result of such evaluation can be the generation of alarms, as a result of which the condition monitoring system applying the present method may enter an alarm status corresponding to the generated alarm. The processed measurement data and characterising moments can be stored in the long-term storage if the evaluation shows a change in an alarm state, or if a given time has lapsed since the last storing of processed measurement data and characterising moments. FIG. 2 shows a schematic overview of the evaluation as to whether the processed measurement data and characterising moments are to be stored in a long-term storage. Hence, the FIG. 2 deals with handling and storing the processed measurement data and characterising moments after evaluating to which bin these data belong to. In step 191, the processing unit determines the point of time at which processed measurement data and characterising moments served as the basis for an evaluation has been stored for the last time. In step 192, the processing unit determines if a change of the alarm status has occurred between the actual measurement data and characterising moments and the last data storing, as determined in step 191. If such a change of alarm status has occurred, the method is continued in step 193 in which the processed measurement data and characterising moments is stored in the long term storage. If no change in the alarm status has occurred since the time of last storing, method execution is continued in step 194, in which the processing unit determines if a given time has lapsed since the last storing of processed measurement data and characterising moments. If the given time has lapsed, the method continues to step 193, in which the processed measurement data and characterising moments is stored in the long term storage. If no change in the alarm status has occurred and the given time has not lapsed since the last storing, the processed measurement data and characterising moments can be discarded in step 195. FIG. 3 provides a schematic overview of a wind turbine condition monitoring system for use with the embodiment of the method described above. Wind turbine condition monitoring system 1 comprises a processing unit 2 and a memory or buffer 3. As indicated by the arrow, the wind turbine condition monitoring system can be coupled to a wind turbine, or a wind turbine control system, for obtaining the input values required by the condition monitoring system and further for outputting data or commands generated by the condition monitoring system. Memory 3 comprises a rule definition repository 10, in which a number of rules 11, 15 are defined. Rule definition 11 comprises a definition of an evaluation method 12 with a point of time 13 that is associated to the defined evaluation method 12 to define when the evaluation method is to be applied and further comprises an interval definition 14 which specifies the time interval of input data required for data evaluation. Likewise, rule definition 15 comprises a definition for an evaluation method 16 and an associated point of time for execution 17 and further an associated interval of measurement data for use with the evaluation method as defined in 16. Note that the number of rules shown in FIG. 3 is only an exemplary description and that the number of rules may, in reality, be larger than two. Memory 3 also comprises a storage area 20 for measurement data to be evaluated 21, 22, and operating condition parameters 23, 24. Note that the number of measurement data to be evaluated shown in FIG. 3 and the number of operating condition parameters shown in FIG. 3 are only an exemplary description and that the number of measurement data to be evaluated and the number of operating condition parameters may, in reality, be much larger. Further, memory 3 comprises a bin definition repository 30, in which bins 31, 32, 33, and 34 are defined. Each of these bins is a range of acceptable values of the characterizing moments, which is given as an interval of an n-dimensional space. Note that the number of bins shown in FIG. 3 is an exemplary description only and that the number of bins may, in reality, be larger or smaller. The present method and system allows to use the maximum possible amount of captured input data and reduces the amount of discarded data to a minimum. In particular, there is no discarding of valid data due to rapid changes of operating parameters during capturing of the data. In the present method and system, data captured from the various sensors can still be used as long as they can be correlated with the characterizing moments and therefore assigned to their correct bin. The approach proposed herein also implies that measurements do not necessarily have to be performed simultaneously in order to be evaluated, or compared to each other. It is just necessary that the measurements, that are the captured input values, are assigned to the same bin. By correlating all data measured (captured input values, for instance obtained from a condition monitoring system) with a set of values that characterizes the operational state of the turbine (characterizing moments), changes in the measured data that are caused by changes in the operational state of the wind turbine can be separated from changes which are caused by changes in the general condition of the wind turbine, such as technical faults or wear of the wind turbine. Thus, the present method and system further enables to detect slow, long-term changes in the condition of the turbine by trending. At the same time, the rate of false alarms is kept low while also the risk of missing detection is kept to a minimum.
056132433	description	The following examples further illustrate the invention. EXAMPLES Chloride solutions having the compositions indicated in the attached Table 1 were first evaporated to dryness at 80.degree. C. to produce solid residues. These residues were then held under a flow of steam at 200.degree. C. for one hour and then under a flow of steam and air at 800.degree. C. for two hours, ensuring both the completion of all possible hydrolysis and the development of crystalline properties. The granular solid residues were then allowed to cool in air. The solid wastes were then leached at room temperature (62.5 gpL) in synthetic groundwater (5 gpL sodium chloride, 500 mgpL sulphuric acid) maintained at pH below 5 by periodic additions of acetic acid. The leach was continued for 24 hours, after which the residue was filtered, washed with fresh synthetic groundwater and dried. Roasted and leached wastes were subjected to chemical analysis and gamma spectroscopy analysis for major elements and radionuclides. Radionuclide extraction from the solid wastes in leaching is also indicated for each case in the attached Table 1. Clearly those samples having lanthanide (eg. Ce) and P additions under circumstances which produced a waste needing little or no acid addition to maintain pH below 5 provided wastes which did not subsequently allow leaching of radionuclides. The absence of these elements or conditions resulted in a far less stable waste. However, it is expected that other elements may substitute for these main constituents, allowing for a range of effective compositions, provided that the effective circumstances as disclosed are maintained. Further, the addition of barium salts (made to liquor A1-9 of the attached table in a separate test) was found to have a strongly negative impact on the stability of uranium and radium in the wastes produced by otherwise identical treatment. Hence wastes containing barium, lanthanide and phosphorus (as have previously been produced in waste forms, due to the composition of wastes from nuclear fuel processing which contain zirconium and phosphorus) are herein disclosed as ineffective for the purposes for which the present invention is practised. In general where the effectiveness of the process depends on the presence of phosphorus and lanthanides the presence of elements which form more stable phosphates than lanthanides may require the addition of incremental compensating phosphorus for all other identical conditions. Solutions derived from the production of synthetic rutile by acid leaching of thermally treated ilmenite to which additives were made to result in solutions having the composition indicated in the attached Table 2 were also treated according to the method described above. Roasted and leached wastes were subjected to chemical analysis and gamma spectroscopy analysis for major elements and radionuclides. Radionuclide extraction from the solid wastes in leaching is also indicated for each case in the attached Table 2. TABLE 1 __________________________________________________________________________ Liquor Compositions and Waste Stability, Illustrating the Process Disclosed. A1-6 A1-9 A2-1 A2-2 A2-3 A2-4 A3-1 A3-2 A3-4 A3-5 A1-2 __________________________________________________________________________ Liquor, g/L Fe 0.25 23.2 0.27 0.27 0.27 0.27 0.27 0.27 58.1 0.27 36.8 Zr 2.01 2.01 1.98 1.98 1.98 1.98 1.98 1.98 -- 1.98 0.73 Si 0.058 0.058 0.29 0.29 0.29 0.29 0.29 0.29 -- 0.29 -- Ti 0.064 0.064 0.064 0.064 0.064 0.064 0.064 0.064 -- 0.064 -- Y 0.172 0.172 -- 0.172 -- 0.172 0.172 0.172 0.172 0.172 -- Mg 0.169 0.169 -- -- 0.169 0.169 0.169 0.169 -- 0.169 -- Al 0.43 0.43 0.43 -- -- 0.43 0.43 0.43 -- 0.43 -- P <0.020 <0.020 -- -- -- -- 0.09 0.09 0.09 0.135 -- Ca 6.50 6.50 6.50 6.50 6.50 6.50 6.50 6.50 -- 6.50 4.69 Ce 0.01 0.01 -- -- -- -- -- 0.011 -- 0.011 -- Hf 0.062 0.062 0.062 0.062 0.062 0.062 0.062 0.062 -- 0.062 -- Cl 71.0 114.7 71.0 71.0 71.0 71.0 71.0 71.0 110.7 71.0 78.5 Na -- -- 0.42 0.42 0.42 0.42 0.42 0.42 -- 0.42 -- U -238 0.028 0.028 0.029 0.029 0.029 0.029 0.029 0.029 0.029 0.029 0.5 Th -232 0.070 0.070 0.070 0.070 0.070 0.070 0.070 0.070 0.070 0.070 0.5 Ra -226 400 400 400 400 400 400 400 400 400 400 6000 H.sub.2 SO.sub.4 14.2 14.2 15.7 15.7 15.7 15.7 15.7 15.7 15.7 15.7 10.9 Addition (g/l) Waste Leach Results Acetic 0 0 47.6 33.2 41.4 46 3.9 48.4 0 0 0 Acid Addition 0.5.M mL/L U Extraction 0 6.2 8.0 3.6 8.6 6.5 13.0 0 13.3 0 69 Th Extraction 0 3.6 6.1 0 3.8 0 16.0 0 14.6 0 0 % Ra Extraction 0-10 21 18 34 71 44 0 28 11 15 44 % __________________________________________________________________________ TABLE 2 ______________________________________ Liquor Compositions and Waste Stability ______________________________________ Liquor, g/L A4-1 A4-2 A4-3 ______________________________________ Fe 84.4 86.9 83.8 Zr 0.009 5.15 5.12 Si 0.023 0.028 0.028 Ti 0.177 0.171 0.150 Y 0.011 0.012 0.012 Mg 2.29 2.41 2.10 Al 0.146 0.175 2.70 P 0.097 1.38 2.65 Ca 0.110 0.115 0.116 Ce 0.048 0.158 0.168 Hf -- -- -- Cl n.d. n.d. n.d. Na 0.515 0.555 0.546 U -238 0.180 0.182 0.158 Th -232 0.102 0.106 0.090 Ra -226* H.sub.2 SO.sub.4 Addition (g/l) 0 0 0 Waste Leach Results Acetic Acid Addition 0.5 .M mL/L 0 5.2 5.0 U Extraction % 19.8 0.13 0.08 Th Extraction % 0.11 0 0 Ra Extraction % 3 7 4 ______________________________________ n.d. = not determined *in radiochemical equilibrium with uranium
	abstract	A collision-avoidance propulsion system and method for orbiting satellites and other spacecraft takes advantage of ambient cosmic rays in space to catalyze micro-fusion events via particle-target fusion and muon-catalyzed fusion processes, using the reaction products to produce thrust upon orbiting satellites and other spacecraft. A supply of deuterium-containing particle fuel material is propelled in a specified direction of the spacecraft in response to indication of a potential collision with another space object (e.g. orbiting debris). In one embodiment, this may be performed by propellant gas expelling the fuel material through conduits to specified ports on the exterior of the spacecraft. The propelled material interacts with the ambient cosmic rays and muon generated from those cosmic rays to induce micro-fusion. A portion of the energetic reaction products (e.g. alpha particles) are received upon the spacecraft to alter its trajectory in a manner that avoids the potential collision.
	description	This application claims the priority of Japanese application nos. JP 2004-006198, filed Jan. 14, 2004, and JP 2004-215095 filed Jul. 23, 2004, the disclosures of which are expressly incorporated by reference herein. This is a divisional application from U.S. application Ser. No. 11/033,953, filed Jan. 13, 2005. 1. Field of the Invention The present invention relates to a nuclear power plant and an operation method thereof, and particularly, to augmenting a power generation capacity. 2. Description of the Related Art In a conventional newly-constructed nuclear power plant, a power output is augmented by, for example, improving either a composition or a shape configuration of a fuel assembly, or the like, and by increasing a main steam flow rate at an outlet of a reactor. A technology of such a conventional example is disclosed in Japanese Patent Laid-Open Publication Hei. 9-264983. When applying the conventional technology described above to an existing nuclear power plant, the main steam flow rate increases substantially proportional to an increase of the power output. In order to suppress an increase of the main steam flow rate, a feedwater temperature may be lowered; however, if an extraction steam for heating the feedwater is simply decreased, thermal efficiency is extensively deteriorated and the power output hardly increases. This is not realistic option. Further, the increase of the main steam flow rate decreases a design margin of pressure vessel internals such as feedwater piping, a feedwater heater, a feedwater pump, and a steam dryer, and almost all power plant components, such as a main steam pipe, a high pressure turbine, a low pressure turbine, and a condenser. In a power plant using a normal boiling water reactor, the high pressure turbine is one of the components most likely to be the first to lose its design margin due to the increase of the main steam flow rate. Also in a nuclear power plant system other than a boiling water reactor, there is a similar problem with respect to a plant having a comparatively small design margin of the high pressure turbine, such that when applying a conventional technology to augment power output to an existing nuclear power plant, large scale improvement and change of the plant instruments is required. Consequently, there is a need for a nuclear power plant and operation method thereof that enable a power uprate of the plant without extensively changing a configuration of the plant, including its instruments. A first embodiment of the invention to solve the above problem is, after an operation cycle (i.e., a period from an activation of a nuclear power plant to an operation stop thereof for changing fuel), to augment a second reactor thermal power output in a second operation cycle to a level larger than a first reactor thermal power output in the previous operation cycle by decreasing a ratio of extraction steam which is led to a feedwater heater from a steam loop in the second operation cycle. A second embodiment of the invention to solve the above problem is, after an operation cycle, to augment a second reactor thermal power output in a second operation cycle to a level larger than a first reactor thermal power output in a previous operation cycle by decreasing a ratio of extraction steam which is led to a feedwater heater specifically from a middle area and an outlet of a high pressure turbine (the outlet steam extraction actually may be taken anywhere between the outlet of the high pressure turbine to any one of the inlets of a moisture separator, a moisture separator and heater, and a moisture separator and reheater). In addition, a third embodiment of the invention to solve the above problem is to augment a second reactor thermal power output in a second operation cycle of a reactor to a level larger than a first reactor thermal power output in a previous operation cycle by decreasing a mass flow rate of extraction steam led to a feedwater heater specifically from a middle area and outlet of a high pressure turbine out of extraction steam. In addition, a fourth embodiment of the invention to solve the above problem is to augment a second reactor thermal power output in a second operation cycle of a reactor to a level larger than a first reactor thermal power output in a previous operation cycle by decreasing a temperature rise amount at a high pressure feedwater heater placed downstream of a main feedwater pump. In addition, a fifth embodiment of the invention to solve the above problem is to augment a second reactor thermal power output in a second operation cycle of a reactor to a level larger than a first reactor thermal power output in a first operation cycle by stopping at least not less than one loop of an extraction steam pipe specifically from a middle area and outlet of a high pressure turbine. Here will be described an embodiment where the present invention is applied to the boiling water reactor of one of direct-cycle nuclear power plants. FIG. 1 shows a heat balance example of a boiling water reactor (BWR) after a power uprate according to the present invention, and FIG. 2 shows a heat balance example of the boiling water reactor before a power uprate. FIG. 3 shows a heat balance example of the boiling water reactor after a conventional power uprate. FIG. 4 shows an example for realizing a heat balance of the present invention shown in FIG. 1. Although in FIG. 1 an extraction steam amount is reduced by placing a valve at a middle area of an extraction pipe, the approach shown in FIG. 4 is applied when there is no space at the middle area of the extraction pipe and placement cost of the valve is high. In addition, each of FIGS. 5 and 6 show a conceptual drawing of an operation cycle of an embodiment of the present invention. In FIGS. 1, 2, and 3 reactor thermal power output is represented as Q, each mass flow rate of water and steam as G, and each enthalpy of water and steam as H. The reactor thermal power output Q and a mass flow rate G are expressed as ratios (%) based on their respective values at the reactor thermal power output of a reactor and a steam flow rate at an outlet of a reactor pressure vessel before a power uprate as shown in FIG. 2, and enthalpy is expressed in units of kJ/kg. In addition, each embodiment of the present invention shows a normal operation condition; operation conditions of an activation, stop time, transient state, and furthermore accident are excluded. This embodiment of the present invention is shown in FIG. 1, and the conceptual drawing of the operation cycles for complementing the embodiment is shown in FIG. 5. FIG. 1 is a drawing schematically showing the heat balance example in a case of performing the power uprate in a boiling water reactor that comprises a recirculation pump and a jet pump within a reactor pressure vessel 1, and has a main steam pipe 2, a high pressure turbine 3 and low pressure turbine 5 connected to the main steam pipe, a moisture separator 4 between the high pressure turbine and the low pressure turbine and a condenser 6 receiving steam from low pressure turbine 5. FIG. 5 contrasts relationships between an operation cycle and a reactor thermal power output, a main steam flow rate (steam flow amount flowing in the main steam pipe from the reactor pressure vessel), and an extraction steam amount together with a conventional power uprate method. One operation cycle is defined as a period from an activation out of a stop condition of a reactor operation to a stop thereof for a fuel change. In FIG. 5, an Nth operation cycle is shown before a power uprate method of the present invention is applied, and at this time the reactor thermal power output is Q=100%. A heat balance example before the power uprate is shown in FIG. 2. An (N+1)th operation cycle increases the reactor thermal power output by 5% and thereby makes Q=105%. An increase of the reactor thermal power output can be realized by any method, such as by: enlarging a pull-out amount of control rods in the (N+1)th cycle larger than in the Nth cycle; increasing a reactor core flow rate in the (N+1)th cycle larger than in the Nth cycle by increasing a rotation speed of the recirculation pump; and changing a kind of a fuel assembly. In addition, because applying the present invention results in lowering a temperature of feedwater supplied to the reactor pressure vessel, it can also be expected that the reactor thermal power output will naturally rise by coolant density feedback for core reactivity due to a lowering of the reactor-core-inlet coolant temperature. In some plants an extraction steam flow rate and main steam flow rate in one cycle are changed as shown in FIG. 6. In a case of a plant adopting the operation cycle as shown in FIG. 6, it is assumed that the heat balance, extraction steam flow rate, main steam flow rate, feedwater heating amount, and the like are compared at an operation point where the main steam flow rate becomes maximum in the operation cycle excluding transients, such as activation, stop, accident/transient phenomenon occurrence time, and test operation. When increasing the reactor thermal power output, it is necessary to increase a feedwater flow rate or to widen an enthalpy difference of a coolant between an inlet/outlet of the reactor pressure vessel in order to remove the additional increment of thermal energy from the reactor. The conventional power uprate method adopts the former method, increasing the feedwater flow rate in proportion to the reactor thermal power output. A heat balance example by the conventional power uprate method is shown in FIG. 3. As a result, in the conventional power uprate method the main steam flow rate of the (N+1)th operation cycle shown in FIG. 5 becomes 105%. The present invention adopts the latter method and is characterized by widening the enthalpy difference of the coolant between the inlet/outlet of the reactor pressure vessel by intentionally lowering a feedwater enthalpy at the inlet of the reactor pressure vessel. In order to lower the feedwater enthalpy at the inlet of the reactor pressure vessel, it is possible to decrease an extraction steam from a steam loop and thereby decrease a steam amount sent to feedwater heaters 7, 9. However, if only an extraction steam amount is decreased, thermal efficiency decreases and the total power generation increase is limited. Accordingly, by selectively decreasing an extraction steam amount from any of a middle area and outlet of the high pressure turbine (actually at any location from the outlet of the high pressure turbine and an inlet of the moisture separator), a steam amount flowing in the low pressure turbine is increased and thus the power generation amount is increased. Because most extraction steam from the middle area and outlet of the high pressure turbine is used at a feedwater heater downstream of a main feedwater pump 8, the power uprate method of the present invention may be viewed as a method of decreasing feedwater heating downstream of the main feedwater pump. In a case of a plant where an original extraction steam amount from the middle area and outlet of the high pressure turbine is little, in order to sufficiently decrease a feedwater temperature it may be necessary to also decrease an extraction steam amount extracted from the low pressure turbine in such a plant the extraction steam amount from the middle area and outlet of the high pressure turbine is decreased more, some extent of effect can be obtained. In the embodiment, in spite of increasing the reactor thermal power output by 5% compared to that of the Nth cycle, the main steam flow rate can be made same as that of the Nth cycle. The embodiment shows an ideal power uprate method in which the main steam flow rates of the Nth and (N+1)th operation cycles are assumed to be the same, however, they need not always be entirely the same and may be increased within a range of component design margin, for example, within the design margin of the high pressure turbine. When there are a plurality of extraction points at the middle area and outlet of the high pressure turbine, decreasing an extraction steam amount is most effective if the extraction point is selected at the most upstream side of the high pressure turbine. In this case although it is possible to place an extraction pipe flow rate adjustment valve 10 for controlling the extraction steam amount at this location, it is possible to completely close at least one extraction pipe. As a closing method, it is possible to place a shut-off valve in the extraction pipe or to plug the pipe. When an extraction pipe is completely closed, control loop instruments for monitoring the extraction steam amount become unnecessary and operation control is also simplified. Whether controlling the extraction steam amount or completely closing the extraction pipes is preferred depends on the heat balance and the power uprate range, for example, it may be necessary to be able to adjust the steam extraction amount if an extraction steam amount per extraction pipe is high and when the extraction pipes are completely closed, a feedwater temperature lowers too much. In addition, instead of placing a shut-off valve in an extraction pipe, a feedwater flow rate flowing in a feedwater heater may be decreased. This embodiment is shown in FIG. 4, in which a feedwater heater bypass loop 11 is placed in the feedwater piping, and a part of feedwater is made to flow in the bypass loop 11. A low temperature coolant flowing in the bypass loop 11 bypasses at least one feedwater heater and then mixes with high temperature main feedwater. Thus a lowering of a feedwater temperature can be realized at an inlet of the reactor pressure vessel. Because when augmenting the reactor thermal power output and increasing the power generation amount of a nuclear power plant, the embodiment can suppress an increase of a feedwater flow rate and a main steam flow rate, it can suppress an increase of a load on a feedwater pipe, main steam pipe, and pressure vessel internals. Compared to the case of simply decreasing the extraction steam amount, the present invention can suppress the lowering of the thermal efficiency and obtain a larger power output. In addition, although in an extensive power uprate by a conventional power uprate method it generally becomes necessary to change the high pressure turbine, with the present invention a power uprate range performable without a change of the high pressure turbine widens compared to the conventional method. Further, as the feedwater temperature lowers, a thermal margin (corresponding to an MCPR (Minimum Critical Power Ratio) in a case of the BWR) of a reactor core increases, there is also the benefit of an increase of a design margin compared to the conventional method. Although in a power uprate a pressure loss and stability of the reactor core deteriorates, in the power uprate method of the present invention a void fraction of the reactor core becomes lower and an absolute value of void coefficient of the reactor core becomes larger, and thus the pressure loss of the reactor core is reduced, and the deterioration of the stability of the reactor core is also suppressed. The decrease of the pressure loss of the reactor core means that an increase of a load on the jet pump and recirculation pump for recirculating a coolant by a power uprate can also be suppressed. Because an increase in the amount of generation steam in the reactor core also becomes small compared to the increase of the thermal power output, an increase of carry under that occurs due to a steam entrainment into recirculation water is also small, and even in an extensive power uprate, it becomes easy to ensure a flow window. A direct-cycle nuclear power plant other than the boiling water reactor may also have a power uprate by a similar method. Table 1 shows a relationship among a reactor thermal power output, main steam flow rate, extraction steam flow rate, and feedwater enthalpy when applying the power uprate method of the embodiment to various output increase amounts. The reactor thermal power output and the main steam flow rate show ratios in the case of a reactor thermal power output of 100%, and the extraction steam flow rate shows a ratio for the main steam flow rate in the case of the reactor thermal power output of 100%. As seen from Table 1, even when making the reactor thermal power output 110%, the power uprate method of the present invention is widely applicable. A reason why the output is not shown only until 110% in Table 1 is that in a higher power uprate a change of the moisture separator and the like becomes necessary; if the moisture separator is changed or combined with a reactor pressure increase, the power uprate method of the present invention is more extensively applicable. Generally in the boiling water reactor a reactor thermal power output may be increased to 102% solely by improving measurement accuracy of a feedwater flowmeter and the like. Therefore, the present invention has greater applicability to a power uprate in ranges above 102%. Furthermore, in the power uprate up to a reactor thermal power output of 105%, it is generally unnecessary to extensively change system plant components, such as a change of the high pressure turbine. Using the present invention, particularly a large effect can be obtained because the change of the high pressure turbine becomes unnecessary even in the power uprate exceeding the reactor thermal power output 105%. TABLE 1Reactor thermalMain SteamExtraction SteamFeedwaterpower output (%)Flow Rate (%)Flow Rate (%)Enthalpy (kJ/kg)100100459241031004386910510042831107100407951101003873911010542831 Next will be shown an embodiment of the present invention applied to a pressurized water reactor (PWR) of an indirect cycle nuclear power plant. FIG. 7 shows a heat balance example of the pressurized water reactor of the present embodiment after a power uprate, and FIG. 8 shows a heat balance example of the pressurized water reactor before a power uprate. FIG. 9 shows a heat balance of the pressurized water reactor after applying a conventional power uprate method. Each of FIGS. 5 and 6 shows the conceptual drawing of the operation cycle of one embodiment of the present invention. In FIGS. 7, 8, and 9 reactor thermal power output is represented as Q, each mass flow rate of water and steam as G, and each enthalpy of water and steam as H. The reactor thermal power output Q and a mass flow rate G are expressed as ratios (%) based on their respective values at the reactor thermal power output and steam flow rate (steam amount flowing in a secondary main steam pipe from a steam generator) of a reactor before a power uprate as shown in FIG. 8, and enthalpy is expressed in units of kJ/kg. A heat exchange amount at a steam generator is an amount where a heat leak in a primary loop is subtracted from a reactor thermal power output, and because a normal heat leak amount is sufficiently small compared to the reactor thermal power output, the heat exchange amount at the steam generator and the reactor thermal power output are assumed equal. This embodiment of the present invention is shown in FIG. 7, and the conceptual drawing of the operation cycle for the embodiment is shown in FIG. 5. FIG. 7 schematically shows a heat balance example in the pressurized water reactor that comprises a reactor pressure vessel 1, a steam generator 13 transferring heat generated at a reactor core within the reactor pressure vessel to a secondary loop, a main steam pipe 2 leading secondary loop steam going out of the steam generator, a high pressure turbine 3 and low pressure turbine 5 connected to the main steam pipe, a moisture separator and heater 12 between the high pressure turbine and the low pressure turbine, and a condenser 6 receiving steam from low pressure turbine 5. FIG. 5 contrasts relationships between an operation cycle and a reactor thermal power output, a main steam flow rate, and an extraction steam amount in a case of using the embodiment together with a conventional power uprate method. One operation cycle is defined as a period from a reactor activation to a reactor operation stop for a fuel change. In FIG. 5 an Nth operation cycle is shown before an power uprate method of the present invention is applied, and at this time the reactor thermal power output is Q=100%. A heat balance example before the power uprate is shown in FIG. 8. An (N+1)th operation cycle increases the reactor thermal power output by 5% and thus makes Q=105%. An increase of the reactor thermal power output can be realized by any method, such as by: enlarging a pull-out amount of control rods in the (N+1)th cycle larger than in the Nth cycle; and changing a kind of a fuel assembly. In some plants an extraction steam flow rate and main steam flow rate in one cycle are changed as shown in FIG. 6. In a case of a plant adopting the operation cycle as shown in FIG. 6, it is assumed that the heat balance, extraction steam flow rate, main steam flow rate, feedwater heating amount, and the like are compared at an operation point where the main steam flow rate becomes maximum in the operation cycle excluding transients, such as activation, stop, accident/transient phenomenon occurrence time, and test operation. When increasing the reactor thermal power output, it is necessary to increase a primary coolant flow rate into the reactor pressure vessel and a secondary feedwater flow rate into the steam generator, or to enlarge an enthalpy difference of a primary coolant between an inlet/outlet of the reactor pressure vessel and that of a secondary coolant between an inlet/outlet of the steam generator in order to remove the additional increment of thermal energy from the reactor. The conventional power uprate method adopts the former method, increasing the primary coolant flow rate and the secondary feedwater flow rate in proportion to the reactor thermal power output. A heat balance example by the conventional power uprate method is shown in FIG. 9. As a result, in the conventional power uprate method the main steam flow rate of the (N+1)th operation cycle shown in FIG. 5 becomes 105%. The present invention adopts the latter method and is characterized by enlarging the enthalpy difference of the secondary coolant between the inlet/outlet of the reactor pressure vessel with intentionally lowering a secondary feedwater enthalpy at the inlet of the steam generator. In order to lower the feedwater enthalpy at the inlet of the reactor pressure vessel, although it is possible to decrease an extraction steam from a steam loop and thereby to decrease a steam amount sent to the feedwater heaters 7, 9. However, if only an extraction steam amount is decreased, thermal efficiency decreases and the total power generation increase is limited. Accordingly, by selectively decreasing an extraction steam amount from any of a middle area and outlet of the high pressure turbine (actually at any location from the outlet of the high pressure turbine and an inlet of the moisture separator), a steam amount flowing in the low pressure turbine is increased and thus the power generation amount is increased. Because most extraction steam from the middle area and outlet of the high pressure turbine is used at a feedwater heater downstream of a main feedwater pump 8, the power uprate method of the present invention may be viewed as a method of decreasing feedwater heating downstream of the main feedwater pump. In a case of a plant where an original extraction steam amount from the middle area and outlet of the high pressure turbine is little, in order to sufficiently decrease a feedwater temperature it may be necessary to also decrease an extraction steam amount extracted from the low pressure turbine. If in such a plant the extraction steam amount from the middle area and outlet of the high pressure turbine is decreased more, some extent of effect can be obtained. In the embodiment, in spite of increasing the reactor thermal power output by 5% compared to that of the Nth cycle, the main steam flow rate can be made same as that of the Nth cycle. The embodiment shows an ideal power uprate method in which the main steam flow rates of the Nth and (N+1)th operation cycles are assumed to be the same, however, they need not always be entirely the same and may be increased within a range of component design margin, for example, within the design margin of the high pressure turbine. When there are a plurality of extraction points at the middle area and outlet of the high pressure turbine, decreasing an extraction steam amount is most effective if the extraction point is selected at the most upstream side of the high pressure turbine. In this case although it is possible to place an extraction pipe flow rate adjustment valve 10 for controlling the extraction steam amount at this location, it is possible to completely close at least one extraction pipe. As a closing method, it is possible to place a shut-off valve in the extraction pipe or to plug the pipe. When an extraction pipe is completely closed, control loop instruments for monitoring the extraction steam amount become unnecessary and operation control is also simplified. Whether controlling the extraction steam amount or completely closing the extraction pipes is preferred depends on the heat balance and the power uprate range, for example, it may be necessary to be able to adjust the steam extraction amount if an extraction steam amount per extraction pipe is high and when the extraction pipes are completely closed, a feedwater temperature lowers too much. In addition, instead of placing a shut-off valve in an extraction pipe, a feedwater flow rate flowing in a feedwater heater may be decreased. This embodiment is shown in FIG. 10, and it shows an example for realizing a heat balance of the present invention shown in FIG. 7. Although in FIG. 7 an extraction steam amount is reduced by placing a valve at a middle area of an extraction pipe, a method shown in FIG. 10 is applied when there is no space at the middle area of the extraction pipe and placement cost of the valve is high. In this embodiment a feedwater heater bypass loop 11 is placed in the feedwater piping, and a part of feedwater flow is made to flow in the bypass loop 11. A low temperature coolant flowing in the bypass loop 11 bypasses at least one feedwater heater and then mixes with high temperature main feedwater. Thus a lowering of a feedwater temperature can be realized at an inlet of the reactor pressure vessel. Because when augmenting the reactor thermal power output and increasing the power generation amount of a nuclear power plant, the embodiment can suppress an increase of a feedwater flow rate and a main steam flow rate, it can suppress an increase of a load on the feedwater pipe, main steam pipe, and steam generator. It is also possible to lower the reactor pressure vessel inlet temperature of a primary loop without increasing the primary coolant flow rate, and in this case it is more effective to suppress the increase of a load on the steam generator and a load on the primary coolant pump is also reduced. Furthermore, if the reactor pressure vessel inlet temperature of the primary loop lowers, a thermal margin (corresponding to a DNBR (Departure from Nucleate Boiling Ratio) in the case of the PWR) of a reactor core increases, there is also the benefit of an increase of a design margin compared to the conventional method. A indirect-cycle nuclear power plant other than the pressurized water reactor may also have a power uprate by a similar method. Thus, although the embodiments of the present invention are described, the invention is not limited thereto, and various variations are available without departing from the spirit and scope of the invention.
050248030	description	DESCRIPTION OF PREFERRED EMBODIMENT FIG. 1 shows schematically a portion 1 of the enclosure of a nuclear reactor, notably of the so-called pressurized water type, including thick concrete protection walls 2 defining an inner zone 3 filled with water. A portion of the reactor shell is shown at 4, reference 5 denoting a nuclear fuel assembly adapted to be introduced or removed from the reactor core. The reactor structure as well as the means used for loading or unloading the core with the fuel assembly are well known in the art and therefore require no particular description, said means moreover having no direct relation with the object of the present invention. In FIG. 1, a fuel assembly 5 shown in solid lines is mounted in a substantially horizontal position, carried by a conveying carrier 6 brought into the inner zone 3, in a space (I) of the latter which forms a conventional a loading-unloading station. The mechanism allowing gripping the fuel assembly 5 on its carrier, transferring it from its horizontal position to a vertical position, and then transferring it so as to place it in vertical alignment with the core above shell 4, is schematically denoted as a whole by reference numeral 7. The conveying carrier 6 moves on a rolling path 8 provided in the bottom of enclosure 1, so as to pass from space (I) and from the loading-unloading station to a space (II) situated on the other side of the protection wall 2 with respect to space (I) and forming a receiving or removal station for the fuel assemblies, to selectively allow the positioning on the carrier of a new fuel assembly to be introduced in the core, or reversely receiving and removing a used fuel assembly removed from said core. During its transfer from space (I) toward space (II) or vice versa, the conveying carrier 6 passes through a guiding tunnel 9 provided in a passage 10 formed in the lower portion of wall 2, whereby said tunnel can be opened or closed by the closing block 11 of a lock-gate mounted at the lower end of a control stem 12, operated from the outside of the enclosure by a control mechanism 13. Space (II), where the removal or receiving station is placed, is filled with water 14 to a height necessary to provide a convenient protection to the outer environment during handling of the fuel assemblies 5, and particularly those extracted from the core which are highly radio-active. The control of the reciprocating movements of conveying carrier 6 from one to the other of spaces (I) and (II) is provided by means of a handling chain 15 the details of which will be explained hereinafter. Moreover, one of the ends of carrier 6 is connected to a traction cable 16 allowing the carrier to be returned inside space (II) to the receiving or removal station when an incident occurs on handling chain 15. Cable 16 is returned on a guide pulley 17 toward a control winch 18 provided at the upper portion of enclosure 1. FIGS. 2 and 3 show in greater detail a particular embodiment of the conveying carrier 6 and of the fuel assembly 5 monted in the latter inside a carrying structure 19 (FIG. 2), provided with side trunnions 20 pivoting inside bearings 21 so as to allow assembly 5 to be lifted respect to the carrier for loading or unloading operations, briefly mentioned hereinabove. Carrier 6 includes a support chassis 22 provided laterally with wheels 23 for the displacement of the carrier on path 8, particularly on carrying walls 24 conveniently arranged for receiving and guiding said wheels. The operating chain 15 is a chain of the "push-pull" type and includes successive links (FIGS. 3 and 4) each formed with two parallel side flanges, respectively 25 and 26. Between the flanges of any one of the chain links, preferably at the end of the chain, a fastening member 27 is engaged underneath the chassis 22 of carrier 6, flanges 25 and 26 being rigidly connected to said fastening member 27 by a connection means the details of which will be explained hereinafter. Flanges 25 and 26 extend into tongues 25a and 25a through which extends a connection shaft 28, including substantially in its central portion a double flange 29 defining a groove in which the teeth 30 of a control wheel 31 are successively engaged. The latter is rigidly connected to a horizontal shaft 32 and rotates inside a casing 33 provided between the carrier walls 24 of the rolling path 8. At one end of shaft 32 is supported a bevel pinion 34 in mesh with a pinion of same profile 35 keyed at the end of a control shaft 36, driven by an appropriate motor reduction 37 (FIG. 3). Chain 15, the detailed description of which does not directly refer to the invention, allows, depending on the direction of rotation of wheel 31, displacement of the conveying carrier 6 to the right or to the left in FIG. 1, respectively in the direction of spaces (I) or (II). When the chain winds itself into a spiral, it pulls the carrier to the removal or receiving station in space (II); on the contrary, when the chain is unwound, it becomes rigid and forms a one arm jack of the carrier toward space (I) in the direction of the loading and unloading station. According to the invention, the connection between the fastening member 27 rigidly connected to the carrier and the end link of chain 15, notably with the side flanges 25 and 26 of said link, is provided by means of a transverse spindle 38 penetrating bores in register, respectively 39 and 40, formed in the flanges on the one hand and in the fastening member on the other hand, as shown in FIG. 4. Moreover, spindle 38 is secured against motion with respect to fastening member 27, and therefore flanges 25 and 26 and also chain 15, by means of a pin 41 extending through the fastening element and the spindle, parallel to the link flanges. Spindle 38 is mounted inside a first housing 42 of corresponding shape, provided inside a support block 43. Said spindle 38 includes on on one of its sides a rectilinear rack 44, the housing 42 being open laterally to allow the teeth of a pinion 45 mounted in the support block 43 to mesh with rack 44. Pinion 45 is rigidly connected to shaft 46 extending vertically underneath chassis 22 of the conveying carrier (FIG. 2), said pinion 45 being substantially horizontal when such conditions prevail. The support block 43 also includes a second housing 47 extending perpendicularly to the direction of the first housing 42 so as to intersect the latter. Inside the second housing 42 is mounted a second spindle 48 including, like the first one, a rectilinear rack 49 adapted to mesh with the teeth of pinion 45. As may be seen in FIGS. 4 and 5, the relative position of spindles 38 and 48 is determined by the construction in such manner that, depending on the direction of pinion 45, one of the spindles penetrates inside support block 43 in its associated housing while the other extends outside the block by passing through the end of its own housing, and vice versa. Spindle 48 is rigidly connected outside block 43 to a stem 50, ending into a whorl 51 on which is attached the end of cable 16, the connection between the whorl and the cable being effected by any appropriate means, e.g., by crimping, screwing or other clamping and connection means. Whorl 51 includes one end 52 adapted for come to bear against an abutment 53 formed in alignment underneath chassis 22, with the cable 16 extending freely through said abutment via a passage 54. FIG. 6 shows schematically the control means for cable 16. Said cable is first passed around pulley 17, the position of which is fixed at the bottom of enclosure 14 by a support bracket 55, then upwardly to a second pulley 56. After passing around the latter, cable 16 is tackled on a mobile pulley 57 supporting a counter-weight 57a which exerts force on the cable maintaining it permanently taut. Beyond pulley 57, the cable winds itself on drum 58 of winch 18, controlled by a crank 59 or any equivalent driving means. The operation of the device according to the invention is easily understood from the explanations already given. Under normal operation conditions, the movements of carrier 6 between its two spaces (I) and (II), i.e., the loading and unloading station on the one hand, and the removal-receiving station on the other hand, are provided by means of the push-pull chain 15 which, according to the direction of its winding, causes the displacement of the carrier on its rolling path 8 on one side or the other. When an incident occurs and notably when there is a jamming or blockage of carrier 6 or of chain 15, it is of course important to be able to bring the carrier back rapidly to space (II) so as to be able to proceed, under acceptable conditions, to the necessary repairs and to the immediate removal of the assembly it supports, particularly if it is an irradiated assembly removed from the core. To the effect, this emergency device according to the invention consists in controlling, via winch 18, cable 16 in the direction where the latter exterts a traction force on spindle 48 engaged in its housing 47, the second spindle 38 being in that case in the position shown in FIG. 4 where it rigidly connects fastening member 27 flanges 25 and 26 of the end link of chain 15. The force thus exerted on the cable, transmitted from spindle 48 to spindle 38 via pinion 45, is applied to the connection pin 41 which, when judiciously calculated, can also abruptly break into three fragments, 41a and 41b remaining immobile inside fastening member 27, and 41c captive inside spindle 38 and permitting the removal of said spindle and simultaneously the progressive exit of the other spindle 48, respectively in and out of said block 43. After being broken, fastening member 27 is freed from chain 15, carrier 6 which is rigidly connected to the fastening member being no longer connected to the chain. As the force is still applied on cable 16, whorl 51 comes to bear via its end 52 on abutment 53 provided on chassis 22. At that moment, cable 16 can itself entrain the carrier and bring it back from any position of the latter on its rolling path 8 toward the receiving station (II) in the outer portion of enclosure 3. It should be noted that the whole of these operations is controlled remotely by a single operator, controlling only the drum 58 of winch 18. The contact established between whorl 51 and abutment 53 triggers by any known means a signal which is a confirmation for the operator to start driving the carrier and retracting it, until it reaches its final position. The invention thus provides a simple and particularly safe device enabling, when an incident occurs on the carrier handling chain, the return of the latter to a position which allows recovery of the fuel assembly which it supports. Cable 16 assumes under any circumstances the function imparted to it, which is the return of the carrier, particularly by using a tackle and counter-weight maintaining it permanently in a taut state. Of course, it goes without saying that the system according to the invention for the return, should an incident occur, of any remotely controlled mobile element, particularly of the freight-elevator type, or a loading member for a thermal treatment oven, a rolling mill, etc.
059873990	abstract	A method and apparatus for monitoring a source of data for determining an operating state of a working system. The method includes determining a sensor (or source of data) arrangement associated with monitoring the source of data for a system, activating a method for performing a sequential probability ratio test if the data source includes a single data (sensor) source, activating a second method for performing a regression sequential possibility ratio testing procedure if the arrangement includes a pair of sensors (data sources) with signals which are linearly or non-linearly related; activating a third method for performing a bounded angle ratio test procedure if the sensor arrangement includes multiple sensors and utilizing at least one of the first, second and third methods to accumulate sensor signals and determining the operating state of the system.
	description	The present invention relates to an X-ray focusing device that is used for focusing X-rays in various apparatuses that use X-rays such as electron probe micro-analyzer (EPMA), scanning electron microscope (SEM), transmission electron microscope (TEM), X-ray fluorescence spectrometer, XRD, X-ray CT and medical X-ray devices. With micro-area X-ray fluorescence spectrometers that are used for performing component analysis on a micro-area of a specimen, X-rays that are emitted from an X-ray source must be focused to a very small diameter and irradiated onto the specimen. With the micro-area X-ray fluorescence spectrometer that is described in Non-Patent Literature 1, multi-capillary (the term used in the literature is “polycapillary” but the more commonly used term “multi-capillary” is used in this specification) X-ray lens is used. The multi-capillary X-ray lens (“MCX”) is briefly explained next (see Patent Literature 1 and 2, etc.). FIG. 5 shows one mode of a MCX. FIG. 6 shows the principle behind the transmission of X-rays with a MCX. The basic construction of a MCX consists of numerous (approximately several hundred to a million) capillaries that are bundled together, each capillary being made of borosilicate glass and having a very small inner diameter in the range of approximately 2 μm to a dozen μm or so. As FIG. 6 shows, an X-ray beam that enters into a capillary 32 advances through the capillary while engaging in total reflection off the inner wall surface of the glass wall at an angle less than the critical angle. This principle is used to efficiently guide an X-ray. An X-ray can be efficiently guided whether the capillary 32 is linear-shaped such as that shown in FIG. 6(a) or bow-shaped such as that shown in FIG. 6(b). There are many types of MCX. FIG. 5(a) shows a point/point type MCX 30 wherein X-rays that are emitted from an X-ray source that can be considered to be substantially a point are collected at the incident-side end face with a large solid angle and X-rays that are emitted from the emission-side end face on the opposite side is focused to a single point. With the MCX shown in FIG. 5(b), X-rays that are emitted from an X-ray source that similarly can be considered to be substantially a point are collected at the incident-side end face having a large solid angle and the X-rays are emitted as parallel beams from the emission-side end face. The MCX shown in FIG. 5(b) can also be a point/parallel type MCX 31 where the direction of travel is reversed. Because, as afore-described, MCX is capable of collecting and guiding X-rays with a high efficiency, it is capable of irradiating a specimen with an X-ray having a high energy density and is therefore very effective in increasing the analysis sensitivity. On the other hand, it is not always very capable of focusing the X-rays, which have been collected with a high efficiency, onto a small irradiation area. One of the major reasons for this is that MCX, by its very principle of operation, causes blurring of the focal point. To explain, as shown in FIG. 7, because the X-ray travels through one capillary 32 while engaging in total reflection off the inner wall surface, the maximum reflection angle is the critical angle. For that reason, when the X-ray is emitted from the end face of capillary 32, the X-ray will have a divergence angle with respect to the optical axis (the center line of capillary 32) S with the maximum divergence angle being the critical angle θ. As a result, as shown in FIG. 8, the irradiation area of the X-ray that emerges from point-focus side end face 33 of MCX does not form an ideal point and instead forms an area 34 having a certain size. Furthermore, even if the X-ray that emerges from the end face of a single-capillary 32 is made to be non-diverging, because of the limitations with the manufacturing of MCX, it is practically speaking impossible to cause all of the optical axes of a vast number of capillaries to perfectly focus to a single point. This factor also becomes a cause for the blurring of the focal point. Because of the combination of such theoretical factors and the manufacturing limitations, the minimum focal point size of previous MCX has been limited to at most about 20 to 30 μm, and achieving any reduction in focal point size has been difficult. For example, with the device that is described in Non-Patent Literature 1, the size of the micro-area where the X-ray is irradiated is about 50 μm. In recent years, there has been a strong need with analytic instruments such as micro-area X-ray fluorescence spectrometer and the like to perform measurements of components that are present in minute quantities in micro-areas. In response to such need, novel X-ray focusing devices that reduce the size of irradiation diameter of X-rays have been proposed. Patent Literature 3 combines MCX with a focusing member having a truncated cone shape, and Patent Literature 4 combines MCX with a Fresnel zone plate (FZP). Even though it is possible with these configurations to reduce the X-ray irradiation diameter to less than that achieved with MCX alone, the configuration of Patent Literature 3 has a tendency to reduce the intensity of the X-rays in the irradiated areas and is disadvantageous in terms of sensitivity, and the configuration of Patent Literature 4 has a cost disadvantage because of the very expensive cost of FZP required for obtaining a sufficient level of performance. So, both methods have their advantages and disadvantages. Patent Literature 1: Examined Patent Application Publication No. H07-11600 Patent Literature 2: Examined Patent Application Publication No. H07-40080 Patent Literature 3: Unexamined Patent Application Publication No. 2007-93315 Patent Literature 4: Unexamined Patent Application Publication No. 2007-93316 Patent Literature 5: Unexamined Patent Application Publication No. 2007-225314 Non-Patent Literature 1: “Energy Dispersive Micro X-ray Fluorescence Spectrometer μEDX Series,” Online, Shimadzu Corporation, searched Oct. 15, 2009, Internet, URL: http://www.shimadzu.co.jp/surface/products/m_edx/index.html The present invention was made to solve the afore-described problems, and it is the object of the present invention to provide an X-ray focusing device that can focus X-rays to a very small diameter while, at the same time, securing a high X-ray intensity in the X-ray irradiated area and providing cost advantages. As afore-described, MCX is advantageous in efficiently collecting X-rays that are emitted from an X-ray source and increasing the energy density of the X-rays at the irradiated area, but is limited in the ability to reduce the X-ray irradiation diameter. As a way of taking advantage of the afore-described advantages of MCX while compensating for its disadvantages, the inventors of the present application focused on a single-capillary X-ray lens (“SCX”) as an X-ray optical device whose properties are different from (opposite of) those of MCX. As its name literally states, SCX uses only one capillary. As FIG. 9 shows, an X-ray that is introduced into the interior of one glass capillary 40 is focused as the X-ray reflects off the inner wall surface of the glass capillary 40 once or a plurality of times at an angle less than the critical angle. The X-ray that emerges from the tapered end face 41 at the tip can be formed to have a very small focal point with a diameter of 10 μm or less. The afore-described SCX is advantages in terms of reducing the irradiation area of the X-ray while having a low cost because of its relatively easy manufacturing. At the same time, however, because the diameter of the X-ray incident-side end face cannot be made large, the incident efficiency of the X-ray is poor. This results in a low energy density of the X-ray irradiated area. Another way of stating this is that the advantages and disadvantages of SCX and MCX are the exact opposites. The inventors of the present application realized that by suitably combining the two, the advantages of either can be brought to the fore while compensating for the disadvantages, and that an X-ray focusing device with superior performance but a low cost can be realized. The X-ray focusing device according to the present invention, which was invented, for solving the afore-described problems includes: a multi-capillary including a plurality of bundled capillaries for guiding X-rays and whose, at least, one end face is a converging end for concentratedly irradiating X-ray to a micro-area located outside of the end face; and a single-capillary including one capillary for guiding X-rays and whose, at least, one end face is a converging end for irradiating X-rays to a micro-area located outside of the end face and whose other end face is a long-focal length converging end or a parallel end capable of accepting parallel X-ray beams; wherein the parallel end or the long-focal length converging end of the single-capillary is positioned outside the converging end of the multi-capillary, and the multi-capillary and the single-capillary are positioned so that the optical axis of the multi-capillary at the converging end coincides with the optical axis of the single-capillary at the parallel end or the long-focal length converging end. With the X-ray focusing device according to the present invention, one end face of the multi-capillary is a converging end but the other end face may either be a converging end or a parallel end. With the X-ray focusing device according to the present invention, the X-ray that has been efficiently guided through each capillary of the multi-capillary is emitted from the converging end and forms a focal point whose area size is relatively large. As one desirable mode of the present invention, the inner diameter (diameter of the area that can accept X-ray) of the incident end face at the converging end with a long-focal length or the parallel end is made larger than the diameter of the X-ray irradiated area that is formed at the focal point at the converging end of the afore-described multi-capillary, and the position of the multi-capillary and the single-capillary is set so that the incident end face of the single-capillary is positioned near the position of the focal point. As afore-described, the size of the focal point outside of the converging end of the multi-capillary is large, but the X-ray that is emitted from the converging end, when viewed from the incident end face of the single-capillary, can be deemed as a light source that gradually joins the focal point or as an approximately parallel light source. Because of this, the X-ray that is emitted from the converging end of the multi-capillary is efficiently taken into the single-capillary. The X-ray is then focused onto a very small diameter by the single-capillary and is emitted from its converging end to irradiate a very small area in a concentrated manner. Ignoring the loss in X-ray as it passes through multi-capillaries or a single-capillary and the loss in X-ray as the X-ray that is emitted from multi-capillaries becomes incident to the single-capillary, since numerous X-ray beams that were introduced into the multi-capillaries are ultimately irradiated onto a very small area from the converging end of the single-capillary, the X-ray energy density at the irradiated area becomes extremely high. Needless to say, the loss in X-ray during transit cannot be reduced to zero, but in the afore-described mode, since the loss in X-rays as the X-rays that are emitted from the multi-capillaries becomes incident to the single-capillary can be kept low, the final energy density at the X-ray irradiated area is kept sufficiently high. With the X-ray focusing device according to the present invention, the X-ray that is emitted from an X-ray source is efficiently collected by multi-capillaries, thus increasing the X-ray intensity. The X-ray is then irradiated onto a very small area in a concentrated manner by a single-capillary. By so doing, the area of the X-ray irradiated spot is made much smaller as compared to an ordinary MCX, and at the same time, even if the same X-ray source were to be used, the X-ray energy density at the X-ray irradiated area is made significantly larger as compared to before. This allows information that is obtained by the interaction (transmission, reflection, absorption, etc.) between the X-ray and the substances that exist at the micro-area to be detected with high sensitivity and accuracy. Furthermore, because a single-capillary can be manufactured more easily and inexpensively as compared to a multi-capillary, the X-ray focusing device according to the present invention, which combines these components, is not that much more expensive as compared to a multi-capillary X-ray lens alone, thus providing a X-ray focusing device of a high performance yet low cost. One embodiment of an X-ray focusing device according to the present invention is described next with reference to the attached drawings. FIG. 1 shows the configuration of the major elements of the present embodiment of an X-ray focusing device according to the present invention. FIG. 2 shows a schematic view of the configuration of an X-ray inspection device using the present embodiment of the X-ray focusing device. FIG. 3 is a schematic view showing the effects of the present embodiment of the X-ray focusing device. The present embodiment of the X-ray focusing device 1 comprises a multi-capillary X-ray lens (MCX) 2 and a single-capillary X-ray lens (SCX) 3. MCX 2 has a point/parallel type structure with its one end being a converging end 2b having a point focus that can be considered to be a single point (which, in fact, as described later, is a large size) if one were to assume that light that is emitted from each of the capillaries does not diverge after their emission. Its other end is a parallel end 2a. SCX 3 has a point/parallel type structure with its one end being a parallel end 3a with a substantially tubular shape and its other end being a converging end 3b with a tapered tip. The end face of the converging end 2b of MCX 2 and the end face of the parallel end 3a of SCX 3 oppose each other and are separated by distance L1. The distance L1 is equal to the distance from the end face of converging end 2b of MCX 2 to the focal point that is formed outside the end face of converging end 2b, i.e., the distance L1 is equal to the focal distance. The optical axis C2 at the converging end 2b of MCX 2 coincides with optical axis C3 of the parallel end 3a of SCX 3. Hence, the focal point of the X-ray that is emitted from the converging end 2b of MCX 2 is situated on the end face of the parallel end 3a of SCX 3. The diameter of the X-ray irradiated area for MCX 2, which is minimum at that position, is about several dozen μm to about 100 μm. On the other hand, the diameter φD3 of the area that can accept X-ray at the end face of the parallel end 3a of SCX 3 is usually about 0.1 mm to 1 mm, which is larger than the afore-described diameter of the X-ray irradiated area. This means that all of the X-ray that is emitted from the converging end 2b of MCX 2 become incident to the X-ray acceptable area on the end face of the parallel end 3a of SCX 3. Now, assuming that the loss in X-ray while passing through MCX 2 can be ignored and that the inner diameter of the parallel end 2a is φD1 and the inner diameter of the converging end 2b is φD2, the energy density of the X-ray that is emitted from the converging end 2b is going to be approximately φD22/φD12 times greater than the energy density of the X-ray that is introduced into the parallel end 2a. (Here, the thickness of the walls separating the adjacent capillaries is ignored.) For example, if φD1=3 mm and φD2=0.1 mm, the X-ray energy density will be about 900 times greater. Again assuming that all of the X-ray that is emitted from the converging end 2b of MCX 2 is all incorporated into SCX 3 and that any loss in X-ray during passage through SCX 3 can be ignored, and letting φD4 represent the irradiation diameter at focal point F of the X-ray that is emitted from the converging end 3b of SCX 3, the energy density of the X-ray at focal point F becomes approximately φD42/φD22 times greater. For example, if φD4=10 μm, the final energy density of the X-ray of the X-ray focusing device 1 at focal point F becomes approximately 100 times greater. In other words, the energy density of the X-ray at focal point F is going to be 90,000 times greater than the energy density of the X-ray that was initially incident on MCX 2. In actuality, the loss in X-ray as the X-ray is guided through MCX 2 and SCX 3 is not zero. Also, some of the X-ray that is introduced into SCX 3 through the end face of the parallel end 3a will exceed the critical angle for a total reflection on the inner wall surface of SCX 3, and such X-ray will not be used (will be lost). These factors mean that the actual increase in energy density of the X-ray will be less than the aforesaid approximations, but nevertheless, the X-ray energy density at focal point F will be dramatically higher than the case with MCX alone. In general, the transmittance of MCX or SCX when loss is accounted for is said to be about 30%. When this factor is accounted for, the aforesaid increase of 90,000-fold drops to about 9,000-fold, but this is still a very large effect. The afore-described operation and effect can be easily understood based on FIG. 3. FIG. 3 is a graph that plots along the horizontal axis the divergence in the horizontal direction of the irradiated X-ray at focal point F, and plots along the vertical axis the light quantum count (i.e., the X-ray intensity). The area of the region bounded by the curve such as that shown in the graph represents the total number of light quantum of all irradiated X-ray. To explain, when an MCX alone is used, as shown by curve A in the graph, for reasons already described, the irradiated X-ray cannot be focused very much, resulting in the divergence of the irradiated X-ray to be relatively large (minimum of about 20 to 30 μm). In contrast to this, with the present embodiment of X-ray focusing device 1, the same property is represented by curve B in the figure, showing that the irradiated X-ray can be narrowly focused as compared to previous. Furthermore, the X-ray intensity in the irradiated range is quite high. As shown in FIG. 2, with an X-ray inspection device that employs the present embodiment of the X-ray focusing device 1, an X-ray focusing device 1 is installed between an X-ray source 12 and an inspected object 11 that moves on a manufacturing line 10. Primary X-ray that is emitted from the X-ray source 12 is efficiently focused to a small diameter by the X-ray focusing device 1 and is irradiated onto the inspected object 11. The secondary X-ray that is released from the inspected object 11 is detected by the X-ray detector 13, and information (such as an image) from the X-ray irradiation site on the inspected object 11 is obtained based on the detection signal. It is certainly acceptable to install a MCX on the detection side. As afore-described, with the present embodiment of the X-ray focusing device, MCX 2 is used to efficiently collect the X-ray and to narrow the irradiation diameter of the X-ray to a certain extent. The X-ray is then introduced into SCX 3 without waste where the X-ray is further focused so that the X-ray is irradiated onto a very small area on, for example, inspected object 11. By so doing, even though the intensity of the X-ray that is generated by X-ray source 12 may not be that high, an X-ray of a strong intensity can be irradiated onto a micro-area, allowing information on components that are present at that area to be acquired with a high sensitivity. The X-ray incident end portion of MCX 2 in the afore-described embodiment was a parallel end. This is effective when the X-ray source has a size that is greater than a certain value. If the X-ray source is of the size that allows it to be considered as substantially being a single point and if the X-ray is radially emitted from there, it is acceptable to use a MCX whose X-ray incident end part is a converging end with a point focus. Stated otherwise, MCX 2 that is used here can either be a point/parallel type or a point/point type. FIG. 4 shows the configuration of the major elements of a variation of the X-ray focusing device 1′ according to the present invention. Here, MCX 2 is the same as that in the embodiment shown in FIG. 1, but the shape of SCX 4 is different. To explain, with SCX 4, both ends are spheroid-shaped converging ends. However, the end portion 4a that opposes the converging end 2b of MCX 2 is a spheroid-shaped converging end with a long-focal length. The end part 4b at the side where the X-ray is irradiated externally is a spheroid-shaped converging end of a short focal length. The X-ray that is emitted from the converging end 2b of MCX 2 proceeds towards the focal point, and the X-ray diverges once the focal point is passed. However, as stated earlier, the size of the focal point is relatively large. If the focal point on the output side of MCX 2 is situated near the spheroid focal point of the spheroid-shaped converging end 4a of SCX 4 with a long-focal length, the X-ray is efficiently incorporated into SCX 4. The X-ray undergoes total reflection inside SCX 4 and is irradiated in a concentrated manner from spheroid-shaped converging end 4b having a short focal length onto a very small focal point F. In this way, with the X-ray focusing device according to the present invention, the SCX that is combined with the MCX need not necessarily be a parallel/point type and can also be a point/point type. Furthermore, the afore-described embodiments are just examples of the present invention, and needless to say, various modifications, changes and additions can be made within the scope of the thrust of the present invention and still be included within the scope of the claims. 1. X-ray focusing device 2. Multi-capillary X-ray lens (MCX) 2a. Parallel end 2b. Converging end 3, 4, Single capillary X-ray lens (SCX) 3a. Parallel end 3b, 4a, 4b. Converging end
	abstract	A protective screen for the screening off of a suction space and a suction duct connected to it in a cooling system, include at least one screen wall element which has a suction side and an outflow side. The screen wall element is built up from a plurality of modular rectangular cassette units, which respectively contain a plurality of suction pockets open towards the suction side, with the screen pockets being surrounded by outflow gaps which are open towards the outflow side.
	description	This application is a Divisional of U.S. Non-Provisional patent application Ser. No. 14/061,345 entitled SYSTEM AND METHOD FOR THE CAPTURE AND STORAGE OF GASES (corrected title is SYSTEM AND METHOD FOR THE CAPTURE AND STORAGE OF WASTE), filed Oct. 23, 2013, which claims benefit of U.S. Provisional Patent Application No. 61/717,209 entitled “ONE-STEP CAPTURE AND STORAGE OF VOLATILE FISSION GASES”, filed Oct. 23, 2012, the entirety of which are incorporated herein by reference. The United States Government has rights in this invention pursuant to Contract No. DE-AC04-94AL85000 between the United States Department of Energy and Sandia Corporation, for the operation of the Sandia National Laboratories. The present disclosure is generally directed to systems and methods for waste capture and storage, and is more particularly directed to systems and methods for capturing and storing hazardous waste gases with metal-organic frameworks (MOFs). The capture and storage (CSS) of waste materials continues to be a problem is many industries. In the nuclear industry, the capture and storage of fission gases is important in both the disposal of large quantities of high level radioactive wastes generated in the reprocessing of spent power reactor fuel and from nuclear reactor accidents. It is generally accepted that the most promising approach is to convert these radioactive wastes to a dry solid form which would render such wastes chemically, thermally and radioactively stable. This problem of dry solid stability is closely related to the safety of human life on earth for a period of over 20,000 years. For example, radioactive wastes contain isotopes including 129I, 90Sr, 240Pu, and 137Cs, whose half-lives are >15 million years, 29 years, 66,000 years, and 30 years respectively. These isotopes alone pose a significant threat to life and must be put into dry, solid forms that are stable for thousands of years. The solid radioactive waste form must be able to keep the radioactive isotopes immobilized for this length of time, preferably even in the presence of a water environment. For radioactive gases, one present immobilization route is the so-called dry solids approach which involves the method of fixation of waste materials in glasses via melting glass procedures. This approach offers some improvement regarding isolation and decrease in the rate of release of radioactive elements when the outer envelopes or containers are destroyed. However, standard nuclear waste glasses (such as borosilicate glass) glasses with high chemical durability and low alkali ion conductivities are melted at very high temperatures, e.g., 1800° C. and higher. Such high melting processes are economically unsound and moreover, cause a dangerous problem due to the volatilization of pernicious radioactive materials. Additionally, none or very small amounts of gaseous radioactive materials are further trapped in case of volatilization during glass formation. In nuclear power accident clean-up, the removal of radioactive Cs and/or Sr from seawater or containment fluids is a critical issue. Other scenarios include the removal of uranium (U) based compounds and ions from water systems in case of accidents of for reuse, and/or in the act of resource extraction (in a sense ocean water “mining”). In the power generation industry, and in particular coal fired power generation, the generation and release into the atmosphere of large quantities of CO2 remains an area of concern. A number of CO2 CCS technologies have been developed. One method for CCS uses metal oxide solutions, such as potassium or magnesium oxide, to remove CO2 from flue gas or other CO2 containing vent gases. The general capture mechanism involves reaction of the metal oxides with CO2 to form metal carbonates. These carbonate salts can either be land-filled, or be regenerated via oxidation to form a concentrated CO2 stream that can be compressed and injected into geological formations for storage. Present methods of gas capture, and in particular, waste gas capture, lack the ability to store the captured gas in a stable form for a lengthy period of time. In particular, present day storage does not provide sufficient isolation and immobilization of such waste material, sufficient long-term resistance to chemical attack by the surroundings, and sufficient stability at high temperature. In addition, a process for fixating radioactive materials inside a dry solid form having high resistance to leaching and other forms of chemical attack would not only be suitable for the containment, interim storage and possible eventual disposal of radioactive nuclear wastes. The need remains, therefore, for a method and system for the capture and storage of gases that provides a high degree of stable containment. The need also remains for a method of forming waste forms below the volatilization temperature of the captured waste. The present invention is directed to novel systems and methods for the containment of waste. The waste may be a waste gas or an element or molecule in solution. The methods of the invention do not involve any steps which would expose material to temperatures above room temperature (ambient temperature), thereby eliminating the environmental hazard due to volatilization of the captured material into the atmosphere. In addition, there are provided novel systems and methods for the fixation and immobilization of gases and ionic wastes such as, but not limited to the neutral and radioactive forms of I, H, CO2, Kr, Xe, Ra, Cs, Ba, Y, Sr, and Rb. According to an embodiment, a modified metal-organic-framework (MOF) material is disclosed that includes a waste material adsorbed in the MOF material. The MOF material is modified by converting the crystalline structure of the MOF to an amorphous structure. The waste material may be an element, molecule, or ion in gas or liquid phase. According to another embodiment of the invention, a method of forming a waste storage material is disclosed that includes providing a MOF material having a crystalline structure, adsorbing a waste material into pores of the metal-organic framework material, and applying pressure to the metal-organic framework material to convert the crystalline structure of the metal-organic framework to an amorphous structure. According to another embodiment of the present invention, a method of remediating waste is disclosed that includes adsorbing a waste material onto a metal-organic framework material, and applying pressure to the metal-organic framework material to convert the crystalline structure of the metal-organic framework to an amorphous structure One advantage of the present disclosure is to provide a waste form that has improved waste confinement. Another advantage of the present disclosure is to provide a waste form that physically entraps the gas or molecule with minimal processing including no heating, and non-specialized mechanical pressed amorphization. Other features and advantages of the present disclosure will be apparent from the following more detailed description of the preferred embodiment, taken in conjunction with the accompanying drawings which illustrate, by way of example, the principles of the disclosure. Wherever possible, the same reference numbers will be used throughout the drawings to represent the same parts. The present invention now will be described more fully hereinafter with reference to the accompanying drawings, in which preferred embodiments of the invention are shown. This invention may, however, 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. The present invention is directed to systems and methods for capturing waste, including gases and elements, ions or molecules in solution. The systems and methods provide for a high level of confinement and long term stability. The systems and methods include adsorbing waste into a metal-organic framework (MOF), and applying pressure to the MOF material's crystalline framework to change the framework to an amorphous MOF. This change or modification of the MOF material changes the sorption characteristics without collapsing the MOF cage framework. In such a manner, the modified MOF has improved adsorption retention, reduced desorption kinetics. The modification results in a more compacted, less porous, MOF material with lowered possibility of desorption of guest material. Additionally, modified MOF material, for example modified MOF adsorbed radioactive contaminated I2 are less powdery and have higher resistance to flow compared to unmodified MOFs. The modified MOF may then be further treated/and or contained. The waste may be a gas or an element or molecule in aqueous solution. The captured waste may be referred to as a “guest.” In an embodiment, the waste may be stable or radioactive isotope form of a light gas. In an embodiment, the light gas may be a stable or radioactive isotope form of I, H, CO, CO2, Kr, Xe, Ra, Cs, Ba, Y, Sr, and Rb; also including CH4 and H2O. In an embodiment, the radioactive gas is selected from the group including 129I, 131I, 3H, 14CO2, 85Kr, 133Xe, 90Sr, 135Cs and 137Cs, plus the other isotopes of each of these radioelements. In an embodiment, the waste may be an element or molecule in solution. In an embodiment, the waste may be radioactive cesium in seawater. In another embodiment, the waste may be radioactive uranyl-containing ions in seawater. The MOF material has an open structure through which the target molecules can diffuse and be adsorbed for capture and storage. MOFs are crystalline framework structures with metal clusters interconnected by organic linker groups, a design that endows the materials with large pores, open channels, and huge internal surface areas for adsorbing molecules. A particular MOF may be selected based on the size of the pore required to receive the target species or guest. The pore size may be selected by the smallest dimension of the molecule that can fit into the pore opening of the MOF and may be aided by temperature and pressure of operating system. In this invention, MOFs can be 3D (cage) or 2D (layered) framework types. MOFs are highly porous crystalline materials, with a very diverse structural and chemical profile. A large set of metal and organic linkers are available. As such, MOFs which can be used in this invention can be categorized following several criteria, including topology (ex. MOFs with zeolitic topologies: zeolitic imidazolate frameworks (ZIFs), zeolite-like metal-organic frameworks (ZMOFs), or based on the organic linkers they include: carboxylate-based MOFs, phosphonate-based MOFs, N-based linker MOFs, N—O-heterofunctional linkers based MOFs. In an embodiment, the MOF may be selected from the group including, but not limited to IRMOFs series (MOF-5), MOF-74 series, Sandia Metal-Organic Frameworks (SMOFs) series, and ZIFs series. ZIFs are a type of MOF framework. They are generally built from tetrahedral metal nodes and imidazolate ligands and form analogs to zeolite mineral structures. In an embodiment, the ZIF may be selected from the group including, but not limited to ZIF-6, ZIF-8, ZIF-10, ZIF-11. The framework of ZIF-8 has a chemical composition of ZnL2 (wherein L=2-Methylimidazolate, i.e., the anion of 2-Methylimidazole) and a topology defined by the Zn cations that is identical to the zeolitic framework type SOD. SOD is a three letter framework type code for a sodalite structure type, as defined by the International Zeolite Association (“IZA”) in the “Atlas of Zeolite Framework Types” (Ch. Baerlocher, L. B. McCusker, D. H. Olson, Sixth Revised Edition, Elsevier Amsterdam, 2007). In an embodiment, the MOF may be ZIF-4, ZIF-5, or ZIF-8. In another embodiment, the crystalline MOF may be a highly porous coordination polymer, HKUST-1. The framework of HKUST-1 is [Cu3(benzene-1,3,5-carboxylate)2. It has interconnected [Cu2(O2CR)4] units (where R is an aromatic ring), which create a three-dimensional system of channels with a pore size of 1 nanometer and an accessible porosity of about 40 percent in the solid. In an embodiment the MOF material may be treated by the addition of a metal ion or gas that reacts to form nanoparticles within the pores of the MOF. Those nanoparticles will participate in gas recovery by reaction with the gas as it enters the MOF pore. In an embodiment, the reactive substance may be silver or palladium. Silver provides a reactive nanoparticle component to selectively adsorb I2 gas molecules, resulting in containment of the radioactive iodine. The iodine loaded silver containing MOF is then treated to for a long term storage waste form. In another embodiment, the reactive particle may be Pd, which may be used to enhance the selectivity and storage capacity of neutral and radioactive forms of hydrogen. Palladium can either impregnate the MOF pore from solution with a Pd precursor, followed by hydrogen reduction (reference: M. Sabo, A. Henschel, H. Frode, E. Klemm, S. Kaskel, J. Mater. Chem. 2007, 17, 3827-3832); another method to impregnate the MOF pore is by vapor deposition, again followed by hydrogen reduction (reference: S. Hermes, M. K. Schroter, R. Schmid, L. Khodeir, M. Muhler, A. Tissler, R. W. Fischer, R. A. Fischer, Angew. Chem. 2005, 117, 6394-6397; Angew. Chem. Int. Ed. 2005, 44, 6237-6241.) Yet another method of impregnating the MOF pore is via solution route, by an autoredox reaction between the organic ligand incorporated in the MOF solid and palladium(II) ions (reference: Cheon, Y. E.; Suh, M. P. Angew. Chem. Int. Ed. 2009, 48, 2899-2903). The MOF is modified post capture by the application of pressure. The pressure is in an amount sufficient to mechanically modify the MOF to trap the waste gas inside the MOF phase for longer periods of time and to higher temperatures than in the original unmodified MOF. In an embodiment, the amount of pressure is sufficient to create pressure-induced amorphization of the MOF, which alters the pore structure and sorption characteristics of the MOF without collapsing the MOF. In an embodiment, the amount of pressure to amorphized ZIF-8 may be up to 9 tons per square inch (18,000 psi). Pressure may be applied to the MOF by a variety of applications, such as, but not limited to pressing, grinding, ball milling, sonication, and hot isostatic pressing (HIPping) below the volitization temperature of the waste. In an embodiment, pressure may be applied by HIPping, at a temperature below the volitization temperature of the waste, of ZIF-8 between pressures of between about 7,350 psi and 18,000 psi. In another embodiment, pressure may be applied by HIPping between pressures of between about 10,000 psi and 18,000 psi. In another embodiment, pressure may be applied by HIPping at a pressure of about 15,000 psi. The modified MOF containing the stored waste may then be further processed by encapsulating in glass, stored in containers, or forming a core/shell waste form in which formed pellets (the core) are stacked in a low temperature sintered Bi—Si Glass tube (shell) and sealed by a sintered cap of the same composition as the shell, for example by the method disclosed in U.S. Pat. No. 8,262,950, which is herein incorporated by reference in its entirety. Detailed analysis of the local structure of I2 guests, using a PDF method which can probe both crystalline and amorphous materials, indicates that the MOF crystallographic cage and gas trapping ability remains unaffected by the framework amorphization. In contrast to chemical post synthetic modification of MOFs, mechanical modification through pressure can be applied at an intermediate point in a sorption or sequestration process, rather than simply being used to generate the initial sorbent and sorption characteristics. Consequently, when employed after guest-sorption, mechanical modification alters desorption properties to physically trap guests. This approach decouples guest sorption process from the guest trapping process; the process changes the bulk porosity of the MOF, and the gas sorption pathway into MOF. This is unique and distinct from instances in which adsorption of a molecule into a pore causes the MOF pore to either constrict or expand (sometimes named “guest-triggered switching of pores”). Whatever affects the adsorption of a molecule into the MOF pore on the resultant shape of the pore, the overall resulting MOF-guest molecule can be treated with this amorphization process to form an interim waste form. ZIF-8, Zn(2-methylimidazole)2, is part of a broad family of MOFs with expanded zeolite topologies—zeolitic imidazolate frameworks (ZIFs)—where the Zn-imidazolate-Zn link replicates the characteristic T-O-T angle of zeolites. The sodalite-type topology of the cubic ZIF-8 framework defines 12.0 Å diameter pores connected via 3.5 Å diameter apertures (6-rings). The unmatched I2 sorption capacity, retention, and selectivity of the ZIF-8 pore network benefits from the close correspondence of these 6-ring apertures to the I2 molecular dimensions. Activated (which means desolvated) ZIF-8 (approx. 1.5 g, Sigma-Aldrich) was loaded with iodine at approx. 75° C. to different final concentrations (20, 40, 60, 80, 100, and 120 wt % I2). A portion of each sample was annealed at 125° C. for 6 h to remove I2 from the external surface. While the as-loaded samples were brown, with the color intensity increasing for higher loadings, all samples faded to a pale brown or tan shade upon annealing, independent of I2 loading. This suggests that the color of the bulk sample is correlated to external surface-sorbed I2. Variable pressure X-ray diffraction measurements, for a 40 wt % I2 sample, indicate that the I2-containing sample can be amorphized at the same pressure as the vacant framework (approx. 0.34 GPa). Bulk powders of I2-loaded ZIF-8 (as-loaded and annealed) were amorphized within a pellet press (9 ton, 10-mm-diameter die, approx. 1.2 GPa average pressure), and redispersed as powders for subsequent analysis. The structure of the crystalline and amorphized materials were compared using PDF analysis of high energy X-ray scattering data collected at beamline 11-ID-B at the Advanced Photon Source at Argonne National Laboratory. The PDF provides local structure information, independent of crystallinity, as a weighted histogram of all atom-atom distances within a material. Not only can the PDF provide insight into the structural features that are retained in the amorphous ZIF-8 framework, but using a differential approach, and subtracting the contribution associated with the framework, the local I-I and I-framework interactions in the crystalline and amorphized materials can be directly compared. The well-defined long-range correlations, evident in PDFs for the crystalline materials, are eliminated for the amorphized ZIF-8 systems (FIG. 1). However, the shorter range features, including those up to 6 Å, which correspond to the Zn-imidazolate-Zn links, are entirely preserved in the amorphous materials. The combined retention of guests, porosity, and the Zn . . . Zn connectivity in the pressure-amorphized materials suggests that the sodalite topology of ZIF-8 is preserved, despite the local structural changes that destroy the long-range order, that is, the crystallinity. These structural changes are likely to involve symmetry—reducing distortions of the 6-ring apertures, eliminating the well-defined features in the PDF beyond approx. 6 Å and impeding diffusion of guest molecules through the framework. This displacive amorphization contrasts with the reconstructive transition to a dense amorphous phase induced thermally in ZIFs containing unsubstituted imidazole. While the long-range framework order is eliminated upon amorphization, the short-range I-I and I-framework interactions remain unchanged. Indeed, a larger change in local structure is associated with the annealing and surface-desorption compared to the amorphization itself (FIG. 1). Specifically, the nearest neighbor I-I peak shifts from 2.8 to 2.6 Å while simultaneously narrowing, indicating less disorder (dynamic or static). This is accompanied by an increase in the relative intensity of the second and third peaks at 3.85 and 4.3-4.4 Å, associated with intermolecular interactions within pores. These changes may reflect a refinement of the I2 arrangement within the pores upon annealing. The retention of I2 is enhanced in the amorphized relative to the crystalline ZIF-8, as evidenced from thermogravimetric analysis (TGA). The mass losses upon heating (10° C./min, N2 flow, see FIG. 2) were shifted to higher temperatures for the amorphized materials, by up to 150° C. These gains were most pronounced for the intermediate I2 loadings. At the highest loadings, there appeared to be some destabilization of the framework, with a greater overall mass loss for the I2-containing framework than for the vacant ZIF-8. This is consistent with the reduced crystallinity observed at high loadings. For the as-loaded samples, the TGA showed mass loss associated with surface desorption starting at 100-120° C., with a further loss at 170-240° C. associated with release of I2 from within the pores. For the surface-desorbed samples, the mass loss occurred at 170-240° C., starting at lower temperatures for higher loadings. The improvement in I2 retention was quantified by comparing the I2 mass loss from ZIF-8 samples (annealed 80 wt %) at constant temperature in the TGA apparatus (200° C., N2 flow). The I2 loss is retarded by a factor of approximately I2 in the amorphized ZIF-8 (see FIG. 2). The enhanced guest retention of the amorphized material is associated with a kinetic trapping mechanism, rather than a change in binding energy, with identical host-guest interactions in the crystalline and amorphous materials. Leach testing, which evaluates the long-term durability of a final waste form, by heating in deionized water for 7 days at 90° C., showed minor differences in I2 retention upon amorphization. As such, the amorphized pellets are most suitable for applications as an interim waste-form. These can be incorporated into core-shell structured final waste form that combines excellent long-term stability with potentially unmatched I2 densities. The ZIF-8 framework can be amorphized at the same mild pressure and temperature conditions as empty ZIF-8. This is particularly striking considering the extremely high I2 loading (up to approx. 6 I2 molecules per sodalite cage). This is generally consistent with the relative independence of certain MOFs' compressibility to different guest loadings. The structural changes to the framework associated with this amorphization improve the I2 retention upon heating, by up to 150° C., retaining the local structure of the captive I2 despite the changes to the framework crystallinity. An efficient radioactive waste capture process, in reprocessing nuclear fuel or cleanup following inadvertent environmental release, is perhaps one of the highest impact (in terms of cost savings per unit quantity) potential sorption applications for MOFs. In this field, there are overwhelming economic drivers that dictate that contamination, and accordingly, processing steps and materials, must be minimized, with the same storage requirements applied to any component used as part of the waste capture as for the waste itself. The amorphization of I2-loaded ZIF-8 provides for secure interim storage before incorporation into a long-term waste form, ensuring non-contamination of the environment. The unmatched I2 uptake capacity for ZIF-8, relative to existing zeolite getters, and the possibility to minimize subsequent release through amorphization, makes this a promising I2 capture method. Furthermore, the invariance of the host-guest interactions to pressure treatment, answers an important question in materials for radioactive waste storage. Here, an optional technology to densify radioactive waste to a monolithic form that is suitable for transport is hot isostatic pressing (HIPping). The modified MOFs of the present invention provide a route to an interim waste form based on amorphized ZIF-8, but more generally, it provides insight into the behavior of other I2 containing MOFs under HIPping conditions, showing that the pore structure can retain iodine under pressure. More generally, in contrast to chemical post synthetic modification of MOF structure and sorption-desorption behaviors which must be applied before guest-loading, this mechanical modification through pressure can be applied at any point in a sorption-sequestration process. Consequently, pressure-induced structural changes can be used as a macro-scale handle with which to control the nanoscale sorption properties. Specifically, they can be used for increasing hysteresis in the sorption-desorption kinetics, at will, to kinetically trap I2. In another embodiment, modified MOFs may be used for the controlled release of agrochemicals (insecticides, herbicides, and fungicides). This could reduce the harmful effects on the environment and have a targeted release on as needed basis. A similar concept is also valid for the targeted release of various drugs. In an example, 129I in the form of I2 gas, which may be from spent nuclear fuel reprocessing or from nuclear reactor accidents, which is of particular concern due to its very long half-life, its potential mobility in the environment and its deleterious effect on human health, may be treated to remove and store 129I. 129I is separated from spent fuel during fuel reprocessing as 129I2 vapor. The gas containing 129I2 vapor is passed through a bed of ZIF-8, such that the ZIF-8 selectively captures the 129I. the ZIF-8 is then modified by pressure to form a stable storage media. In another example, radiological Cs+ ion in aqueous solutions, for example in seawater, from nuclear reactor accidents or from spent nuclear fuel reprocessing is captured by a charged framework MOF material, and the MOF is modified by pressure treatment. In another example, radiological Uranyl (e.g., UO22+, (UO2)CO3(OH)3−, UO2(CO3)22−, UO2(CO3)34−) ion in aqueous solutions, for example seawater, is captured by a charged framework (2D) MOF material, and the MOF is modified by pressure treatment. The invention being thus described, it will be obvious that the same may be varied in many ways. Such variations are not to be regarded as a departure from the spirit and scope of the invention, and all such modifications as would be obvious to one skilled in the art are intended to be included within the scope of the appended claims. It is intended that the scope of the invention be defined by the claims appended hereto. The entire disclosures of all references, applications, patents and publications cited above are hereby incorporated by reference. In addition, many modifications may be made to adapt a particular situation or material to the teachings of the disclosure without departing from the essential scope thereof. Therefore, it is intended that the disclosure not be limited to the particular embodiment disclosed as the best mode contemplated for carrying out this disclosure, but that the disclosure will include all embodiments falling within the scope of the appended claims.
	summary
	summary
	abstract	A system and method of determining shape and position corrections of a beam such as a particle or other beam used in a system such as a particle beam lithography. The method of providing corrected deflector voltages may include determining a voltage step value by subtracting a previous deflector voltage value with a current deflector voltage value; determining a plurality of correction values using the voltage step value and an exposure time for the current deflector voltage value; selecting a current voltage correction value from the plurality of correction values using the current deflector voltage value; and calculating a corrected deflector voltage value by adding the current voltage correction value to the current deflector voltage value.
	abstract	The present invention is based on the property that the electric and magnetic fields are independent of each other and normal to each other and the property that the deflection of a charged particle beam by the electromagnetic field follows the rule of linear combination. The present invention employs a system that creates a region in which there exist both electromagnetic field and controls the deflection of a charged particle beam in each of the electric and magnetic fields.
	abstract	We disclose a precision positioner based on an inertial actuator, an optical instrument for accurate positional readout and control, and an electrostatically clamped assembly for holding any instrument or device. All aspects of the present invention present a significant improvement over the prior art: a positioner is robust and compact; an optical instrument for positional control is a profoundly simple and compact module; a clamping assembly is self-aligning and suitable for robotic hot-swapping of objects being positioned.
044977689	summary	BACKGROUND OF THE INVENTION The present invention relates generally to the quantitative evaluation of total fissile and total fertile nuclide content in samples, and more particularly to simultaneous photon and neutron interrogation of a sample coupled with the measurement of resulting prompt and delayed neutron emission as the basis for analysis of the totality of the fissile material present in the form of .sup.233 U, .sup.235 U or .sup.239 Pu, and the totality of fertile material present in the form of .sup.232 Th and .sup.238 U in the sample under investigation. Current U.S. Department of Energy guidelines for the management of transuranic waste have created a need for instrumentation to monitor such wastes at the 10 nCi/g level of fissile nuclides in the presence of fertile material. Solely passive systems which rely on the detection of gamma rays or neutrons from the decay or spontaneous fission of transuranic wastes are generally not suitable because of (1) the attenuation of high density matrix material with the attendant loss of sensitivity, and (2) interference from gamma and alpha emitting contaminants in the matrices. Active interrogation methods, wherein the waste is probed by an externally generated neutron or photon pulse, which cause fission in many of the nuclides present, or a combination of active and passive methods, overcome many of the problems associated with the passive methods alone. Two currently used active methods are: (1) photofission where a high energy photon beam induces fission in the waste sample and fission and delayed neutrons thereby produced are detected, and (2) thermal neutron fission where fast neutrons from a pulse source, after moderation, induce fissions in those nuclides present in the sample which are fissile. Active assay systems based on these methods have been used but neither approach itself is entirely adequate. Photofission offers good sensitivity for a large number of transuranic waste samples, but because of similarity of photofission cross-sections, identification of specific nuclides or classes of nuclides is difficult. For example, the important fissile and fertile groups cannot readily be distinguished. Thermal neutrons, on the other hand, offer very high sensitivity for fissile elements but essentially none for fertile elements. A combination of the two methods would be most desirable but technical complexities such as the need for two pulsed sources, together with longer assay times, have made such an analytical system impracticable. A combination of neutron and photon interrogation offers several distinct advantages over either applied alone, including a direct and unequivocable separation of fissile and fertile nuclides within the sample under investigation. The method and apparatus of the instant invention demonstrates that dual interrogation can be achieved using an electron linear accelerator (LINAC) as a pulsed source for both photons and neutrons. Moreover, both interrogations are initiated during each pulse of the LINAC, and the resulting prompt and delayed neutrons can be monitored with the same detection system. It is known in the art that high energy gamma radiation is produced as a result of bremsstrahlung in a heavy-metal target placed in the path of a high energy electron beam. The production of neutrons through the use of electron beams is also known, and occurs when the high energy gamma photons subsequently pass through additional layers of a target causing neutrons to be emitted in (.gamma.,n) processes. "Efficient Neutron Production Using Low-Energy Electron Beams," by C. D. Bowman, Nucl. Sci. Eng., 75, 12 (1980). However, the combination of neutrons and gamma radiation produced from a single source and used for analysis of transuranic waste samples for total fissile and total fertile nuclides present has not been reported. Most of the photons will pass into the volume of the waste sample where some will cause photofission. Prompt neutrons emitted from the photofission of either fissile or fertile nuclides will not be distinguishable from photoneutrons that are formed in the materials of the chamber containing the sample under investigation and/or the matrix materials which contain these fissile and fertile nuclides. However, delayed neutrons from photofission will be emitted on a continual basis during the whole period between LINAC pulses. Photoneutrons and prompt photofission neutrons will thermalize in a few tens of microseconds and will persist as thermals for hundreds of microseconds, during which time they will generate thermal neutron fissions among the fissile transuranic nuclides that may be present. Therefore, fission neutrons from thermal fission are separated in time from the photoneutrons, and can serve, along with delayed neutrons, as a quantitative signature. Essentially then, after an initial burst of photoneutrons and neutrons from photofission, the bulk of the fast prompt neutrons derive from thermal fission of fissile materials. Subsequent to the emission of these prompt neutrons are the emission of delayed neutrons which derive from both photofission of fertile and fissile material as well as delayed neutrons from thermal neutron fission of fissile nuclides. If the latter contribution to the total delayed neutron flux is made small, the delayed neutron emission is representative of the photofission events only. To achieve this result, some iteration of the interrogation neutron flux may be necessary when analyzing samples of completely unknown fissile content. This flux can be varied by choosing different target materials for the (.gamma.,n) source. Thus, in the instant invention, the events detected following a single LINAC pulse are separable into neutron fission (prompt fission neutrons) and photofission (delayed neutrons) events. This data can then be analyzed to yield the individual quantities of fertile and fissile isotopes present. SUMMARY OF THE INVENTION An object of the apparatus and method of the instant invention is to quantitatively evaluate total fissile and total fertile nuclide content in samples. Another object of our invention is to screen transuranic wastes for storage arrangements at the 10 nCi/g level. Additional objects, advantages and novel features of the invention will be set forth in part in the description which follows, and in part will become apparent to those skilled in the art upon examination of the following or may be learned by practice of the invention. The objects and advantages of the invention may be realized and attained by means of the instrumentalities and combinations particularly pointed out in the appended claims. To achieve the foregoing and other objects in accordance with the purpose of the present invention, as embodied and broadly described herein, the method of this invention may comprise generating repetitively pulsed gamma radiation and neutrons, using these radiations to interrogate the sample of interest, and detecting the emitted prompts and delayed fast neutron flux. High energy repetitively pulsed gamma radiation is generated and directed onto a partially transparent target from which photoneutrons are generated while allowing a substantial portion of the gamma radiation to pass through. The transmitted gamma radiation and generated photoneutrons are then allowed to simultaneously impinge on a chamber which surrounds and contains the sample under investigation. The photoneutrons are then thermalized by collisions within the walls of the chamber resulting in an enhanced probability that any fissile isotopes present within the sample will undergo a fission reaction. The high energy gamma radiation causes photofission in both the fertile and fissile materials, while the thermalized neutrons produce fission in only the fissile material present. From the actual fission processes, and from the fragments produced therefrom, both prompt and delayed fast neutrons are emitted. These emitted neutrons, along with the thermalized photoneutron flux, are measured in between the gamma radiation pulses. A plurality of such measurements are accumulated until a statistically significant signal is obtained, and the results, normalized by the accumulated thermalized neutron flux and gamma radiation photon flux, are related to the total fissile and the total fertile nuclide concentration contained in the sample to be quantitatively assayed, using known gamma and neutron cross-sections for these nuclides. Preferably, the pulsed gamma radiation has a photon energy in excess of about 10 MeV in order to produce significant numbers of photoneutrons. Preferably, also, the photoneutrons are thermalized within about 0.5 ms after the termination of a particular gamma radiation pulse. Since the fissile nuclides undergo both photofission and thermal neutron fission, the delayed neutron emission is composed of contributions from both the fertile nuclides and the fissile nuclides. Therefore, the gamma radiation generated neutron flux is adjusted such that the emitted delayed neutron flux is comprised principally of neutrons from only the photofission reactions, while allowing sufficient prompt fast neutron emission from the thermal neutron fission process in the fissile nuclides to enable statistically significant neutron measurements to be obtained in a practical accumulation time period. This is possible because of the high sensitivity of the method for prompt fast neutron emission, derived from the temporal and energy selectivity of the .sup.3 He-proportional counting system utilized. Preferably, prompt fast neutron measurements are made between about 0.5 and and 2.5 ms after the termination of a particular gamma radiation pulse, and this measurement determines the number of thermal-neutron-induced fissions. The delayed fast neutron measurements are performed between about 5.5 ms after the termination of a particular gamma radiation pulse and the commencement of the following gamma radiation pulse. Adjustment of the thermal-neutron interrogating flux, if necessary, is performed by allowing larger or smaller amounts of photoneutron emitters to be placed in the gamma radiation beam pathway. It is also preferred that the gamma radiation is derived from an electron accelerator, the high energy electrons produced therefrom, being caused to impinge upon a metal target thereby producing the desired radiation by a bremsstrahlung process. It is finally preferred that the photoneutron emitting target include beryllium. In a further aspect of the present invention, in accordance with its objects and purposes, the apparatus hereof may also comprise means for simultaneously generating repetitively pulsed gamma radiation and photoneutrons produced therefrom, a chamber which surrounds and contains the sample under investigation and into which the neutrons and gamma radiation are directed, means for thermalizing these photoneutrons to increase their fission reaction effectiveness for fissile nuclides present in the sample, means for measuring the thermalized photoneutron flux in the vicinity of the sample and for measuring prompt and delayed fast neutron emission from the sample as a result of photofission of fertile and fissile nuclides interacting with the gamma radiation, and from thermal neutron fission of fissile nuclides capturing thermalized photoneutrons during a time interval after the termination of the gamma radiation pulse, and the means for accumulating a plurality of such neutron measurements until a statistically significant signal is obtained and means for recording these accumulated measurements. Preferably, the thermalization of the photoneutrons occurs within the walls of the chamber. It is also preferred that the thermalized photoneutron flux is measured using at least one bare low partial pressure .sup.3 He-proportional counter located inside the chamber in the vicinity of the sample, and the prompt and delayed fast neutron emission measurements are made using at least one high pressure .sup.3 He-proportional counter surrounded by polyethylene which is further surrounded by cadmium foil and located within the chamber in the vicinity of the sample. Such shielded detectors block the large thermalized photoneutron background while allowing very sensitive prompt fission neutron detection. Preferably also, the gamma radiation is produced by means of an electron accelerator capable of providing greater than about 1 ma of electron beam current in short duration pulses at a repetition rate between 1 and 60 Hz and at energies in excess of about 10 MeV, said electrons impinging upon a heavy metal target producing high-energy gamma radiation by means of a bremsstrahlung process. Preferably, the pulsed gamma radiation is filtered to remove photons of low energy to reduce the gamma radiation background, and to remove background neutrons from the bremsstrahlung target using a polyethylene slab placed in between the bremsstrahlung target and the chamber. The polyethylene slab often provides sufficient photoneutrons for thermal fission interrogation without the need for the presence of beryllium which is sometimes inserted in the polyethylene slab in small quantities in the path of the gamma radiation. It is also preferred that the chamber wall be constructed of a thick polyethylene and graphite inner wall surrounded by a layer of cadmium foil which is further surrounded by a thick outer wall of borated polyethylene in order to reduce the effect of stray neutrons generated by the electron accelerator away from the direction of the electron beam, since it is desired that the interrogation neutrons originate in the chamber, and to thermalize the photoneutrons incident on this chamber. Finally, it is preferred that the number of neutrons generated from the pulsed gamma radiation is controlled by selection of a suitable target material which for low neutron flux requirements is simply the chamber wall, and for high neutron flux requirements may include a beryllium sheet placed in between the gamma radiation filter and the chamber. In summary, the apparatus and method of the instant invention allows the distinguishing of fissile and fertile nuclides in a sample under investigation, which is not available from separate interrogation by either neutrons or gamma radiation alone. For example, thermal neutron interrogation offers very high sensitivity for fissile elements but essentially none for fertile nuclides. It is important that the fertile elements be measured since a particular assay for fissile nuclides made alone cannot identify the material as being of low enrichment (small fissile-to-fertile ratio) or high enrichment (large fissile-to-fertile ratio). For proper disposal and handling a fissile-to-fertile ratio is required. The simultaneous production of pulsed gamma radiation and thermal neutrons from a single source provides great simplification for the experimental procedures necessary to do the assay of samples, and also greatly reduces the time involved in performing such assays.
	summary
	description	The present disclosure relates to a radiation attenuation system. More particularly, the present disclosure relates to a radiation attention system adapted for use with Computed Tomography procedures such as Computed Tomography scanning procedures and Computed Tomography fluoroscopy procedures. The present disclosure further relates to radiation attenuation system that is intended to reduce radiation exposure to at least one of a patient and medical personnel during Computed Tomography procedures. Computed Tomography (CT) procedures are commonly used to obtain cross-sectional images of the patient's body, including images of a patient's brain, lungs, heart, liver, bones, blood vessels, etc. CT procedures are often used to diagnose different kinds of diseases such as cancer, to plan radiation treatments and surgeries, and to guide physicians during biopsies and other invasive procedures. CT procedures involve the use of CT machines that use x-ray radiation to obtain the cross-sectional images. In conducting a CT procedure, a patient is placed in the CT machine between an x-ray generating source and an x-ray detecting sensor. The CT machine delivers controlled amounts of x-ray radiation from the x-ray generating source to the portion of the patient's body being examined. The x-ray detecting sensor is positioned on the other side of the patient and captures the x-ray radiation passing through the body of the patient. The x-ray detecting sensor sends an output signal to a processor representative of the amount of x-ray radiation absorbed by the patient. The processor receives the output signal from the x-ray detecting sensor and processes the signal to create the cross-sectional images of the patient on a display. As presently configured, areas in which CT procedures are conducted (i.e. CT areas) expose not only the patient to radiation, but also the physicians and other medical personnel that may be present during the procedure. In CT procedures, significant amounts of radiation may be scattered to the patient and to the physician, or other medical personnel in the CT area (i.e. scatter radiation). The likelihood of having radiation scattered to the physician or other medical personnel is increased for CT fluoroscopy guided interventional procedures during which the medical personnel is in the CT area during the scan. Exposure to radiation may create potential health concerns. Radiation specialists and government agencies recognize the potential health risks caused by ionizing radiation and have developed the principle of ALARA (As Low As Reasonably Achievable). The principle of ALARA requires that radiation levels be reduced to the greatest degree possible taking into account a reasonable cost and physical application. Accordingly, it would be advantageous to provide a radiation attenuation system that may be used during CT procedures to minimize a patient's exposure to radiation. It would further be advantageous to provide a radiation attenuation system that reduces the amount of radiation exposure for medical personnel working in a CT area. It would also be advantageous to provide a radiation attenuation system that is relatively flexible and compliant, and adaptable for use with a variety of CT machines and CT procedures. It would also be advantageous to provide a radiation attenuation system that is disposable. It would also be advantageous to provide a radiation attenuation system that is sterilizible before use. It would also be advantageous to provide a radiation attenuation system that may be coupled to CT devices having different configurations. It would further be advantageous to provide a radiation attenuation system for protecting medical personnel that is suitable for use with CT fluoroscopy procedures where medical personnel may need to insert biopsy needles or other instrumentation without hindrance. It would also be advantageous to provide a radiation attenuation system which provides a relatively high degree of comfort to the user. It would be desirable to provide for a radiation attenuation system having one or more of these or other advantageous features. An exemplary embodiment relates to a system for the attenuation of radiation during a Computed Tomography procedure. Computed Tomography procedures are conducted using a Computed Tomography machine having a gantry defining an opening. The system includes a shield made of a radiation attenuation material. The shield is configured to be disposed at least partially in front of the opening defined by the gantry of the Computed Tomography machine to reduce radiation exposure during the Computed Tomography procedure. Another exemplary embodiment relates to a system for the attenuation of radiation during a Computed Tomography procedure. Computed Tomography procedures are conducted using a Computed Tomography machine having a gantry defining an opening. The system includes a shield made of a radiation attenuation material. The shield is configured to be positioned between a medical personnel and the Computed Tomography machine to protect the medical personnel from radiation exposure during the Computed Tomography procedure. Another exemplary embodiment relates a system for the attenuation of radiation during a procedure that emits ionizing radiation. The system includes a shield made of a radiation attenuation material. The shield is configured to be draped over and around substantially all secondary areas of the patient to protect the secondary areas of the patient from radiation exposure. Another exemplary embodiment relates to a method of attenuating radiation exposure to a medical personnel during a Computed Tomography procedure preformed by a Computed Tomography machine. The method includes the steps of disposing a radiation attenuation material on the Computed Tomography machine between the medical personnel and the Computed Tomography machine. Another exemplary embodiment relates to a system for the attenuation of radiation during a Computed Tomography procedure conducted using a Computed Tomography machine. The system includes a means for reducing radiation exposure to a medical personnel during the Computed Tomography procedure. The means is coupled to the Computed Tomography machine and positioned between the Computed Tomography machine and the medical personnel. FIG. 1 shows a Computed Tomography (CT) machine 20 of the type commonly used to create cross-sectional images the body of a patient 10. CT machines are well known and widely used in the medical field. Accordingly, CT machine 20, as illustrated, is intended to be representative of all conventionally known CT machines and is not intended to be limited to the exact configuration shown. CT machine 20 may be used for both CT scanning procedures and CT fluoroscopy procedures. As used herein, the use of the term “CT scanning procedures” is intended to mean CT procedures conducted as part of a noninvasive examination during which a medical personnel 12 (e.g. a physician, nurse, technician, and the like) administering or otherwise involved with a CT procedure is likely to be outside of the area in which the CT procedure is conducted (i.e., the CT area). The term “CT fluoroscopy procedure”, as used herein, is intended to mean CT procedures conducted as part of an invasive surgical procedure during which the medical personnel 12 is likely to remain in the CT area and substantially near the CT machine during the CT procedure. CT machine 20 includes a housing 22 having a front side 23 and a back side 25. Housing 22 encloses a support structure, commonly referred to as gantry 24, that is configured to support at least one x-ray emitter 26 and at least one x-ray detector 28. Gantry 24 may support the x-ray emitter 26 and the x-ray detector 28 in a manner sufficient to allow for the orbital rotation of x-ray emitter 26 and x-ray detector 28 around patient 10. The gantry 24 defines an opening 30 in which at least a portion of patient 10 is inserted during the CT procedure. In its most common form, opening 30 is a generally circular opening. The illustration of opening 30 as a circular opening is not intended to limit the applicability of the present invention to CT machines having circular openings. As can be appreciated, the present invention is equally applicable with alternative CT machines having openings configured in any of a variety of shapes. CT machine 20 further includes a patient table 38 configured to support the body of patient 10. Patient table 38 is generally positioned perpendicular to the front side 23 of housing 22 and may be movable in the vertical and horizontal directions relative to opening 30 as well as transversely. As can be appreciated, for alternative CT machines, patient table 38 may remain stationary and housing 22 may move relative to patient table 38. To obtain an image, patient 10 is placed on patient table 38 and moved into opening 30 wherein patient 10 is positioned between x-ray emitter 26 and x-ray detector 28. A primary beam of x-ray radiation emanating from x-ray emitter 26 passes through patient 10 before being captured by x-ray detector 28. The x-ray radiation beam emanating from x-ray emitter 26 and passing through patient 10 is referred to herein as entrance radiation. During CT fluoroscopy procedures, wherein medical personnel 12 is standing near patient 10 and CT machine 20, medical personnel 12 may be inadvertently exposed to entrance radiation and radiation leakage from CT machine 20. In addition to entrance radiation and radiation leakage, CT procedures are likely to generate scatter radiation. Scatter radiation refers to radiation emanating from x-ray emitter 26 that reflects off of an object such as patient 10, CT machine 20, the floor in CT area, etc. and scatters throughout the CT area. During a typical CT scanning procedure, the only person likely to be exposed to scatter radiation is patient 10. However, during CT fluoroscopy procedures, or any other CT scanning procedure in which medical personnel 12 remain in the CT area, medical personnel 12 may also be exposed to scatter radiation. As explained above, exposure to radiation may create a health risk and should be reduced whenever practicably possible. Referring to FIG. 2, a radiation attenuation system 100 configured to minimize radiation exposure during a CT procedure is shown. Radiation attenuation system 100 includes a first radiation attenuation system 200 that is intended to assist in the protection of patient 10 from unnecessary exposure to radiation during a CT procedure and a second radiation attenuation system 300 that is intended to assist primarily in the protection of medical personnel 12 from exposure to radiation during a CT procedure. Radiation attenuation system 200 and radiation attenuation system 300 include at least one radiation barrier article for reducing radiation exposure. Depending on the CT procedure being performed, first radiation attenuation system 200 and second radiation attenuation system 300 may be used in combination, or alternatively may be used separately as individual radiation attenuation systems. During a CT procedure, patient 10 must be exposed to x-ray radiation (i.e. entrance radiation) in order for cross-sectional images of the patient's body to be obtained. CT procedures are often focused on a specific portion of the patient's body (i.e. the target area). While the target area must be exposed to entrance radiation, the surrounding portions of the patient's body (i.e. secondary areas) do not have to be exposed. Radiation attenuation system 200 is intended to minimize a patient's exposure to entrance radiation, radiation leakage and scatter radiation present during a CT procedure by shielding the secondary areas. Referring to FIG. 3, radiation attenuation system 200 includes a radiation attenuation wrap, shield, cloth, or garment 210. Garment 210 may be useful in blocking or attenuating radiation, and assisting in the protection of patient 10. Garment 210 may be made of any radiation attenuation material and preferably is made of a light-weight and flexible radiation attenuation material. Preferably garment 210 is made of a radiation attenuation material that provides a relatively high degree of comfort to the patient. Garment 210 may used to cover the portions of patient 10 during a CT procedure that are not going to be examined. Garment 210 preferably includes a body cover portion 212 and a head cover (e.g. hood, hat, helmet, etc.) portion 214. Body cover portion 212 is not limited covering a patient's torso and may be configured to include leg cover portions, foot cover portions, arm cover portions, and hand cover portions. Preferably, garment 210 wraps around (e.g. underneath) patient 10 and does not simply drape over the top of patient 10. Head cover portion 214 is intended to protect a patient's head from radiation exposure, and may include portions covering a patient's face, forehead and neck. As can be appreciated, the configuration of garment 210 may vary depending on the application and portion of the patient's body that is to be scanned. For example, it would be anticipated that garment 210 would be configured differently for scanning of the chest as compared to the abdomen or an extremity. Garment 210 may be made in range of sizes to fit adult or adolescent patients as well as infants. Garment 210 may include a fenestration area 216 for providing access to the target area (i.e. the portion of the patient's body to be scanned) through an aperture (shown as an rectangular strip 218). Fenestration area 216 further provides an opening for allowing medical personnel 12 to access patient 10 for conducting various invasive procedures, such as the fluoroscopic guidance and/or manipulation of instruments during surgical procedures. According to a preferred embodiment, fenestration area 216 may be selectively sealed or opened by coupling a fastener 220 to garment 210 near fenestration area 216. According to a particularly preferred embodiment, a hook and loop fastener is coupled to garment 210 and allows fenestration area 216 to be selectively sealed or opened depending on the CT procedure being conducted. According to a particularly preferred embodiment, garment 210 is configured as a combination of a skirt, a vest, and a helmet. Such a configuration may be particularly suitable for procedures wherein the target area is the patient's abdomen or chest area. During a procedure of a patient's abdomen or chest, medical personnel can access the target area by moving a portion of the vest upwards to expose the desired area. However, the garment 210 is not limited to such a configuration, and such a garment could be used for procedures wherein the target area is not the patient's abdomen or chest. While garment 210 is shown as an attenuation system that may be useful during CT procedures to protect a patient from radiation exposure, garment 210 is equally applicable with any procedure that emits ionizing radiation such as, but not limited to, intraoperative use of radiation equipment and implanting radiation therapy devices into patients that emit radiation. As stated above, physicians, nurses, technicians, and other health care employees (collectively referred to as medical personnel) present during a CT procedure may be exposed radiation. Medical personnel present for numerous CT procedures may be exposed to significant cumulative radiation doses over time. Radiation attenuation system 300 is intended to reduce radiation exposure to medical personnel 12 present in the CT area during a CT procedure. Radiation attenuation system 300 may be particularly applicable with CT fluoroscopy procedures wherein medical personnel 12 is likely to be near the primary beam of x-ray radiation emanating from the CT machine or at least in an area susceptible to secondary scattered radiation or radiation leakage. Radiation attenuation system 300 includes at least one radiation barrier article coupled substantially near or to CT machine 20 configured to reduce radiation exposure to medical personnel 12. Referring to FIG. 4, radiation attenuation system 300 may include a first radiation panel, shield, or pad 320 that may be useful in blocking or attenuating radiation, and assisting in the protection of medical personnel 12. Pad 320 is made of a radiation attenuation material and is positioned between CT machine 20 and medical personnel 12. Pad 320 may be coupled near or to the CT machine 20. Preferably, pad 320 is substantially rectangular shape having an outer edge that includes a top edge 322 and an opposite bottom edge 324. Pad 320 may be coupled near or to CT machine in any position that may protect medical personnel 12 and/or patient 10 from unintentional radiation exposure. According to an exemplary embodiment, pad 320 is coupled to the side of patient table 38. Pad 320 is coupled near the top surface of patient table 38 and hangs, extends, or drapes over the side of patient table 38 so that bottom edge 324 is near the floor of the CT area. Depending on the size of pad 320 and patient table 38, multiple pads 320 may be coupled to patient table 38 in order to provide sufficient protection for medical personnel 12. According to a second exemplary embodiment, pad 320 is coupled to the front side 23 of CT machine 20. Pad 320 is coupled near gantry 24 substantially tangential to the bottom of opening 30 and extends downward so that bottom edge 324 is near the floor. Bottom edge 324 may be weighted in order to urge bottom edge 324 in the direction of the floor and help maintain pad 320 in a protective position. As can be appreciated, pad 320 is suitable for use anywhere in the CT area so long as pad 320 is between CT machine 20 and medical personnel 12. According to a preferred embodiment, shown in FIG. 5, pad 320 is a flexible member sized to span across both the areas covered by the first and second exemplary embodiments described above. For such an embodiment, pad 320 may be described as having two portions, a first panel 328 that is integrally formed with a second panel 330. First panel 328 is coupled to the front side 23 of CT machine 20 near gantry 24 and second panel 330 is coupled to a side portion of patient table 38 that is near opening 30. Pad 320 is positioned between CT machine 20 and medical personnel 12. First panel 328 and second panel 330 conform to the contour of CT machine 20 and are substantially perpendicular to each other. As previously stated, bottom edge 324 may be weighted. Such a configuration is intended to reduce the amount of radiation exposure experienced by medical personnel 12 while enabling medical personnel 12 to remain substantially close to opening 30 and patient 10. Referring to FIGS. 4 and 5, to couple pad 320 to CT machine 20, pad 320 may include a fastener 326. According to a preferred embodiment, fastener 26 allows for the detachable coupling of pad 320 to CT machine 20. According to a particularly preferred embodiment, pad 320 includes a hook and loop fastener coupled to the outer edge of pad 320 for allowing the detachable coupling of pad 320 to CT machine 20. As shown in FIGURES, fastener 26 is coupled to top edge 322. In alternative embodiments, fastener 26 may be coupled anywhere along the outer edge of pad 320, or anywhere else along pad 320. As can be appreciated, a number of suitable fasteners may provide the detachable coupling of pad 320 to CT machine 20 in addition to hook and loop fasteners such as, snaps, grommets, adhesives, zippers, etc. Preferably, pad 320 is coupled to CT machine 20 and patient table 38 on both sides of patient table 38 as shown in FIG. 2. If pad 320 includes a detachable fastener 26, a single pad 320 can be utilized by selectively positioning pad 320 along CT machine 20 and patient table 38 to protect medical personnel 12. As can be appreciated, pad 320 may be dimensioned and shaped in any of a variety of ways depending on the application. For example, pad 320 may be configured in any of a variety of shapes such as a pad having a curvilinear portion to more readily conform to a CT machine. Referring to FIG. 6, radiation attenuation system 300 may also include a second radiation barrier article, shown as radiation curtain, shield, or drape 340. Drape 340 may be useful in blocking or attenuating radiation, and assisting in the protection of medical personnel 12. Drape 340 is intended to be positioned between CT machine 20 and medical personnel 12. Drape 340 is coupled near gantry 24 of CT machine 20 and substantially covers opening 30. Drape 340 may be made of any attenuation material and is intended to reduce the amount of entrance radiation, radiation leakage and scatter radiation that medical personnel 12 or patient 10 may be exposed to during a CT procedure. In its most preferred form, drape 340 is a made of a flexible attenuation material having an outer edge that includes a bottom edge 344 that hangs downward from a top edge. Preferably, bottom edge 344 drapes around patient 12 and conform to the patient's body and patient table 38. Similar to pad 320, drape 340 may include a fastener, such as a hook and loop fastener, along the outer edge and may further be weighted along bottom edge 344 to maintain a desired position. According to an exemplary embodiment, drape 340 is a solid shield or member covering opening 30 (shown in FIG. 7a). Configuring drape 340 as a solid member may be particularly useful during CT scanning procedures during which medical personnel 12 do not need access to the portion of the patient's body being scanned. Drape 340 may include a viewing panel (shown as a window 346 in FIG. 7b) that is relatively clear or translucent for the viewing of patient 12 within CT machine 20. Window 346 may be of a variety of shapes and sizes, which may be dictated at least in part by the particular application. To accommodate CT procedures during which it would be desirable for medical personnel 12 to access the portion of the patient's body being scanned, drape 340 may include a fenestration area 342 for providing access to the portion of the patient that is within CT machine 20 during the CT procedure. Fenestration area 342 may be an aperture (shown as a rectangular opening in FIG. 7c) that allows medical personnel 12 to insert medical instrumentation when conducting various invasive procedures, such as the fluoroscopic guidance and/or surgical procedures. According to a preferred embodiment, shown in FIG. 7d, drape 340 may be configured as a plurality of flaps 348 which do not substantially restrict medical personnel 12 from accessing patient 10. According to an alternative exemplary embodiment, as shown in FIG. 7e, drape 340 is a solid member having a slit or cut extending from the bottom edge in a substantially vertical direction to define flaps 348 thereby providing access to patient 10. According to a particularly preferred embodiment, drape 340 is a solid barrier having a plurality of slits formed in a substantially vertical direction to define flaps 348 (shown in FIG. 7f). The use of flaps 348 in combination with drape 340 is intended to reduce the radiation exposure experienced by medical personnel 12 without substantially restricting access to patient 10. As shown in FIGS. 7a–7f, drape 340 is a generally rectangular shield that is disposed across opening 30. As can be appreciated, drape 340 may be dimensioned and shaped in any of a variety of ways depending on the CT machine and the application. For example, drape 340 may be configured in any of a variety of shapes such as a shield having a curvilinear portion to more readily conform to a CT machine (shown in FIG. 7g). Alternatively, drape 340 may be configured as having a circular shape (shown in FIG. 7h). According to a preferred embodiment, as shown in FIG. 2, radiation attenuation system 300 includes the use of both pad 320 and drape 340 to assist in the protection of patient 10 and medical personnel 12. The combination of pad 320 and drape 340 may increase the level of protection relative to the use of any one of the articles alone. The radiation barrier articles of radiation attenuation system 300 (i.e. pad 320 and drape 340) may be selectively positionable to allow medical personnel 12 to move an article out of the way if the article is not needed. Each of the barrier articles of radiation attenuation system 100 described above may be made of any radiation attenuation material including, but not limited to, bismuth, barium, lead, tungsten, antimony, copper tin, aluminum, iron, iodine, cadmium, mercury, silver, nickel, zinc, thallium, tantalum, tellurium, and uranium. Anyone of the aforementioned attenuation materials alone or in a combination of two or more of the attenuation materials may provide the desired attenuation. According to a particularly preferred embodiment, the articles of radiation attenuation system 100 are made of the attenuation material disclosed in U.S. Pat. No. 6,674,087, which is hereby incorporated by reference. However, the articles of radiation attenuation system 100 are not limited to such an embodiment. The degree of radiation transmission attenuation factor by the radiation attenuation material may be varied depending upon the specific application. According to an exemplary embodiment, the radiation attenuation material will have a radiation transmission attenuation factor of a percent (%) greater than about 50%, suitably greater than about 90%, suitably greater than about 95%. Preferably, the radiation attenuation material is generally light and flexible, to maximize workability for processing, bending, folding, rolling, shipping, etc. The material may be formable (e.g. deformable) or compliant, and relatively “stretchable” (e.g. elastic). According to alternative embodiments, the material used may be generally rigid and inflexible, and/or substantially weighted. According to a preferred embodiment, the articles of radiation attenuation system 100 are generally disposable in whole or in part, thereby minimizing ancillary sources of contamination that may arise from multiple uses. According to another suitable embodiment, the articles of radiation attenuation system 100 are generally non-toxic, recyclable, and/or biodegradable. According to an alternative embodiment, the articles of radiation attenuation system may be reusable (e.g. for attenuation of radiation from atomic/nuclear disaster, clean up, rescue operations, etc.). According to a preferred embodiment, the articles of radiation attenuation system may be sterilized between uses to minimize the likelihood of bacteriological or virus contamination. Sterilization may be performed in any convenient manner, including gas sterilization and irradiation sterilization. The construction and arrangement of the articles of the radiation attenuation system as shown in the preferred and other exemplary embodiments is illustrative only. Although only a few embodiments of the present inventions have been described in detail in this disclosure, those skilled in the art who review this disclosure will readily appreciate that many modifications are possible (e.g. variations in sizes, dimensions, structures, shapes and proportions of the various elements, values of parameters, mounting arrangements, use of materials, colors, orientations, etc.) without materially departing from the novel teachings and advantages of the subject matter recited in the claims. For example, garment 210 may be configured in a variety of ways (e.g. as a vest, gown, pants, skirt, shirt, etc.) depending on the application. Further, pad 320 and drape 340 may be configured as screens or curtains that are not coupled to CT machine 20 but are instead positioned near CT machine 20. Accordingly, all such modifications are intended to be included within the scope of the present invention as defined in the appended claims. The order or sequence of any process or method steps may be varied or re-sequenced according to alternative embodiments. In the claims, any means-plus-function clause is intended to cover the structures described herein as performing the recited function and not only structural equivalents but also equivalent structures. Other substitutions, modifications, changes and omissions may be made in the design, operating conditions and arrangement of the preferred and other exemplary embodiments without departing from the spirit of the present inventions as expressed in the appended claims.
	claims	1. A sealing mechanism for a nuclear reactor vessel cable penetration tube, the sealing mechanism comprising:a penetration tubeconfigured to penetrate a nuclear reactor vessel from an outside to advance to an inside thereof andcomprising an axially extending penetration hole for communication with the inside of the nuclear reactor vessel;a cable configured to be inserted to the inside of the nuclear reactor vessel through the penetration hole of the penetration tube; anda thimble placed between the cable and the penetration tube such that the thimble surrounds the cable within the penetration tube,the thimble comprising a dimple groove extending in a direction from an outer surface of the thimble toward the cable,wherein a diameter of the penetration hole at a first end portion of the penetration tube and a diameter of the penetration hole at a second end portion of the penetration tube are both smaller than a diameter of the penetration hole at a middle portion of the penetration tube,wherein the thimble is supported by the first end portion and the second end portion of the penetration tube,wherein the thimble is axially shorter than the penetration hole, wherein a thermal expansion space is defined between one or more of the thimble and the first end portion of the penetration tube or the thimble and the second end portion of the penetration tube, andwherein the penetration tube is provided in a lateral side of the nuclear reactor vessel. 2. The sealing mechanism of claim 1, wherein the dimple groove portion has a round groove shape. 3. The sealing mechanism of claim 1, further comprising a guide tube protruding from an end of the penetration tube, the guide tube comprising a guide hole communicating with the penetration hole to receive the cable. 4. The sealing mechanism of claim 1, further comprising a guide tube and a Swagelok fitting,wherein the guide tube protrudes from an end of the penetration tube and comprises a guide hole communicating with the penetration hole to receive the cable, andthe Swagelok fitting is installed on the end of the penetration tube or an end of the guide tube. 5. The sealing mechanism of claim 1,wherein the thimble comprises a thimble protrusionbetween the first end portion and the middle portion of the penetration tube orbetween the second end portion and the middle portion of the penetration tube.
040509869	abstract	A nuclear reactor, particularly a liquid-metal breeder reactor whose upper internals include provision for channeling the liquid metal flowing from the core-component assemblies to the outlet plenum in vertical paths in direction generally along the direction of the respective assemblies. The metal is channeled by chimneys, each secured to, and extending from, a grid through whose openings the metal emitted by a plurality of core-component assemblies encompassed by the grid flows. To reduce the stresses resulting from structural interaction, or the transmissive of thermal strains due to large temperature differences in the liquid metal emitted from neighboring core-component assemblies, throughout the chimneys and the other components of the upper internals, the grids and the chimneys are supported from the heat plate and the core barrel by support columns (double portal support) which are secured to the head plate at the top and to a member, which supports the grids and is keyed to the core barrel, at the bottom.. In addition to being restrained from lateral flow by the chimneys, the liquid metal is also restrained from flowing laterally by a peripheral seal around the top of the core. This seal limits the flow rate of liquid metal, which may be sharply cooled during a scram, to the outlet nozzles.. The chimneys and the grids are formed of a highly-refractory, high corrosion-resistant nickel-chromium-iron alloy which can withstand the stresses produced by temperature differences in the liquid metal. The chimneys are supported by pairs of plates, each pair held together by hollow stubs coaxial with, and encircling, the chimneys. The plates and stubs are a welded structure but, in the interest of economy, are composed of stainless steel which is not weld compatible with the refractory metal. The chimneys and stubs are secured together by shells of another nickel-chromium-iron alloy which is weld compatible with, and is welded to, the stubs and has about the same coefficient of expansion as the highly-refractory, high corrosion-resistant alloy.
	summary
	claims	1. A microscope examination method comprising:a step of introducing into a specimen a substance in which a structural change is caused by irradiation with light of a first wavelength;a step of specifying an optical stimulation site in the specimen by irradiating the specimen with visible light of a second wavelength that does not cause a structural change in the substance, while examining the specimen in which the substance is introduced using a microscope examination apparatus; anda step of irradiating the specified optical stimulus site with the light of the first wavelength. 2. The microscope examination method according to claim 1, wherein the microscope examination apparatus is formed of a fluorescence microscope that irradiates the specimen with excitation light of a third wavelength different from the first wavelength and the second wavelength to observe the fluorescence produced thereby. 3. A microscope examination method according to claim 1, wherein the microscope examination apparatus is a stereomicroscope. 4. An optical stimulation apparatus comprising:a first light source that emits light of a first wavelength that irradiates a substance introduced into a specimen to cause a structural change thereof;a second light source that emits visible light of a second wavelength, different from the first wavelength, the second wavelength not causing a structural change in the substance;an optical-axis combining unit that aligns the optical axis of the first light source and the optical axis of the second light source; anda focusing optical system that focuses the light from the first and second light sources, wherein the focusing optical system is removably attached to the optical-axis combining unit. 5. An optical stimulation apparatus according to claim 4, further comprising:optical fibers that connect the optical-axis combining unit with the first and second light sources, respectively; andfocus adjusting mechanisms that can adjust the tip positions of the optical fibers at the connection points between the optical-axis combining unit and the optical fibers. 6. An optical stimulation apparatus according to claim 4, further comprising:an optical fiber that guides the light beams whose optical axes are aligned by the optical-axis combining unit to the focusing optical system;a focus adjusting mechanism that can adjust the distance between an exit face of the optical fiber and each light source; anda chromatic-aberration correcting member, provided in the focusing optical system, for correcting aberration of the light of the first wavelength and the second wavelength. 7. A microscope examination apparatus including the optical stimulation apparatus according to claim 4, the microscope examination apparatus comprising:an objective optical system disposed adjacent to the specimen;an illumination optical system that introduces illumination light to the objective optical system; andan eyepiece optical system that allows observation of return light from the specimen;wherein the objective optical system is provided with:a light introducing member that introduces light from the first and second light sources whose optical axes are aligned;an aperture that limits the image of the first and second light sources formed on the specimen; andan aperture moving mechanism that moves the aperture in directions orthogonal to the optical axis.
045044392	description	DETAILED DESCRIPTION OF THE EMBODIMENTS As seen from FIG. 1, the heated gas flows from the core 1 through a hot gas channel 2 into the loop 3. Between the hot gas channel 2 and the loop 3 an apertured plate 4 is arranged, which diverts the hot gas by 90.degree.. In the lower area of the loop 3 a mixing means 5 is arranged according to the invention comprising a plurality of tubes 6, 7, 8, 9 traversing the apertured plate 4 and having their distributor heads 10, 11, 12, 13 above the apertured plate 4. Over the apertured plate 4, the cold gas coming from the mixing device is intermixed with the hot gas. The mixed gas subsequently flows through a heat absorbing component 14, wherefrom it is suctioned by a blower 15 and conducted back through a cold gas line 16 into the core 1. In this embodiment, the gap between the heat absorbing component 14 and the loop 3 is chosen as the cold gas conduit 16. When the valve 17 is opened, the cold gas flows through the mixing device. The flow of the gas is represented by arrows. FIG. 2 shows a mixing means 5 arranged in a loop 3. The valves 18, 19, 20, 21 of the mixing means 5 are arranged in the partition 22 and connected by means of rods 23, 24, 25, 26 with the regulator 27. The tubes 6, 7, 8, 9 are secured in a gas tight manner in the partition 28 and are passing through the apertured plate 4, whereby the distributor heads 10, 11, 12, 13 are arranged above the apertured plate 4. The mixing of the gases takes place above the apertured plate 4. FIG. 3 shows a mixing means 5 with its tubes arranged under the apertured plate 4 and in the hot gas channel 2. The cold gas flows from a mixing means 5 in a direction opposite to the flow direction of the hot gas. The valve 17 is arranged outside the loop 3, thereby establishing a connection between the cold gas line 16 and the valve 17, through the line 16a. FIG. 4 shows a mixing means 5 with its distributor heads arranged above the apertured plate 4 and placed orthogonally to the direction of flow of the hot gas. The mixing of the cold gas with the hot gas takes place under the apertured plate 4. In FIG. 5, a part of the tube 29 is shown with a distributor head 30. The distributor head 30 has several orifices 31, 32 arranged with their principal axes vertically to the surface 35.
	description	The invention relates to a lithium fluoride doped with a divalent positive ion, to a single crystal of said fluoride and to the use of said single crystal as an X-ray monochromator, especially in X-ray fluorescence analysis machines, in X-ray diffraction machines, in electron microprobes and in transmission microscopes. All these analytical techniques employ a monochromator crystal. This crystal receives the X-radiation that it is desired to analyze and separates the various components (or wavelengths) that it contains by diffraction according to Bragg's law and these components leave the crystal at specific angles. This separation results in a set of diffraction lines. A detector placed in the path of the line at a suitable angle determined by Bragg's law converts the X-radiation into an electrical signal. Within the context of the present invention, the X-radiation is understood to mean an electromagnetic wave having an energy between 0.1 and 1000 keV, more particularly between 1 and 100 keV. In the aforementioned analytical machines, the X-radiation may be of various origins. To take an example, the elemental analysis of a specimen by X-ray fluorescence is a nondestructive method based on the detection and analysis of the X-rays that are emitted by said specimen and are then collected by a monochromator crystal that diffracts said X-rays according to Bragg's law. According to that method, the specimen is irradiated by a beam of high-energy X-rays (in the case of an X-ray fluorescence spectrometer) or by a beam of electrons (in the case of a microprobe that may, for example, be incorporated into a scanning electron microscope). This primary beam excites the specimen, which then emits a secondary beam of X-rays, also called X-ray fluorescence. This X-ray fluorescence secondary radiation contains wavelengths characteristic of the chemical elements contained in the specimen. The monochromator crystal separates the various components that it contains by diffraction and they leave the crystal at specific angles. This separation results in a set of diffraction lines. A detector placed in the path of the line at a suitable angle (according to Bragg's law) converts the X-ray fluorescence radiation into an electrical signal. Specific intensities may be accumulated for each X-ray fluorescence line characteristic of a chemical element contained in the specimen. The chemical concentrations of various elements may thus be determined by reference to a prior calibration. A high X-ray sensitivity is expected of such an analytical system, resulting, in the case of X-ray fluorescence, in a high detectability with the best possible precision for detecting very small quantities of an element. The sensitivity is higher the more intense the X-radiation reaching the detector. In the case of X-ray fluorescence, this intensity depends, of course, on the specimen itself, on the targeted chemical element and on the chosen fluorescence line, but also on the monochromator and on the detector. This is because the monochromator may reflect the X-radiation to a greater or lesser extent. Whatever the type of analysis, the aim is to have monochromators that reflect as much as possible, in order to lose the least possible amount of intensity during the diffraction step. The quality of the detector is also important, as there is no point using a highly reflective monochromator sending very intense radiation back to the detector if the latter is capable of measuring it. The detector generally comprises a scintillator and a photoreceiver. The scintillator converts the X-ray energy lost on ionizing into light pulses. The light pulses are received by a photoreceiver, which converts them into an electrical signal. The photoreceiver is usually a photomultiplier tube (or PMT) or a photodiode or other device. The X-radiation striking the scintillator is saturated above a certain intensity (i.e. after a certain number of counts received for a given period of time). This is because, after each X-ray detection (that is to say each time a pulse is counted), the scintillator has a decay time over which any other detection is impossible. A common scintillator such as thallium-doped sodium iodide (Tl:NaI) has a decay time of about 230 ns. This time is determined by fitting a sum-of-exponentials (or integral-of-exponentials) law to the scintillator signal as a function of time. By convention and throughout the rest of the text, only the decay time will be relative to the predominant light component given. With such a scintillator, the maximum counting rate is about 4 million interactions per second. In practice, this value is generally reduced by a factor of 2. In many cases, owing to long decay time of the scintillator, a filter has to attenuate the intensity of the X-radiation. According to the invention, it has been discovered, firstly, that a single crystal of LiF doped with a divalent positive ion, used as monochromator, has a high reflectivity and, secondly, that the strong radiation emanating from the monochromator can be effectively received by a fast light scintillator, for example of the rare-earth halide type. The lithium fluoride used in the context of the invention contains at least 0.014 mol and preferably at least 0.018 mol per kg of a divalent positive ion M present in the fluoride state. The ion M is present in fluoride form, that is to say MF2, in the lithium fluoride LiF. The M contents are given in mols of M (and not in mols of MF2) per total weight in kilograms of doped fluoride, that is to say per kilogram of fluoride containing Li and M (and not pure LiF). Preferably, the atomic number of M is from 10 to 35. Preferably, the ionic radius of divalent M varies from 55 to 80 picometers. The ion M is such that MF2 exists. The ion M may especially be Mg+, Co2+ or Zn2+. The ion M may also be a mixture of at least two ions chosen from Mg2+, Zn2+ and Co2+. The ion M is preferably Mg2+. The table below gives a few characteristics of these ions. Mg2+Co2+Zn2+Atomic Number122730Ionic radius (in657274picometers) The following table gives the equivalences between M contents expressed in mol/kg and Mg2+, Co2+ or Zn2+ contents expressed in ppm by weight. MMg2+Co2+Zn2+(mol/kg)(ppm by weight)(ppm by weight)(ppm by weight)0.018438106011770.020486117913080.023559135515040.025608147316350.0451094265229420.082199348325361 The concentrations of M may be analyzed by ICP spectroscopy (induction-coupled plasma spectroscopy). Preferably, the fluoride contains at least 0.02 mol and even at least 0.023 mol and even at least 0.025 mol of M per kg of fluoride. The fluoride generally contains at most 0.082 mol and even more generally at most 0.045 mol of M per kg of fluoride. If the fluoride according to the invention contains too much M (more than 0.045 mol of M per kg), the single crystal may become brittle and cracks may be observed. The LiF may be manufactured in the single-crystal state from pure LiF and pure MF2 (for example MgF2, CoF2 or ZnF2) powders. The powders are placed in a crucible compatible with its contents, generally a platinum or graphite crucible. The whole assembly is then heated until the powders melt, generally at between 800 and 1000° C., more particularly to above the melting point of LiF, which is about 870° C., the materials then undergoing congruent crystallization resulting in a single crystal or a few large single crystals. The crystallization technique may be the Czochralski, Kyropoulos or Bridgeman-Stockbarger method. The latter technique generally results in a polycrystal containing large single crystals (single-crystal volume of the order of 1 to 10 cm3). The Czochralski method and the Kyropoulos method lead to single crystals and involve a seed. The seed may be pure LiF or LiF doped with M. The material obtained by these growth methods is then used to obtain single crystals generally having the form of a cube or parallelepiped, the thickness of which varies from 0.05 mm to 10 mm in thickness and the two main parallel surfaces of which (one being intended to receive and reflect the X-radiation) have an area ranging from 0.5 to 30 cm2. These single-crystal components may be made from the material coming directly from the growth, for example by cleaving (essentially along the (200) crystal plane). When the material is used in a sequential spectrometer, a parallelepiped with a thickness ranging from 1 to 10 mm is generally prepared, the surface of which may be obtained by cleaving or more generally by mechanical erosion with an abrasive or by chemical-mechanical erosion. When the material is used in a simultaneous spectrometer, then parallelepipedal, and generally cleaved, thin plates ranging in thickness from 0.05 to 1 mm are generally prepared, on which in general a concave shape is imposed by application to a concave support. In this case, the monochromator also has a focusing action. Thus, within the context of the invention, individual single crystals (not agglomerated with another single crystal), the volume of which ranges from 2.5×10−3 cm3 to 30 cm3 and more generally from 0.01 to 20 cm3, are prepared and used. It has been observed that the intensity reflected by the M:LiF single crystal according to the invention (especially when M is Mg) increases very substantially when the wavelength of the reflected line decreases, especially for wavelengths below 3 Å, and even below 2 Å and even below 1.5 Å. The invention also relates to a method of analysis for an element using an analytical machine that includes a monochromator made of the fluoride according to the invention and to a scintillator coupled to said monochromator, said scintillator being locked onto a line of wavelength below 3 Å, or below 2 Å, or below 1.5 Å. The increase in intensity reflected by the M:LIF single crystal according to the invention is particularly spectacular when the M content increases, in particular at short wavelengths. This wavelength effect is more particularly observed for a cleaved surface finish. For a surface finish prepared by mechanical erosion (for example using an abrasive such as silicon carbide, boron carbide or diamond) or even by chemical-mechanical erosion, an increase in the intensity reflected by the M:LIF single crystal according to the invention is also observed (but this is less spectacular than in the case of the cleaved surface) in particular at short wavelengths. The increase in intensity (for the (200) crystal plane) is maintained after the plane single-crystal plates have been curved, especially in the case of a cleaved surface finish. Without the Applicant being held to any theoretical consideration, the X-ray reflectivity properties could be attributed to insertions or substitutions of M (as Mg2+) in the cation lattice. The invention therefore relates to LiF doped with at least one divalent ion M such as Mg2+ having an ionic radius close to that of Li+ (60 picometers), in particular Mg2+, Co2+ and Zn2+. These ions offer the advantage of an atomic number that is still low (therefore providing a lower X-ray absorption) and have, in the form of fluorides (especially MgF2, CoF2 and ZnF2), physical properties that are compatible with the melting of LiF (melting points: 1200° C. and 872° C., respectively; boiling points: 1400° C. and 1500° C., respectively). The invention also relates to the combination of the M:LiF single crystal according to the invention as monochromator with a detector comprising a fast scintillator (delay time less than 30 ns on its principal component) and allowing counts of at least 10 million per second to be achieved. It is also preferable to use a scintillator exhibiting good resolution of its energy spectrum. The energy resolution (ΔE/E) is usually determined, for a given detector and for a given incident energy, as the half-height width of the peak in question in an energy spectrum obtained from this detector, with respect to the energy of the centroid of the peak (see especially: G. F. Knoll “Radiation detection and measurement”, John Wiley & Sons, Inc., 2nd edition, page 114). This combination according to the invention increases the number of X-ray fluorescence photons analyzed. The analysis statistics are therefore improved. The result, in the case of the analytical machine, is better analytical quality and a reduction in measurement time. As suitable scintillator, it is possible to use a polycrystalline or single-crystal material containing a rare-earth halide. These crystals have the advantage of having both a short decay time (for example 28 ns in the case of La0.9Ce0.1Cl3, a figure obtained by fitting to a simple exponential model) and good energy resolution (3.9% with 137Cs). As rare-earth halides more particularly concerned, mention may especially be made of: ALn2X7 in which Ln represents one or more rare earths, X represents one or more halogen atoms, chosen from Cl, Br or l, and A represents an alkali metal such as Rb and Cs; LaCl3, which may in particular be doped with 0.1 to 50 wt % CeCl3; LnBr3, which may in particular be doped with 0.1 to 50 wt % CeBr3; LaBr3, which may in particular be doped with 0.1 to 50 wt % CeBr3; GdBr3, which may in particular be doped with 0.1 to 50 wt % CeBr3; LaxLn(1-x)X3, which may in particular be doped with 0.1 to 50% CeX3, where x may range from 0 to 1, Ln being a rare earth different from La, and X being a halogen as mentioned above; LaxGd(1-x)Br3, which may in particular be doped with 0.1 to 50 wt % CeBr3, it being possible for x to range from 0 to 1; LaxLu(1-x)Br3, which may in particular be doped with 0.1 to 50 wt % CeBr3, it being possible for X to range from 0 to 1; Ln′xLn″(1-x)X′3(1-y)X″3y, in which Ln′ and Ln″ are two different rare earths of the Ln type, X′ and X″ being two different halogens of the X type, especially Cl and Br, it being possible for x to range from 0 to 1 and for y to range from 0 to 1; RbGd2Br7, which may in particular be doped with 0.1 to 50 wt % CeBr3; RbLn2Cl7, which may in particular be doped with 0.1 to 50 wt % CeCl3; RbLn2Br7, which may in particular be doped with 0.1 to 50 wt % CeBr3; CsLn2Cl7, which may in particular be doped with 0.1 to 50 wt % CeCl3; CsLn2Br7, which may in particular be doped with 0.1 to 50 wt % CeBr3; K2LaCl5, which may in particular be doped with 0.1 to 50 wt % CeCl3; K2Lal5, which may in particular be doped with 0.1 to 50 wt % Cel3; and Lul3, which may in particular be doped with 0.1 to 50 wt % Cel3. The term “dopant” or the term “doped” refers to a rare earth present in a minor proportion) that substitutes for one or more rare earths present in a major proportion, the rare earths present in both minor and major proportions being included by the symbol Ln. As preferred rare-earth halide, mention may be made of: LaBr3 doped with 5 to 15 wt % CeCl3; and LaCl3 doped with 5 to 15 wt % CeCl3. The invention is not limited to the use of a Tl:NaI or lanthanum halide crystal as detector. Detectors giving good energy resolution (especially over a wide energy range) and/or a good response time (in particular less than 100 ns) may usefully be employed in combination with the LiF crystal according to the invention. Such crystals may for example be YAP (yttrium aluminum perovskite) especially Ce( )-doped or YAG (yttrium aluminum garnet) or Ge (germanium). Single crystals of pure LiF or those doped with Mg in fluoride form were prepared from a uniform blend of pure LiF and pure MgF2 powders in various concentrations. The blend was placed in a platinum crucible and then melted by heating to 950° C. A crystallization operation was then carried out, resulting in a single crystal 2300 cm3 in volume. The reflected X-radiation intensity was measured on single-crystal pieces cleaved along the (200) plane, with the molybdenum Kα line. The intensity varied little within an Mg concentration range between 0 and 400 ppm by weight. The intensity from the highly magnesium-doped Mg:LiF single crystals was expressed as a percentage of the intensity from the specimen containing 300 ppm Mg by weight. These results are given in Table 1. In this table, the specimen name contains the Mg content. TABLE 1Mg content (inMg content% intensitySpecimenppm by weight)(mol/kg)relative to LiF300pure LiF00100%LiF3003000.0123100%LiF5015010.0206336%LiF6646640.0273476%LiF7657650.0314552%LiF106310630.0437589% The effect of wavelength on the reflected intensity was also measured on specimens having a different Mg content. Table 2 gives the results. These results are expressed as a percentage of the reflected intensity for LiF300. The reflected intensity increases very strongly when the wavelength decreases for specimens having a higher Mg content. TABLE 2λ =λ =λ = 3.359 Åλ = 2.750 Åλ = 1.937 Å1.542 Å0.7107 ÅCaTiFeCuMoLiF300100%100%100%100%100%LiF765207%252%358%410%572%LiF1063238%285%404%441%595% The increase in intensity, measured with a cleaved surface finish (on a (200) crystallographic plane), is maintained after plane plates have been curved. For example, at the iron wavelength (λ=1.937 Å), the intensity reflected by the cleaved plane LiF664 was 2.8 times higher than that of cleaved plane LiF300. The intensity reflected by curved LiF664 plates, curved over a cylinder whose axis was parallel to the X-ray direction, remained higher than the intensity reflected by the curved LiF300 plates, again curved over the same cylinder with its axis parallel to the direction of the X-rays. The intensity ratio of the plates curved over a cylinder of axis parallel to the X-ray beam remained the same (i.e. 2.8).
056152396	abstract	A core differential pressure and liquid control line apparatus for a nuclear reactor is described. The apparatus includes a first portion configured to be positioned within and extend through an opening in the pressure vessel wall. At least the first tube portion has a diameter less than the diameter of the opening in the pressure vessel wall. An annulus is formed between the exterior surface of the first tube portion and the pressure vessel wall so that a neutron absorbent can be injected into the pressure vessel at the location of the annulus. The apparatus further includes a second L-shaped tube portion configured to be coupled to the first tube portion, and a third tube portion configured to be coupled to the second tube portion. The open end of the third tube portion extends to an elevation above the core plate.
	description	The present invention relates to a sintered nuclear fuel pellet including a burnable absorber capable of controlling self-shielding and a manufacturing method thereof. A nuclear reactor is a reactor that uses the nuclear fission of nuclear fuel to get energy. In the nuclear reactor, energy is generated by the chain reaction of a series of nuclear fissions induced by the neutrons emitted during the nuclear fission. At this time, it is necessary to control properly the reactor core reactivity and the nuclear reactor power distribution in order to operate the nuclear reactor safely and economically. In a general nuclear reactor, the reactivity is controlled by inserting or withdrawing a control rod made of a neutron absorbing material mechanically. Control of the reactor core reactivity by using such a control rod has the advantage of quick control of the reactor core reactivity. However, such a control rod is locally inserted into the reactor core, so that the insertion of the control rod itself is not enough to control the reactor core reactivity and the power distribution simultaneously. When the excess reactivity is high, the mechanical movement of the control rod has to grow, and accordingly the uncertainty of controlling the reactivity is increased, indicating the risk of accident is also increased. So, in the conventional nuclear reactor, other methods are used to lower the excess reactivity and then control the reactivity using a control rod. It is one of the well known methods in this field to mix a coolant with a boric acid solution, a neutron absorbing material. When the reactor core reactivity is controlled by regulating the concentration of boron included in the coolant, the reactivity can be controlled while minimizing the distortion of the reactor core power distribution since boron is evenly distributed in the coolant of the nuclear reactor. However, distribution of boron in the coolant and the following dilution take a long time. So, if a quick control of the reactor core reactivity is requested, the method using boric acid to control the reactor core reactivity is not able to be used. Another problem of using boron to control the reactor core reactivity is that a massive radioactive liquid waste is generated in the course of lowering the concentration of boron. In order to regulate the concentration of water-soluble boron in the primary coolant system, such an expensive device as ‘chemical and volume control system (CVCS)’ is required. To neutralize pH of the coolant containing boron (boric acid solution), which is acidic, LiH has to be added to the coolant. At this time, LiH reacts with neutrons to generate large amounts of tritium and the boric acid solution itself causes corrosion of structural materials and nuclear fuel cladding tubes forming the primary coolant system of the nuclear reactor, resulting in the deterioration of operation performance of the nuclear reactor. If the concentration of boron in the coolant is very high, the coolant temperature coefficient might be very close to zero or positive, which is not preferred for the viewpoint of safety. Therefore, overcoming the problem above has been an issue related to the safety of the pressurized light water reactor. It is very natural to consider non-boric acid or low-boric acid operation to improve the safety of nuclear reactors. So, a burnable absorber has been generally tried. The burnable absorber acts as a strong neutron absorber. However, once it absorbs neutrons and converted to other nuclides, the neutron absorption cross section is reduced significantly. The most representative burnable absorbers are exemplified by gadolinium (Gd), erbium (Er), and boron (B). In pressurized light water reactors, gadolinium (Er) and erbium (Er) are properly mixed with UO2 nuclear fuel in the form of Gd2O3 and Er2O3. In the meantime, erbium, one of the rare-earth burnable absorbers, is not frequently used because it has a relatively high residual toxic activity, but it can be used as a relatively efficient burnable absorber in a core having a long cycle length. Gadolinium has a low residual toxic activity, so that it is frequently used as a burnable absorber. When gadolinium is mixed with nuclear fuel, the thermal conductivity of the nuclear fuel decreases, suggesting that the power density of the nuclear fuel mixed with gadolinium is very low. Therefore, when gadolinium is used, it is hard to use a large amount of gadolinium. Also, when the nuclear fuel is mixed with Gd2O3, gadolinium is fast burned, indicating that it is hardly used for such a nuclear reactor that has a long core cycle length. Boron is used as IFBA (Integrated Fuel Burnable Absorber), wherein ZrB2 is coated very thinly on UO2 fuel rod. Boron is also used in another way, wherein a boron compound such as B4C may is prepared in a special shape and loaded in the control rod guide tube. A typical concept thereof is WABA (Wet Annular Burnable Absorber). When boron absorbs neutrons, helium gas is generated. Therefore, it is difficult to mix boron with nuclear fuel, so it is used in the same way of IFBA or WABA. The neutron absorption cross section of boron is relatively small, so when it is used in the IFBA form, a comparatively large number of fuel rods must be loaded with IFBA. When it is used in the WABA form, so that it is loaded in the control rod guide tube as a burnable absorber, the insertion of the control rod might be limited. Thus, when boron is used in the manner of WABA, the use of a burnable absorber is limited. As described hereinbefore, various types of burnable absorbers have been used in nuclear fuel assemblies, but they are used in a limited manner in their use. Therefore, it is an object of the present invention to provide a novel burnable absorber that can control the core reactivity and power distribution more efficiently. To overcome the problems above, the present invention provides a sintered nuclear fuel pellet that can be loaded in the conventional nuclear reactors without major design changes and can optimize the excess reactivity to match the characteristics of each nuclear reactor. The present invention also provides a manufacturing method of the sintered nuclear fuel pellet capable of preventing the occurrence of cracks due to the difference in thermal expansion and phase transformation during sintering and use and solving unevenness in the degree of combustion of nuclear fuel. The objects that the present inventors try to solve are not limited to the above and it is clearly understood by those in the art that other objects supported by the following descriptions can also be included. The present invention relates to a sintered nuclear fuel pellet wherein a consolidated body of a burnable absorber is inserted. In a preferred embodiment of the present invention, the consolidated body of the burnable absorber can be included in one or more forms selected from the group consisting of amorphous, cylinder, disk, spherical, rod, film, and polyprism. In a preferred embodiment of the present invention, the consolidated body of the burnable absorber can be inserted in the sintered nuclear fuel pellet at the volume of up to 50% by the total volume of the sintered nuclear fuel pellet. In a preferred embodiment of the present invention, the consolidated body of the burnable absorber can be inserted either or both in the center area and the surface area of the sintered nuclear fuel pellet. According to a preferred embodiment of the present invention, a single or multiple green bodies of the burnable absorber can be inserted in the sintered nuclear fuel pellet, and when multiple green bodies are inserted, they can be same or different in the shape and the size. In a preferred embodiment of the present invention, the consolidated body of the burnable absorber includes Gd2O3; Er2O3; Gd2O3 stabilized with one or more selected from the group consisting of CeO2, In2O3, Y2O3, UO2, ThO2, TiO2, ZrO2, Al2O3, and Y2O3—ZrO2 (yttria-stabilized zirconia); Er2O3 stabilized with one or more selected from the group consisting of CeO2, In2O3, Y2O3, UO2, ThO2, TiO2, ZrO2, Al2O3, and Y2O3—ZrO2 (yttria-stabilized zirconia); or a combination thereof. In a preferred embodiment of the present invention, the content of CeO2, In2O3, Y2O3, UO2, ThO2, TiO2, ZrO2, Al2O3, and Y2O3—ZrO2 can be 1˜60 weight % by the volume of Gd2O3 and Er2O3. In a preferred embodiment of the present invention, the consolidated body of the burnable absorber can be inserted at the volume of 2˜20% by the total volume of the sintered nuclear fuel pellet. In a preferred embodiment of the present invention, the consolidated body of the burnable absorber can be included in one or more forms selected from the group consisting of amorphous, cylinder, disk, spherical, rod, film, and polyprism. According to a preferred embodiment of the present invention, the manufacturing method of the sintered nuclear fuel pellet with the consolidated body of the burnable absorber inserted therein comprises the steps of forming the consolidated body of the absorber; mixing the consolidated body and the nuclear fuel and then inserting the consolidated body in the nuclear fuel; forming the nuclear fuel-consolidated body composite by press-molding the mixture of the nuclear fuel and the consolidated body; and forming the sintered nuclear fuel pellet by sintering the nuclear fuel-consolidated body composite. In a preferred embodiment of the present invention, the step of forming the consolidated body of the burnable absorber can include the substeps of homogenizing and mixing the burnable absorber powder and press-molding the mixed burnable absorber powder. In a preferred embodiment of the present invention, the step of press-molding can be composed of the first press-molding at the pressure of 30 MPa˜100 MPa and the second press-molding at the pressure of 200 MPa˜500 MPa by an hydrostatic press (CIP). In a preferred embodiment of the present invention, the step of press-molding can be performed at the pressure of 100 MPa˜500 MPa just one time. In a preferred embodiment of the present invention, the step of forming the consolidated body of the burnable absorber can include the substeps of homogenizing and mixing the burnable absorber powder, injecting the burnable absorber powder into the hole of the hollow green body, and pressing the inside of the hole. In a preferred embodiment of the present invention, the hollow green body can include one or more materials selected from the group consisting of CeO2, In2O3, Y2O3, UO2, ThO2, TiO2, ZrO2, Al2O3, and Y2O3—ZrO2. In a preferred embodiment of the present invention, the step of pressing can be uniaxial pressing the inside of the hole at the pressure of 30 MPa˜100 MPa. In a preferred embodiment of the present invention, the hollow green body is in the shape of ring or tube, and the volume of the hollow green body is 2˜20 volume % by the total volume and the diameter of the hollow is 1˜3 mm. In a preferred embodiment of the present invention, the consolidated body of the burnable absorber includes Gd2O3; Er2O3; Gd2O3 stabilized with one or more selected from the group consisting of CeO2, In2O3, Y2O3, UO2, ThO2, TiO2, ZrO2, Al2O3, and Y2O3—ZrO2 (yttria-stabilized zirconia); Er2O3 stabilized with one or more selected from the group consisting of CeO2, In2O3, Y2O3, UO2, ThO2, TiO2, ZrO2, Al2O3, and Y2O3—ZrO2 (yttria-stabilized zirconia); or a combination thereof. In a preferred embodiment of the present invention, the content of CeO2, In2O3, Y2O3, UO2, ThO2, TiO2, ZrO2, Al2O3, and Y2O3—ZrO2 (yttria-stabilized zirconia) can be 3˜30 weight % by the volume of Gd2O3 and Er2O3. In a preferred embodiment of the present invention, the step of forming the consolidated body of the burnable absorber can additionally include the step of heat-treating the consolidated body of the burnable absorber in a reducing gas atmosphere or an inert gas atmosphere at the temperature of 100° C.˜400° C. In a preferred embodiment of the present invention, the step of forming the nuclear fuel-consolidated body composite can be composed of the first press-molding at the pressure of 30 MPa˜100 MPa and the second press-molding at the pressure of 200 MPa˜500 MPa by a hydrostatic press (CIP). In a preferred embodiment of the present invention, the step of forming the nuclear fuel-consolidated body composite can be performed at the pressure of 100 MPa˜500 MPa just one time. In a preferred embodiment of the present invention, the step of forming the sintered nuclear fuel pellet can be performed by sintering at 1000° C.˜1800° C. via microwave sintering. In a preferred embodiment of the present invention, in the step of inserting the consolidated body in the nuclear fuel material, the volume of the consolidated body of the burnable absorber is 1˜10 weight % by the total weight of the nuclear fuel material. According to the present invention, the excess reactivity of a nuclear reactor can be reduced efficiently by inserting the centrally-shielded burnable absorber (CSBA) in the center of the sintered nuclear fuel pellet. The present invention can optimize the surface area of the burnable absorber to control the combustion rate and self-shielding of the burnable absorber in the nuclear fuel properly. The present invention can optimize the performance of the nuclear fuel by controlling the size, shape and position of the burnable absorber according to the output and the lifetime of the nuclear fuel. The present invention can control the position of the consolidated body of the burnable absorber in order to prevent the occurrence of cracks caused by the difference in thermal expansion and phase transformation during sintering and use of the sintered nuclear fuel pellet and to maximize the fuel efficiency. The present invention is efficient in increasing the lifetime and output of the nuclear fuel by regulating the shape, the size and the volume ratio of the consolidated body of the burnable absorber according to the lifetime and output of the nuclear fuel. The examples of the present invention are described in more detail with the figures attached hereinafter. In the following description of the present invention, a detailed description of known functions and configurations incorporated herein will be omitted when it may make the subject matter of the present invention rather unclear. The terminologies used herein are the terms used to properly express the preferred embodiments of the present invention, which may vary depending on the user, the intention of the operator, or the practice of the field to which the present invention belongs. Therefore, the definitions of these terms should be based on the contents throughout this specification. The same reference symbols used in each figures indicate the same elements. The present invention relates to a sintered nuclear fuel pellet. According to a preferred embodiment of the present invention, the sintered nuclear fuel pellet of the present invention can regulate the excess reactivity efficiently and can control the reaction speed and self-shielding of the burnable absorber variably according to the condition of the nuclear fuel. The present invention relates to a sintered nuclear fuel pellet. According to a preferred embodiment of the present invention, the sintered nuclear fuel pellet can control the volume ratio, the shape, the location of insertion, the size and the number of the burnable absorber variably in order to maximize the performance of the nuclear fuel by applying the consolidated body of the burnable absorber. According to a preferred embodiment of the present invention, the sintered nuclear fuel pellet can contain the consolidated body of the burnable absorber inserted therein. The consolidated body of the burnable absorber is the kind of a consolidated body prepared by bulk-processing of the burnable absorber, and can be inserted by adjusting its shape freely to exhibit its own shielding property and minimize the excess reactivity. In an example of the present invention, the consolidated body of the burnable absorber can be porous, which is advantageous for keeping the gas generated by nuclear fission inside the porous burnable absorber and preventing the gas from leaking out of the nuclear fuel even if the burnable absorber is melted. In an example of the present invention, the volume ratio of the consolidated body of the burnable absorber in the sintered nuclear fuel pellet can be adjusted properly for the optimization of the lifetime and output of the nuclear fuel by regulating the combustion speed and self-shielding. The preferable volume of the consolidated body above in the sintered nuclear fuel pellet is up to 50 volume %, more preferably 2˜30 volume %, 2˜20 volume %, or 2˜10 volume % by the total volume of the sintered nuclear fuel pellet. In an example of the present invention, the excess reactivity can be optimized by regulating the combustion speed and self-shielding according to the shape, the number, the insertion position, or the size of the consolidated body of the burnable absorber. For example, the shape of the consolidated body of the burnable absorber above can be one or more shapes selected from the group consisting of amorphous, cylinder, disk, spherical, rod, film, and polyprism. Preferably, the shape of sphere, amorphous, and cylinder can be selected. The combustion speed and self-shielding of the burnable absorber can be regulated according to the shape of the consolidated body of the burnable absorber above. In a preferred embodiment of the present invention, the consolidated body of the burnable absorber can be inserted singly or plurally. For example, when multiple green bodies are inserted, the green bodies of the burnable absorber can be inserted with the same or different shape and size. When multiple green bodies are inserted, the combustion speed of the burnable absorber can be controlled. As shown in FIG. 1 A, which illustrates the sintered nuclear fuel pellet of the present invention, only one spherical consolidated body of the burnable absorber is inserted in order to slow down the combustion of the burnable absorber by maximizing the self-shielding effect. On the contrary, if it is needed to speed up the combustion of the burnable absorber by minimizing the self-shielding effect, three spherical green bodies of the burnable absorber in the same volume, like 3 spherical balls, are preferably inserted. In a preferred embodiment of the present invention, the consolidated body of the burnable absorber can be inserted throughout the sintered nuclear fuel pellet or inserted in the central region, the surface region, or both regions of the sintered nuclear fuel pellet. According to the location of the insertion of the consolidated body of the burnable absorber, the combustion speed and self-shielding can be regulated. When the consolidated body of the burnable absorber is inserted in the center of the sintered nuclear fuel pellet, the increase of temperature according to the excessive nuclear fission in the center of the nuclear fuel can be prevented and the release of the burnable absorber out of the nuclear fuel can be prevented even when the burnable absorber is melted first before the nuclear fuel is melted. As shown in FIG. 1 B, the consolidated body of the burnable absorber can be in various shapes such as spherical ball core, cylindrical core, surface mini-pellets, mini-pellet core, etc, and the consolidated body can be inserted in different locations. In a preferred embodiment of the present invention, the size of the consolidated body of the burnable absorber can be up to 10 mm; up to 1 mm, 1 μm˜1.5 mm; or 1 μm˜900 μm. The size of the consolidated body herein indicates the diameter, radius, or height. In a preferred embodiment of the present invention, the burnable absorber can be applied without limitation if it is a material that provides a self-shielding effect applicable to nuclear fuel. For example, the burnable absorber can include one or more materials selected from the group consisting of Gd2O3; Er2O3; and Gd2O3 or Er2O3 stabilized with CeO2, In2O3, Y2O3, UO2, ThO2, TiO2, ZrO2, Al2O3, and Y2O3—ZrO2 (yttria-stabilized zirconia). Preferably, the burnable absorber can include Gd2O3; and Gd2O3 stabilized with ZrO2 or Y2O3—ZrO2 (yttria-stabilized zirconia). For example, at least one of CeO2, In2O3, Y2O3, UO2 ThO2, TiO2, ZrO2, Al2O3 and Y2O3—ZrO2 (yttria-stabilized zirconia) can be included in Gd2O3 and/or Er2O3 at the concentration of 1˜60 weight % or 3˜30 weight %. When the content is within the above range, the phase change is stabilized and the soundness of the sintered body can be improved. In a preferred embodiment of the present invention, the consolidated body of the burnable absorber can be a structure composed of a core and a shell. For example, the consolidated body can include a core comprising either or both of Gd2O3 and Er2O3; and a shell enveloping the core and containing one or more materials selected from the group consisting of CeO2, In2O3, Y2O3, UO2, ThO2, TiO2, ZrO2, Al2O3, and Y2O3—ZrO2 (yttria-stabilized zirconia). The shell can additionally include either Gd2O3 or Er2O3 or both of them. For example, the UO2 above is a nuclear fuel material, which can inhibit cracking caused by thermal expansion matching in the course of sintering of the sintered nuclear fuel pellet. In the manufacturing method of the sintered nuclear fuel pellet according to a preferred embodiment of the present invention, the thermal expansion and phase transformation characteristics can be well regulated in the course of sintering by using the consolidated body of the burnable absorber and also the insertion location, the size and the shape of the consolidated body of the burnable absorber can be properly regulated, indicating that the self-shielding can be regulated easily. According to an example of the present invention, the consolidated body of the burnable absorber can be prepared by press-molding the burnable absorber powder and the nuclear fuel consolidated body can be prepared by inserting the burnable absorber in the nuclear fuel material. Also, the sintered nuclear fuel pellet with the burnable absorber inserted therein can be prepared by sintering the nuclear fuel consolidated body above. For example, the burnable absorber powder can have the particle size of up to 100 μm; 50 μm; or 100 nm˜1 μm. For example, the content of the consolidated body of the burnable absorber can be 1 weight %˜10 weight %; or 2 weight %˜4 weight % by the total weight of the nuclear fuel material. For example, the nuclear fuel consolidated body can be sintered in air, in an inert gas atmosphere, or in a reducing gas atmosphere at 1000° C.˜1800° C.; or at 1300° C.˜1600° C. for 1 minute˜10 hours; 5 minutes˜1 hour; or 10 minutes˜30 minutes. For example, the sintering above can be preferably performed by using a sintering furnace or a microwave sintering apparatus and more preferably by using a microwave sintering apparatus. Practical and presently preferred embodiments of the present invention are illustrative as shown in the following Examples. However, it will be appreciated that those skilled in the art, on consideration of this disclosure, may make modifications and improvements within the spirit and scope of the present invention. The effect of CSBA on the commercial core was tested through computerized calculation. For the test, a nuclear fuel assembly model used in Westinghouse AP1000 core was considered. CSBA presented in FIG. 1 B was compared with IFBA generally used for AP1000 model. The results are shown in FIG. 2. As shown in FIG. 2, better excess reactivity control than IFBA was possible through the optimization of CSBA, and CSBA could be optimized for core design. The present invention provides a sintered nuclear fuel pellet that can regulate the excess reactivity in nuclear fuel efficiently by regulating the shape, the size, and the insertion location of the consolidated body of the burnable absorber. According to a preferred embodiment of the present invention, the sintered nuclear fuel pellet can include the consolidated body of the burnable absorber inserted therein. The consolidated body of the burnable absorber is prepared by bulk-processing of the burnable absorber; displays the self-shielding property; can regulate consumption rate of the burnable absorber when the fuel is used; and accordingly can increase the neutron utilization efficiency. In an example of the present invention, the volume ratio of the consolidated body of the burnable absorber in the sintered nuclear fuel pellet can be adjusted properly for the optimization of the lifetime and output of the nuclear fuel. The preferable volume of the consolidated body above in the sintered nuclear fuel pellet is up to 20 volume %, more preferably 2˜20 volume %, 2˜10 volume %, or 2˜5 volume % by the total volume of the sintered nuclear fuel pellet. When the volume ratio is in that range, the consumption rate of the burnable absorber can be regulated as desired and the thermal expansion and phase transformation characteristics can also be regulated properly, resulting in the prevention of cracking. In a preferred embodiment of the present invention, the combustion speed of the burnable absorber can be regulated according to the shape, the number, the insertion location, and/or the size of the consolidated body of the burnable absorber, which can also be regulated according to the output and performance of nuclear fuel. For example, the shape of the consolidated body of the burnable absorber above can be one or more shapes selected from the group consisting of amorphous, cylinder, disk, spherical, rod, film, and polyprism, and is more preferably spherical or cylinder. When the consolidated body of the burnable absorber is in the shape of sphere, the combustion speed of the burnable absorber can be slow down, so that the self-shielding effect can be minimized. In a preferred embodiment of the present invention, the consolidated body of the burnable absorber can be inserted singly or plurally. For example, when multiple green bodies are inserted, the green bodies of the burnable absorber can be inserted with the same or different shape and size. When multiple green bodies are inserted, the surface area of the burnable absorber is increased, so that the combustion of the burnable absorber can be increased. For example, the consolidated body of the burnable absorber can be inserted throughout the sintered nuclear fuel pellet or inserted in the central region, the surface region, or both regions of the sintered nuclear fuel pellet. According to the insertion location of the consolidated body of the burnable absorber, the performance of nuclear can be optimized. As shown in FIG. 3˜FIG. 6, which illustrate the sintered nuclear fuel pellet of the present invention, wherein FIG. 3 illustrates that the cylinder-shaped consolidated body (Gd2O3) of the burnable absorber was inserted in the nuclear fuel (UO2) and at this time a single or two or three green bodies of the burnable absorber were inserted. These green bodies of the burnable absorber could provide different self-shielding effects according to the number and the size of the inserted green bodies and therefore the combustion speed of the burnable absorber could also be different. FIG. 4 illustrates that the spherical consolidated body (Gd2O3) of the burnable absorber was inserted in the nuclear fuel (UO2) and at this time a single or two and four green bodies of the burnable absorber were inserted. These green bodies of the burnable absorber could provide different self-shielding effects according to the number and the size of the inserted green bodies and therefore the combustion speed of the burnable absorber could also be different. In a preferred embodiment of the present invention, the burnable absorber can be applied without limitation if it is a material that provides a self-shielding effect applicable to nuclear fuel. For example, the burnable absorber can include one or more materials selected from the group consisting of Gd2O3; Er2O3; and Gd2O3 and Er2O3 stabilized with CeO2, In2O3, Y2O3, UO2, ThO2, TiO2, ZrO2, Al2O3, and Y2O3—ZrO2 (yttria-stabilized zirconia). Preferably, the burnable absorber can include Gd2O3; and Gd2O3 stabilized with ZrO2 or Y2O3—ZrO2 (yttria-stabilized zirconia). For example, at least one of CeO2, In2O3, Y2O3, UO2 ThO2, TiO2, ZrO2, Al2O3 and Y2O3—ZrO2 (yttria-stabilized zirconia) can be included in Gd2O3 and/or Er2O3 at the concentration of 1˜60 weight % or 3˜30 weight %. When the content is within the above range, the phase change is stabilized and the soundness of the sintered body can be improved. As shown in FIG. 5, the ZrO2 sintered body containing Gd2O3 particles could be inserted in the hollow UO2 to prepare a composite sintered body. In a preferred embodiment of the present invention, the consolidated body of the burnable absorber can be a structure composed of a core and a shell. For example, the consolidated body can include a core comprising either or both of Gd2O3 and Er2O3; and a shell enveloping the core and containing one or more materials selected from the group consisting of CeO2, In2O3, Y2O3, UO2, ThO2, TiO2, ZrO2, Al2O3, and Y2O3—ZrO2 (yttria-stabilized zirconia). The shell can additionally include either Gd2O3 or Er2O3 or both of them. For example, the UO2 above is a nuclear fuel material, which can inhibit cracking caused by thermal expansion matching in the course of sintering of the sintered nuclear fuel pellet. As shown in FIG. 6, the UO2 sintered body containing Gd2O3 particles could be inserted in the hollow UO2 to prepare a composite sintered body. The present invention relates to manufacturing method of the sintered nuclear fuel pellet. In the manufacturing method of the sintered nuclear fuel pellet according to a preferred embodiment of the present invention, the thermal expansion and phase transformation characteristics can be well regulated in the course of sintering by using the consolidated body of the burnable absorber and also the insertion location, the size and the shape of the consolidated body of the burnable absorber can be properly regulated in order to optimize the efficiency of the nuclear fuel. As shown in FIG. 7, which illustrates the flow chart of the manufacturing method of the sintered nuclear fuel pellet according to the present invention, the manufacturing method of the sintered nuclear fuel pellet comprises the steps of forming the consolidated body of the burnable absorber (S100); inserting the consolidated body in the nuclear fuel material (S200); forming the nuclear fuel-consolidated body composite (S300); and forming the sintered nuclear fuel pellet (S400). In a preferred embodiment of the present invention, the step of forming the consolidated body of the burnable absorber (S100) can include the substeps of homogenizing and mixing the burnable absorber powder (S110) and press-molding the mixed burnable absorber powder (S120). In a preferred embodiment of the present invention; the step of homogenizing and mixing the burnable absorber powder (S110) is a step of crushing and sieving the burnable absorber powders and then mixing thereof. At this time, the burnable absorber powder can have a particle size of up to 100 μm; 50 μm; or 100 nm˜1 μm. In a preferred embodiment of the present invention, the step of press-molding (S120) is a step of injecting the mixed burnable absorber powders obtained in the mixing step (S110) into a mold, followed by press-molding. For example, the step of press-molding (S120) can be composed of the first press-molding at the pressure of 30 MPa˜100 MPa and the second press-molding at the pressure of 200 MPa˜500 MPa by s hydrostatic press (CIP). For example, the step of press-molding (S120) can be performed once via uniaxial pressing at the pressure of 100 MPa˜500 MPa. The press-molding above can be performed at 0° C.˜200° C.; 0° C.—room temperature; or 80° C.˜150° C. In another preferred embodiment of the present invention, the step of forming the consolidated body of the burnable absorber (S100) can include the substeps of homogenizing and mixing the burnable absorber powder (S110′) and pressing the mixed burnable absorber powder (S120′). In a preferred embodiment of the present invention, the step of homogenizing and mixing the burnable absorber powders (S110′) can be performed in the same manner as in the step S110. In a preferred embodiment of the present invention, the step of pressing (S120′) is to prepare the consolidated body of the burnable absorber having the core/shell structure by pressing inside the hole after injecting the burnable absorber powders into the hole of the hollow green body. For example, the step of pressing can be performed via uniaxial pressing inside the hole at the pressure of 30 MPa˜100 MPa. For example, the hollow green body can be a ceramic having a high melting point and a neutron transparency. In an example, the hollow green body can contain one or more materials selected from the group consisting of CeO2, In2O3, Y2O3, UO2, ThO2, TiO2, ZrO2, Al2O3, Y2O3—ZrO2 (yttria-stabilized zirconia), and UO2, and is preferably a material having the same high melting point as ZrO2 and ZrO2. For example, the burnable absorber powder forming the core can include the burnable absorber mentioned before in the description of the sintered nuclear fuel pellet. In an example of the invention, the core and the shell can contain different burnable absorbers and if they contain the same burnable absorber, the mixing ratio can be differently regulated. In a preferred embodiment of the present invention, the hollow green body can be in the shape of a ring or a tube. The volume of the hollow is preferably 2˜20 weight % by the total volume of the hollow green body, and the diameter of the hollow is 1˜3 mm. The reactivity of the consolidated body of the burnable absorber can be appropriately controlled depending on the volume ratio and the diameter of the hollow. In a preferred embodiment of the present invention, the step of forming the consolidated body of the burnable absorber (S100) can additionally include a step of heat-treating the consolidated body of the burnable absorber (S130). For example, the step of heat-treating the consolidated body (S130) can be performed in air, a reducing gas atmosphere, or an inert gas atmosphere at the temperature of 100° C.˜400° C.; or 200° C.˜300° C. for 1 minute˜5 hours; for 5 minutes˜1 hour; or for 10 minutes˜30 minutes. In an example of the invention, the step of heat-treating the consolidated body (S130) is performed by using a sintering furnace or a microwave sintering apparatus. In an example of the present invention, the step of inserting the consolidated body of the burnable absorber into the nuclear fuel material (200) is a step of inserting the consolidated body of the burnable absorber into the nuclear fuel material by mixing the nuclear fuel material and the consolidated body of the burnable absorber in a mold. In this step (200), the insertion location and the number of the consolidated body of the burnable absorber can be regulated. In an example of the present invention, the consolidated body of the burnable absorber can be mixed with the nuclear fuel material at the volume of 1˜10 weight % or 2˜4 weight % by the total weight of the nuclear fuel material. In an example of the present invention, the step of forming the nuclear fuel-consolidated body composite (S300) is a step of forming the nuclear fuel-consolidated body composite by press-molding the mixture of the consolidated body and the nuclear fuel after the step of inserting the consolidated body (200). For example, the press-molding (S120) can be composed of the first press-molding performed at the pressure of 30 MPa˜100 MPa and the second press-molding performed at the pressure of 200 MPa˜500 MPa by s hydrostatic press (CIP). For example, the press-molding can be performed once at the pressure of 100 MPa˜500 MPa. The press-molding above can be performed at 0° C.˜200° C.; 0° C.—room temperature; or 80° C.˜150° C. In a preferred embodiment of the present invention, the step of forming the sintered nuclear fuel pellet (S400) is a step of sintering the nuclear fuel-consolidated body composite to form the sintered nuclear fuel pellet. In an example of the present invention, the step of forming the sintered nuclear fuel pellet (S400) is performed by sintering in air, in an inert gas atmosphere, or in a reducing gas atmosphere at 1000° C.˜1800° C.; or at 1300° C.˜1600° C. for 1 minute 10 hours; 5 minutes 1 hour; or 10 minutes 30 minutes. For example, the step of forming the sintered nuclear fuel pellet (S400) above can be preferably performed by using a sintering furnace or a microwave sintering apparatus and more preferably by using a microwave sintering apparatus. (1) Preparation of Consolidated Body The mixed powders of ZrO2-8 mol. % Y2O3(Sigma Aldrich. 99.9%, 8YSZ) and Gd2O3(Sigma Aldrich. 99.9%, <10 μm) were pressurized at 33.5 MPa, and then CIP was performed at 400 MPa. As a result, Gd2O3, Gd2O3-5 wt % 8YSZ, Gd2O3-10 wt % 8YSZ, Gd2O3-20 wt % 8YSZ, Gd2O3-30 wt % 8YSZ, and Gd2O3-40 wt % 8YSZ pellets were obtained. (2) Preparation of Sintered Body The obtained pellets were sintered in a microwave sintering apparatus at 1400° C., 1500° C., and 1600° C. for 20 minutes each. As a result, Gd2O3-5 wt % 8YSZ, Gd2O3-10 wt % 8YSZ, Gd2O3-20 wt % 8YSZ, Gd2O3-30 wt % 8YSZ, and Gd2O3-40 wt % 8YSZ sintered bodies were obtained. The XRD patterns of those sintered bodies were measured and shown in FIG. 8. As shown in FIG. 8, when yttria-stabilized zirconia was added to gadolinia at the concentration of 5 wt %˜40 wt %, gadolinia was stabilized in a cubic phase from the concentration of 10 wt %. Since the phase change does not occur on the cubic phase to the monoclinic phase during sintering, the soundness of the sintered body can be improved Gd2O3 (Sigma Aldrich. 99.9%, <10 μm) powder was molded by uniaxial pressing at a low pressure of 30˜50 MPa and then crushed into large lumps, which were rotated in a ball mill to prepare amorphous particles through granulation. The size of the assembled Gd2O3 particles was hundreds μm˜several mm. The mixed powders of the assembled Gd2O3 particles and ZrO2-8 mol. % Y2O3 (Sigma Aldrich. 99.9%, 8YSZ) were pressurized at 33.5 MPa, and then CIP was performed at 400 MPa. As a result, Gd2O3, 30 wt % Gd2O3-70 wt % 8YSZ, and 50 wt % Gd2O3-50 wt % 8YSZ pellets were obtained. The pellets and ZrO2-8 mol. % Y2O3(Sigma Aldrich. 99.9%, 8YSZ) were placed in a mold, pressurized at 33.5 MPa, and then CIP was performed at 400 MPa. As a result, the nuclear fuel-consolidated body composite with the Gd2O3, 30 wt % Gd2O3-70 wt % 8YSZ, and 50 wt % Gd2O3-50 wt % 8YSZ pellets inserted into the core was prepared. The nuclear fuel-consolidated body composite was sintered in a microwave sintering apparatus at 1600° C. for 20 minutes to prepare the sintered nuclear fuel pellet. The SEM images of the prepared sintered body are shown in FIG. 9. As shown in FIG. 9, it was confirmed that the sintered nuclear fuel pellet with the Gd2O3, 30 wt % Gd2O3-70 wt % 8YSZ, and 50 wt % Gd2O3-50 wt % 8YSZ pellets inserted therein was formed. The method of the present invention can produce the sintered nuclear fuel pellet by applying the consolidated body of the burnable absorber thereto, so that the invention is advantageous in providing the sintered nuclear fuel pellet that can overcome the problem of the irregularity of the burnable absorber in the nuclear fuel; prevent the occurrence of cracks caused by the difference in thermal expansion and phase transformation during sintering; and increase the efficiency of the nuclear fuel. The present invention can regulate the insertion location, the size, and the shape of the consolidated body of the burnable absorber in order to optimize the lifetime and output of the nuclear fuel. Those skilled in the art will appreciate that the conceptions and specific embodiments disclosed in the foregoing description may be readily utilized as a basis for modifying or designing other embodiments for carrying out the same purposes of the present invention. Those skilled in the art will also appreciate that such equivalent embodiments do not depart from the spirit and scope of the invention as set forth in the appended Claims.
	claims	1. A collimation device for a dental imaging apparatus, the collimation device comprising:a collimator housing having a first end and a second end opposite to the first end, the collimator housing being configured to allow x-rays to move through the device in a direction from the first end to the second end; anda collimator plate constructed to at least partially block the passage of some of the x-rays moving through the device,wherein the collimator plate is movable relative to the collimator housing between a first position to a second position along the collimator housing, with the collimator plate being closer to the first end of the collimator housing when the collimator plate is in the first position than the collimator plate is to the first end of the collimator housing when the collimator plate is in the second position. 2. The collimation device of claim 1, wherein the collimator plate is movable to a third position between the first position and the second position. 3. The collimation device of claim 1, further comprising:a light emitting alignment device connected to the collimator housing and configured to generate a patterned light beam. 4. The collimation device of claim 1, further comprising:a collimator chassis,wherein the collimator housing is rotatably connected to the collimator chassis such that the collimator housing is rotatable between a first orientation and a second orientation. 5. The collimation device of claim 1, wherein an exterior surface of the collimator housing includes a first indicator indicating the first position and a second indicator indicating the second position. 6. The collimation device of claim 3, wherein the patterned light beam is a circular pattern. 7. The collimation device of claim 4, wherein the second orientation is ninety degrees from the first orientation. 8. The collimation device of claim 4, further comprising:a laser device connected to the collimator chassis and configured to emanate a laser beam. 9. The collimation device of claim 4, wherein the collimator housing includes a first pair of indexing elements and a second pair of indexing elements,wherein the collimator chassis includes a third pair of indexing elements,wherein the first pair of indexing elements are proximate to the third pair of indexing elements when the collimator housing and the collimator chassis are in the first orientation, andwherein the second pair of indexing elements are proximate to the third pair of indexing elements when the collimator housing and the collimator chassis are in the second orientation. 10. A collimation device for a dental imaging apparatus, the collimation device comprising:a collimator housing having a first end and a second end opposite to the first end, the collimator housing being configured to allow x-rays to move through the device in a direction from the first end to the second end;a collimator plate constructed to at least partially block the passage of some of the x-rays moving through the device; anda collimator mount assembly connected to the collimator plate,wherein the collimator housing includes a guide rail on which the collimator mount assembly is movable between a first position and a second position along the collimator housing, andwherein the collimator plate is closer to the first end of the collimator housing when the collimator plate is in the first position than the collimator plate is to the first end of the collimator housing when the collimator plate is in the second position. 11. An x-ray imaging system comprising:an x-ray source configured to generate an x-ray beam; anda collimation device connected to the x-ray source and arranged to receive the x-ray beam generated by the x-ray source, the collimation device including:a collimator housing having a first end and a second end opposite to the first end, the collimator housing being configured to allow x-rays of the x-ray beam to move through the device in a direction from the first end to the second end; anda collimator plate constructed to at least partially block the passage of some of the x-rays of the x-ray beam moving through the device, the collimator plate being movable relative to the collimator housing between a first position to a second position along the collimator housing, with the collimator plate being closer to the first end of the collimator housing when the collimator plate is in the first position than the collimator plate is to the first end of the collimator housing when the collimator plate is in the second position. 12. The x-ray imaging system of claim 11, wherein the collimator plate is movable to a third position between the first position and the second position. 13. The x-ray imaging system of claim 11, wherein the collimation device further comprises a light emitting alignment device connected to the collimator housing and configured to generate a patterned light beam in a direction of an aiming ring. 14. The x-ray imaging system of claim 11, wherein the collimation device further comprises:a collimator chassis,wherein the collimator housing is rotatably connected to the collimator chassis such that the collimator housing is rotatable between a first orientation and a second orientation. 15. The x-ray imaging system of claim 11, wherein an exterior surface of the collimator housing includes a first indicator indicating the first position and a second indicator indicating the second position. 16. The x-ray imaging system of claim 13, wherein the patterned light beam is a circular pattern. 17. The x-ray imaging system of claim 13, further comprising:an x-ray detector; andan alignment device that includes:a holder for holding the x-ray detector,an alignment ring, anda connecting arm connecting the holder to the alignment ring,wherein the alignment ring includes a primary alignment groove disposed on a side thereof, andwherein when the patterned light beam generated by the light emitting alignment device falls entirely within the primary alignment groove, the x-ray source is properly positioned and aligned relative to the x-ray detector. 18. The x-ray imaging system of claim 14, wherein the second orientation is ninety degrees from the first orientation. 19. The x-ray imaging system of claim 14, wherein the collimation device further comprises:a plurality of laser devices connected to the collimator chassis and configured to emanate respective laser beams. 20. The x-ray imaging system of claim 14,wherein the collimator housing includes a first pair of indexing elements and a second pair of indexing elements,wherein the collimator chassis includes a third pair of indexing elements,wherein the first pair of indexing elements are proximate to the third pair of indexing elements when the collimator housing and the collimator chassis are in the first orientation, andwherein the second pair of indexing elements are proximate to the third pair of indexing elements when the collimator housing and the collimator chassis are in the second orientation.
	abstract	A charged particle beam device for inspecting a specimen includes a charged particle beam source adapted to generate a primary charged particle beam; an objective lens device adapted to direct the primary charged particle beam onto the specimen; a retarding field device adapted to accelerate secondary charged particles starting from the specimen, a first detector device having a central opening, includes at least two azimuthal detector segments for detecting secondary particles, wherein the objective lens device is adapted such that particles with different starting angles from the specimen exhibit crossovers at substantially the same distance from the specimen between the objective lens and the detector device, and an aperture located between the objective lens and the crossovers, having an opening which is equal to or smaller than the central opening in the detector device.
043476228	abstract	Typical nuclear fuel material contains tramp ferromagnetic particles of random size and distribution. Also, selected amounts of paramagnetic or ferromagnetic material can be added at random or at known positions in the fuel material. The fuel material in its nonmagnetic container can be scanned by magnetic susceptibility change detecting apparatus to provide a unique signal waveform of the container of fuel material as a signature thereof. At subsequent times in its life, the container similarly can be scanned to provide subsequent signatures. Comparison of the signatures reveals any alteration or tampering with the fuel material.
	summary
052951650	summary	BACKGROUND This invention generally relates to hole plugs and more particularly relates to a self-locking plug for plugging a hole defined by a surrounding structure, which structure may be a nuclear power reactor pressure vessel core barrel flange. Although hole plugs are known in the prior art, it has been observed that prior art plugs have certain operational problems associated with them which make these plugs unsuitable for plugging a hole formed through a nuclear reactor pressure vessel core barrel flange. However, before these problems can be appreciated, some background is necessary as to the structure and operation of a typical nuclear power reactor and its associated core barrel flange. In this regard, a nuclear power reactor is a device for producing heat by controlled fission of fissionable material contained in fuel assemblies. A plurality of the fuel assemblies are grouped in a sealed reactor pressure vessel to define a nuclear reactor core therein. Pressurized liquid moderator coolant is caused to circulate through the pressure vessel and over the fuel assemblies for assisting in the fission process and for removing the heat produced by fission of the fissionable material contained in the fuel assemblies. Disposed inwardly of the pressure vessel is a core barrel for supporting the fuel assemblies thereon, the pressure vessel and the core barrel defining an annular space therebetween. The core barrel has a core barrel flange for connecting it to the inside of the pressure vessel so that the core barrel is supported by the pressure vessel. The core barrel flange has an access hole therethrough for providing access to a radiation detector disposed in the space between the core barrel and the pressure vessel. The radiation detector is typically a neutron detector or activation foil for measuring the integrated neutron flux seen by the pressure vessel. It is important to measure this integrated neutron flux in order to determine the margin remaining to the maximum allowable integrated neutron flux, also known as the "nil ductility temperature" limit, which has safety significance. Mounted atop the core barrel is an upper support structure for supporting reactor internal components. Although the access hole is open during reactor service operations to retrieve the radiation detector therethrough, the hole is closed during normal reactor operation so that the coolant does not flow through it. It is desirable not to have the coolant flow through the hole during normal reactor operation because coolant that flows through the hole bypasses the core and is therefore unavailable for cooling the core. It is desirable to adequately cool the core to avoid damage to the fuel assemblies. In order to prevent bypass flow through the hole defined by the core barrel flange, a prior art plug has been inserted into the hole to plug the hole. During normal reactor operation the upper support structure bears against this plug to maintain the plug in the hole. However, in order to retrieve the detector during reactor service operations, the upper support structure is removed to provide access to the core and detector. During detector retrieval operations that necessitate removal of the upper support structure, any coolant pump transients occurring at that time may be forceful enough to generate hydraulic forces that may cause the coolant to forcibly dislodge the prior art plug from the hole before it can be grasped and removed from the hole. That is, when the upper support structure is removed, the plug may become dislodged from the hole due to the transient hydraulic force of the coolant acting on the plug because the prior art plug is not secured or locked to the core barrel flange. Such inadvertent dislodgement of the plug from the hole prior to being grasped and controllably removed is undesirable because such a dislodged plug may become a "loose part" that may migrate in the coolant to damage fuel assemblies and other internal reactor components. Therefore, a problem in the art is to provide a plug for plugging a hole defined by a surrounding structure, such as a reactor pressure vessel core barrel flange, so that the plug is not inadvertently dislodged from the hole by upset hydraulic forces generated in the vessel. Hole plugs suitable for use in nuclear reactor pressure vessels are known. A plug for altering the flow of coolant in a nuclear reactor is disclosed in commonly owned U.S. Pat. No. 4,591,068 titled "Core Barrel Plug" issued May 27, 1986 in the name of Ralph W. Tolino, et al. This patent discloses a core barrel plug capable of being remotely installed in a port of a core barrel of a pressurized water nuclear reactor for converting the reactor from a by-pass downflow configuration to a by-pass upflow configuration. The plug comprises a body having an expandable cylindrical portion with a movable mandrel disposed in the body. Remote fluid pressurization causes the mandrel to be advanced thus expanding the body into contact with the port. The plug also comprises a locking mechanism to prevent inadvertent release of the plug. Although the Tolino, et al. patent discloses a core barrel plug comprising a locking mechanism to prevent inadvertent release of the plug, this patent does not appear to disclose a self-locking plug for plugging a hole defined by a surrounding structure, as described and claimed hereinbelow. Therefore, what is needed is a self-locking plug for plugging a hole defined by a surrounding structure, which structure may be a nuclear power reactor pressure vessel core barrel flange. SUMMARY Disclosed herein is a self-locking plug for plugging a hole defined by a surrounding structure, which structure may be a nuclear power reactor pressure vessel core barrel flange. The hole is typically open during reactor service operations to retrieve a neutron radiation detector or activation foil disposed in the pressure vessel, but is typically plugged at other times in order to prevent coolant from flowing through the hole and bypassing the reactor core. However, transient hydraulic forces generated in the pressure vessel will tend to force the plug from the hole. Therefore, according to the invention, a self-locking plug is provided that is capable of being locked to the core barrel flange as it plugs the hole, so that the plug is not susceptible to being inadvertently dislodged from the hole even when acted upon by transient hydraulic forces. The plug includes a plug body sized to be disposed in the hole and a locking member pivotally connected to the plug body for locking the plug body to the flange. A cam, which is capable of engaging the locking mechanism for outwardly pivoting the locking mechanism, is slidably connected to the plug body. A movable piston is connected to the cam for sliding the cam into engagement with the locking mechanism. The locking mechanism pivots to engage the flange as the cam slides to engage the locking member. Moreover, a ram, which is also connected to the plug body, is provided to ram the plug body into the hole for snugly plugging the hole. In this manner, the plug is locked to the flange and resists being dislodged from the hole, which may otherwise occur due to transient hydraulic forces generated in the pressure vessel. In its broad form, the invention is a self-locking plug for plugging a hole defined by a surrounding structure, comprising a plug body adapted to be disposed in the hole; locking means connected to said plug body for locking said plug body to the structure; camming means connected to said plug body for moving said locking means; actuator means connected to said camming means for actuating said camming means; and ramming means connected to said plug body for ramming said plug body into the hole. An object of the present invention is to provide a self-locking plug for plugging a hole defined by a sure rounding structure, which structure may be a nuclear power reactor pressure vessel core barrel flange. A feature of the present invention is the provision of a plug body adapted to be disposed in the hole and movable locking means connected to the plug body for locking the plug body to the structure. Another feature of the present invention is the provision of camming means connected to the plug body for moving the locking means. Yet another feature of the present invention is the provision of ramming means connected to the plug body for ramming the plug body into the hole. An advantage of the present invention is that the plug resists being inadvertently dislodged from the hole due to upset hydraulic forces generated in the pressure vessel because the plug is locked to the structure. Another advantage of the present invention is that the reactor coolant does not substantially flow through the hole to bypass the core when the plug plugs the hole. These and other objects, features and advantages of the present invention will become apparent to those skilled in the art upon a reading of the following detailed description when taken in conjunction with the drawings wherein there is shown and described illustrative embodiments of the invention.
046876278	summary	BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates generally to nuclear reactors and, more particularly, is concerned with an improved water displacer rod containing hollow sealed pellets which prevent the rod from filling with reactor coolant in the event of rod failure. 2. Description of the Prior Art In the conventional designs of pressurized water reactors, an excessive amount of reactivity is designed into the reactor core at start-up so that as the reactivity is depleted over the life of the core there will still be sufficient reactivity to sustain core operation over a long period of time. However, since an excessive amount of reactivity is designed into the reactor core at the beginning of the core life, steps must be taken at that time to properly control it. One technique to control reactivity is to produce an initial spectral shift which has the effect of increasing the epithermal (low reactivity) part of the neutron spectrum at the expense of the thermal (high reactivity) part. This results in production of fewer thermal neutrons and decreased fission. Then, as fission decreases during extended reactor operation, a reverse shift back to the thermal part of the neutron spectrum at the expense of the epithermal part is undertaken. Such control technique is primarily accomplished through the use of displacer rods. As the name implies, these rods are placed in the core to initially displace some of the moderator water therein and decrease the reactivity. Then, at some point during the core cycle as reactivity is consumed, the displacement associated with these rods is removed from the core so that the amount of moderation and therewith level of reactivity in the core are increased. One approach for removing moderator water displacement is described in a U.S. patent application entitled "An Improved Water Displacer Rod Spider Assembly for a Nuclear Reactor Fuel Assembly" by Trevor A. Francis, filed Mar. 30, 1984 and assigned U.S. Ser. No. 595,154. Several constructions of a water displacer rod are illustrated and described in U.S. Pat. No. 4,432,934 to Gjertsen et al. Referring to FIGS. 12 and 13 of this patent, the displacer rod includes an upper first tube composed, for instance, of stainless steel and a lower second tube composed, for example, of Zircaloy-4, with the tubes being interconnected by a mid extension in the for of a rod of solid Zircaloy-4. The upper first tube is sealed at its upper end by a top end plug capable of being threadably attached to a spider, while the lower second tube is sealed at its lower end by a bottom end plug which can be bullet shaped to aid in insertion of the rod into a guide thimble within one of the fuel assemblies of the reactor. In one construction of the displacer rod shown in FIG. 12, the second tube is thin walled and contains a stack of annular pellets extending from the bottom end plug to the mid rod extension which provides structural support for the tube to prevent its collapse under reactor operating pressure. In the alternative, the pellets can be of a solid instead of a hollow construction. As another alternative construction, the second tube can be thick walled and pressurized with an inert gas to prevent collapse of the rod under reactor pressure. In still another construction, the rod contains a spring instead of the stack of pellets. In any event, the construction of the displacer rod is such that it provides a low neutron absorbing rod that is capable of displacing reactor coolant-moderator when inserted into a fuel assembly. While all of the above-cited prior constructions of the water displacer rod perform reasonably well and generally achieve their objectives under the range of operating conditions for which they were designed, a need exists for improvement in the design of the displacer rod so as to increase its general utility in fuel assemblies in a variety of slightly deformed structural states and to facilitate even more efficient use of nuclear fuel in the reactor core through employment of the spectral shift principle. SUMMARY OF THE INVENTION The present invention provides an improved water displacer rod designed to satisfy the aforementioned needs. Design activity leading up to the improved rod of the present invention has brought to light certain important considerations which affect the design of water displacer rods and are met by the rod construction of the present invention. First, the rod must be flexible to permit axial movement of the rod in a fuel assembly whose structure has become somewhat deformed axially, such as warped or bowed. Second, the amount of coolant flow needed to remove heat from the rod must be kept to a minimum. Third, to more precisely calculate nuclear peaking factors in the core, it must be assumed that there is no coolant in any of the rods. Each of the three above-mentioned considerations underlie, and are satisfactorily addressed by, the improved water displaced rod construction of the present invention. As will be explained hereafter, the problems of providing adequate rod flexibility and collapse resistance, of minimizing the need for increased coolant flow to avoid rod overheating, and of minimizing nuclear peaking in adjacent fuel rods are all treated in a satisfactory manner. First, flexibility of the rod can be increased through reduction in the moment of inertia of the rod. The most practical way to minimize the rod's inertia is to use a thin-walled cladding or tube in the rod. The use of a thin-walled tube, in turn, requires the provision of some means within it to prevent collapse of the tube due to external pressure. A stack of pellets therein bolsters the strength of the tube without adversely affecting its flexibility. Compared thereto, the use of another inner thin tube would augment resistance of the outer thin tube against collapse but would decrease its flexibility. Second, to minimize the amount of coolant flow required to remove heat from the rod and to prevent surface boiling due to gamma heating, the mass of the pellets selected to satisfy the first consideration--flexibility and collapse resistance--must be minimized. The use of hollow or annular pellets is a practical way of minimizing the mass of the pellets without deleteriously affecting the flexibility and collapse resistance of the tube. Third, without some additional steps being taken, the use of a thin walled tube increases the probability of clad failure and entry of coolant into the displacer rod and, to further compound the problem, the use of hollow annular pellets ensures that the entering coolant will flow throughout the empty spaces in the rod. Thus, in absence of some additional steps being taken, peaking factors will substantially increase. However, in accordance with the present invention, by individually sealing the center void of each annular pellet, substantially all of the empty space within the tube will be made impenetrable to coolant even if some of the thin walled tubes of the displacer rods should fail. Furthermore, the solid sealed ends of the individual pellets provide radial support at axially spaced locations along the rod which enhances its collapse resistance without adversely affecting its flexibility nor significantly increasing its overall mass. Accordingly, the present invention sets forth in a fuel assembly for a nuclear reactor including an organized array of nuclear fuel rods and moderator-coolant liquid flowing along the fuel rods, at least one improved water displacer rod disposed among the fuel rods of the assembly. The improved displacer rod includes: (a) an elongated hollow hermetically-sealed thin-walled tubular member; and (b) a plurality of pellets disposed in a stacked relationship within the tubular member. Each of at least a substantial number of the pellets have (i) a body with opposite ends and a hollow annular cross-sectional shape defining a central void through the pellet, and (ii) a web extending across at least one of the opposite ends of the body so as to close the void at the one body end such that the respective voids of the bodies of the pellets disposed in the stacked relationship are enclosed individually one from the next. More particularly, each of the pellets has a pair of webs extending across the respective ends of the pellet body so as to close the opposite body ends such that the respective voids of the bodies are sealed individually one from the next. In one form of the pellet, one of the webs is integrally connected with the pellet body at one of its opposite ends and the other of the webs is a cap attached with the pellet body at its other end. In an alternative form, each of the webs is integrally connected with the pellet body at a respective one of its opposite ends. In another alternative form, each of the webs is a cap attached to a respective one of the opposite ends of the pellet body. These and other advantages and attainments of the present invention will become apparent to those skilled in the art upon a reading of the following detailed description when taken in conjunction with the drawings wherein there is shown and described an illustrative embodiment of the invention.
	claims	1. A measuring device comprising:a housing comprising two surfaces facing away from each other, the housing being provided with a flow path structure, a first cavity, and a second cavity, whereinthe flow path structure has two openings on the two surfaces respectively and comprises a first side surface and a second side surface facing each other,the first cavity comprises one end opening on the first side surface,the second cavity faces the first cavity across a center axis of the flow path structure and comprises one end opening on the second side surface;a first cell disposed in the first cavity, made from a transparent material, and having a substantially cylindrical shape, both end surfaces of the first cell being flat surfaces;a second cell disposed in the second cavity, made from a transparent material, and having a substantially cylindrical shape, both end surfaces of the second cell being flat surfaces;a light emitting section configured to irradiate a hydraulic oil flowing in the flow path structure with light via the first cell in a direction substantially orthogonal to the center axis; anda light receiving section disposed facing the light irradiating section across the first cell, the flow path structure, and the second cell;the first cavity and the second cavity having centers substantially coinciding with an optical axis being a center of light emitted from the light irradiating section,wherein:the flow path structure comprises both ends shaped into a round cavity;the flow path structure has such a tapered shape that a shape in a plane substantially orthogonal to the center axis changes from the round cavity to a long cavity having two sides substantially orthogonal to the optical axis; andthe tapered shape is formed such that a taper angle is approximately 60 degrees with respect to the center axis. 2. The measuring device according to claim 1, wherein:the first side surface and the second side surface are flat surfaces;an end surface of the first cell disposed in the first cavity and the first side surface are substantially in the same plane; andan end surface of the second cell disposed in the second cavity and the second side surface are substantially in the same plane. 3. The measuring device according to claim 2, wherein a distance between the first side surface and the second side surface is smaller than a length of the first side surface and the second side surface in the direction substantially orthogonal to the center axis. 4. The measuring device according to claim 2,wherein:the first cell comprises a first main portion having a substantially cylindrical shape and a first flange portion formed at an end of the first main portion and having a substantially circular plate shape with a diameter greater than a diameter of the first main portion;a first sealing member is disposed between the first cavity and the first main portion;a first pressing member is disposed within the housing;the first pressing member presses the first cell; andend surfaces, on the first main portion side, of the first flange portion is pressed against the housing, andwherein:the second cell comprises a second main portion having a substantially cylindrical shape and a second flange portion formed at an end of the second main portion and having a substantially circular plate shape with a diameter greater than a diameter of the second main portion;a second sealing member is disposed between the second cavity and the second main portion;a second pressing member is disposed within the housing;the second pressing member presses the second cell; andend surfaces, on the second main portion side, of the second flange portion is pressed against the housing. 5. The measuring device according to claim 1, wherein a distance between the first side surface and the second side surface is smaller than a length of the first side surface and the second side surface in the direction substantially orthogonal to the center axis. 6. The measuring device according to claim 1,wherein:the first cell comprises a first main portion having a substantially cylindrical shape and a first flange portion formed at an end of the first main portion and having a substantially circular plate shape with a diameter greater than a diameter of the first main portion;a first sealing member is disposed between the first cavity and the first main portion;a first pressing member is disposed within the housing;the first pressing member presses the first cell; andend surfaces, on the first main portion side, of the first flange portion is pressed against the housing, andwherein:the second cell comprises a second main portion having a substantially cylindrical shape and a second flange portion formed at an end of the second main portion and having a substantially circular plate shape with a diameter greater than a diameter of the second main portion;a second sealing member is disposed between the second cavity and the second main portion;a second pressing member is disposed within the housing;the second pressing member presses the second cell; andend surfaces, on the second main portion side, of the second flange portion is pressed against the housing. 7. A measuring device comprising:a housing comprising two surfaces facing away from each other, the housing being provided with a flow path structure, a first cavity, and a second cavity,the flow path structure has two openings on the two surfaces respectively and comprises a first side surface and a second side surface facing each other,the first cavity comprises one end opening on the first side surface,the second cavity faces the first cavity across a center axis of the flow path structure and comprises one end opening on the second side surface;a first cell disposed in the first cavity, made from a transparent material, and having a substantially cylindrical shape, both end surfaces of the first cell being flat surfaces;a second cell disposed in the second cavity, made from a transparent material, and having a substantially cylindrical shape, both end surfaces of the second cell being flat surfaces;a light emitting section configured to irradiate a hydraulic oil flowing in the flow path structure with light via the first cell in a direction substantially orthogonal to the center axis; anda light receiving section disposed facing the light irradiating section across the first cell, the flow path structure, and the second cell;the first cavity and the second cavity having centers substantially coinciding with an optical axis being a center of light emitted from the light irradiating section,wherein a distance between the first side surface and the second side surface is smaller than a length of the first side surface and the second side surface in the direction substantially orthogonal to the center axis. 8. A measuring device comprising:a housing comprising two surfaces facing away from each other, the housing being provided with a flow path structure, a first cavity, and a second cavity,the flow path structure has two openings on the two surfaces respectively and comprises a first side surface and a second side surface facing each other,the first cavity comprises one end opening on the first side surface,the second cavity faces the first cavity across a center axis of the flow path structure and comprises one end opening on the second side surface;a first cell disposed in the first cavity, made from a transparent material, and having a substantially cylindrical shape, both end surfaces of the first cell being flat surfaces;a second cell disposed in the second cavity, made from a transparent material, and having a substantially cylindrical shape, both end surfaces of the second cell being flat surfaces;a light emitting section configured to irradiate a hydraulic oil flowing in the flow path structure with light via the first cell in a direction substantially orthogonal to the center axis; anda light receiving section disposed facing the light irradiating section across the first cell, the flow path structure, and the second cell;the first cavity and the second cavity having centers substantially coinciding with an optical axis being a center of light emitted from the light irradiating section,wherein:the first cell comprises a first main portion having a substantially cylindrical shape and a first flange portion formed at an end of the first main portion and having a substantially circular plate shape with a diameter greater than a diameter of the first main portion;a first sealing member is disposed between the first cavity and the first main portion;a first pressing member is disposed within the housing;the first pressing member presses the first cell; andend surfaces, on the first main portion side, of the first flange portion is pressed against the housing, andwherein:the second cell comprises a second main portion having a substantially cylindrical shape and a second flange portion formed at an end of the second main portion and having a substantially circular plate shape with a diameter greater than a diameter of the second main portion;a second sealing member is disposed between the second cavity and the second main portion;a second pressing member is disposed within the housing;the second pressing member presses the second cell; andend surfaces, on the second main portion side, of the second flange portion is pressed against the housing.
	description	FIGS. 1 and 2 show a transport container for new fuel assemblies of a pressurized water nuclear reactor denoted overall by the reference number 1. The transport container 1, which is designed to transport two fuel assemblies in a horizontal position, has an external envelope 2 formed by a lower shell 2a and an upper shell 2b, both semi-cylindrical in shape and connected one on top of the other along a joining plane of the envelope 2 passing through the longitudinal axis of the cylindrical-shaped envelope. Each of the shells 2a and 2b is made from steel sheet and has semicircular reinforcing ribs 3a, 3b respectively, distributed over the length of the half-shell. Sections 4 and 4xe2x80x2 are also fixed to the lower part of the lower half-shell 2a, said sections forming support feet for the container. Furthermore, adjustable support elements 5 and 5xe2x80x2, which have screw jacks and which are secured to a longitudinal end part of the container, enable the inclination of the container resting on a support surface to be adjusted about the longitudinal axis of the container and about a transverse axis of the container respectively. By using the adjustable feet 5 and 5xe2x80x2 of the container, it is possible to place the container, on its transport support, in a perfectly horizontal position, i.e. in a position in which the longitudinal axis of the container is perfectly horizontal. The two half-shells 2a and 2b are brought together one on top of the other via rectangular peripheral flanges forming an upper planar support part of the lower half-shell 2a and a lower planar support part of the upper half-shell 2b of the container. In the closed position of the container, shown in FIGS. 1 and 2, the flanges of the two half-shells 2a and 2b are brought together and fixed one on top of the other by screws and nuts, forming an assembly flange 6. FIGS. 3A and 3B show a part of the container in the open state, i.e. with the upper half-shell of the container envelope separated from the lower half-shell and removed. FIGS. 3A and 3B show the internal structure of the container, denoted overall by the reference number 7, which has in particular a cradle 8 resting on supports 9 formed by shock-absorber pads, in the lower half-shell 2a of the external envelope 2 of the container. A second part of the internal structure of the container is formed by a unit 10 for receiving and supporting two fuel assemblies in the horizontal position placed side by side. The unit 10, which rests on the cradle 8, defines two completely closed housings for two fuel assemblies, as will be explained hereinafter. The cradle 8 has two side rails 8a, 8b formed by angle brackets fixed on the support pads 9 and which are held in parallel positions, with a separation corresponding to the width of the unit 10 for receiving the container, by crossmembers. At one of its ends, the cradle has a pivoting stiffening and mounting unit comprising two plates 11a and 11b which are parallel to one another and two crossmembers formed by hollow sections fixed to the side rails of the cradle and to the plates 11a and 11b. The mounting of the cradle on the lower shelf of the container such that it can pivot about a horizontal axis of transverse direction, is ensured via the pivoting stiffening and mounting unit comprising the plates 11a and 11b. Furthermore, as will be explained hereinafter, a retaining plate for fuel assemblies is also mounted between the plates 11a and 11b. As can be seen in FIG. 3B, a shock absorber 43 is inserted between the longitudinal end of the internal structure 7 and the internal circular end wall of the external envelope 2, in such a way as to limit the effect of a shock to the fuel assemblies, for example the effect of dropping a container. The shock-absorber 43, in the shape of a disc whose cross section is identical to the internal cross section of the container envelope, is made up of a balsa disc surrounded by an envelope made from stainless steel sheet. Of course, an identical shock absorber is positioned at the second longitudinal end of the container, between the second longitudinal end of the internal structure and the second end of the external envelope. As can be seen in FIG. 4, the fuel assembly support and reception unit 10 has a frame 12 having a Tshaped cross section and two doors 14a and 14b mounted is such that they pivot on the sides of the frame 12, as will be explained hereinafter. In the closed position of the doors, as shown in FIG. 4, the door 14a together with the right part of the frame 12 defines a housing 13a for one fuel assembly and the door 14b together with the left part of the frame 12 defines a second housing 13b. The housings have a square cross section which has the dimensions of the cross section of a spacer-grid of a pressurized water nuclear reactor fuel assembly for which the container 1 ensures transport. To load the container, the cradle 8 is made to tilt about the transverse axis located at one of the ends of the cradle into a position which is substantially vertical. In its tilted position, the fuel assembly reception and support unit 10, is in a vertical position. The doors 14a and 14b are tilted towards the outside, in such a way as to give access to the housings 13a and 13b. A fuel assembly may be placed in each of the housings 13a and 13b, using a fuel assembly lifting tool, for example the hoist of an overhead crane. The fuel assemblies come to rest, via their bottom nozzles, on the fuel assembly support plate fixed between the two plates 11a and 11b of the cradle 8. The doors of the fuel-assembly reception and support unit 10 are closed and the unit 10 is tilted into the horizontal position, coming to rest on the cradle 8. After having placed the upper half-shell back on the lower half-shell of the envelope 2 and fixed the two half-shells by screws and nuts, the container can be handled and transported, for example by lifting the container using lifting lugs 15 and 15xe2x80x2 fixed on the upper half-shell of the external envelope, as shown in FIG. 1. FIG. 5 shows an exploded view in perspective of the cradle 8 and the various elements forming the fuel assembly reception and support unit 10. The frame 12, which has a T-shaped transverse section, has a parallelepipedal base 12a and a wall 12b perpendicular to the base 12a, separating the housings 13a and 13b for two fuel assemblies 16a and 16b, the spacer-grids 17a and 17b, the bottom nozzles 18a and 18b and the top nozzles 18xe2x80x2a and 18xe2x80x2b of which are shown. The housings 13a and 13b of the fuel assemblies 16a and 16b are defined at one of the ends of the frame 12, by a support plate 20 intended to be fixed such that it pivots, via stub shafts, between the plates 11a and 11b of the cradle 8 and a second end plate 21 mounted such that it pivots at the second end of the frame 12, about a transverse pivot axis. The fuel assemblies rest, via their top nozzles 18xe2x80x2a and 18xe2x80x2b, on the plate 20. The transverse holding plate 21 has adjustable supporting end-stops on the bottom nozzles 18a and 18b of the fuel assemblies. The plate 21 could also have adjustable means for holding the fuel assemblies in the longitudinal direction. When the end plates 20 and 21 are pulled down into their closed position, the fuel assemblies are held in the longitudinal direction by being clamped between the support devices 22 and the plate 20. The pivoting lateral doors 14a and 14b of the unit 10 holding and supporting the fuel assemblies 16a and 16b have an inverted L-shaped cross section and have, along their lower edge, at the end of one of the branches of the L, articulating parts 23 in the form of hinges spaced out over the length of the doors 14a and 14b. The doors shown in FIG. 5 have six hinges. 23 spaced apart over the length of a first lower edge of the doors 14a or 14b. Along its opposite second edge, at the end of the second branch of the L, each of the doors 14a and 14b has fixing lugs 24 having a part pierced by an opening and projecting slightly towards the outside with respect to the edge of the door. The hinge shaped articulating parts 23 have all openings aligned in a direction parallel to the edge of the door and each one engages on an articulation axis 25, fixed so that it projects from a lateral edge of the base 12a of the frame 12 of the fuel-assembly support. Similarly, the openings in the parts projecting from the lugs 24 located along the second edge of the doors are aligned in a direction parallel to the edge of the door. The median wall 12b of the frame 12 has on its upper edge guide parts 26 and 26xe2x80x2 having openings which are all aligned in a direction parallel to the upper edge of the median wall 12b of the frame 12. When the doors, which are mounted articulated on the articulation axes 25 via hinges 23, are pulled down to the closed position, the second edges of the doors 14a and 14b along which the lugs 24 are located, are pulled down onto the upper edge of the median wall 12b of the frame 12, each of the lugs 24 coming to a position inserted between two successive guide posts 26 and 26xe2x80x2 fixed on the upper edge of the median wall 12b of the frame 12. The doors 14a and 14b, when in the closed position, can be locked by introducing a rod into the aligned openings of the parts 26 and 26xe2x80x2 and of the lugs 24. Furthermore, the doors 14a and 14b have pegs 27a, 27xe2x80x2a and 27b, 27xe2x80x2b respectively at their longitudinal ends projecting towards the outside in the longitudinal direction. The end plates 20 and 21 of the frame 12 each have, along their upper and lateral edges, slots 28 and 28xe2x80x2, each one intended to receive one of the pegs 27a or 27b or one of the pegs 27xe2x80x2a and 27xe2x80x2b respectively, in the closed position of the doors, after the end walls 20 and 21 have been pulled down. Furthermore the walls 20 and 21 have openings passing through them, facing each of the nozzles of the fuel assemblies, in their transport position inside the housings 13a and 13b. Each of the fuel assembly housings 13a or 13b, which is defined on two lateral faces by two mutually perpendicular surfaces of the frame 12, on its opposite lateral faces by two internal perpendicular surfaces of a door 14a or 14b and at its ends by the plates 20 and 21, is completely closed and ensures effective containment of a fuel assembly. Should the container be subjected to a shock, leading to a partial destruction of the fuel assembly, pieces of fuel assemblies, for example pieces of fuel pellets or rods, cannot escape from the fuel assembly housing and be spread in the container. The doors 14a and 14b and the end walls 20 and 21 which are mounted such that they pivot, form a box having two housings for fuel assemblies, which can be opened to give access to the fuel assembly housings. Furthermore, as will be explained hereinafter, the base 12a and the median wall 12b of the frame and the walls of the doors 14a and 14b are constructed in the form of a double wall inside the thickness of which a neutron-absorbing resin i.e. a synthetic resin to which is added an element which strongly absorbs neutrons, is placed. FIG. 6 shows an exploded view in perspective of the elements forming the frame 12 of the fuel assembly reception and support unit. The frame 12 has a baseplate 30 reinforced by welded ribs 29 and by transverse sections 31 at the end of which are fixed articulation axes 25 for the doors 14a and 14b and lugs 32 for fixing the frame 12 to the lateral sides of the cradle 8, via screws and nuts (FIGS. 7 and 8). On either side of each of the sections 31, on top of the plate 29, in its median part, columns 33 are fixed perpendicular to the plate 29. To the upper part of the columns 33 are fixed elements 26xe2x80x2 for guiding the means of locking the doors of the fuel assembly reception and support unit. The second element forming the frame 12 is a profiled element in folded metal sheet 34 comprising two elements of metal sheet folded into an L-shape extended towards the bottom by two sills and connected at their upper part by elements which are folded and/or attached forming guide parts 26 for guiding the upper edge of the median wall 12b of the frame 12. On the folded-down lateral edges of the profiled sheet-metal element 34 passages are provided for the articulation axes of the doors and the pads for fixing the frame on the cradle which are fixed to the end of the reinforcing sections 31. Two T-shaped spacer parts 35a and 35b are fixed to the end of the plate 29. The frame 12 is produced by assembling the folded sheet-metal element 34 and the baseplate 29 having reinforcing elements and the columns 33. The end spacers 35a and 35b of the baseplate 30 are inserted into the internal profile of the folded sheet-metal element 34. Similarly the six columns 33 are inserted into the vertical part of the internal profile of the folded sheet-metal element 34, between the two vertical branches of the two L-shaped lateral sheet-metal elements. The guide parts 26xe2x80x2 fixed to the end of the columns are inserted between two successive guide parts 26 connecting the two L-shaped folded sheet-metal elements, in the form of the profiled element 34 with a T-shaped transverse cross section. In the assembled position of the frame 12, the horizontal parts of the sheet-metal elements folded into an L-shape come to rest on the spacers 35 and on the sections 31, in such a way that an empty space is kept between the horizontal parts of the sheet-metal element 34 and the baseplate 29. As can be seen in FIG. 8, this free space 36 is filled with a neutron-absorbing resin. The resin is a dense resin whose density is between 1.5 and 2. Similarly, an empty space 37 between the vertical parts of the sheet-metal element 24 is filled with a high density neutron-absorbing resin. The resin and the spacer elements ensure the mechanical integrity of the frame 12. By assembling the plate 30, its reinforcing elements and the columns 33 with the folded sheet-metal element 34, a double walled, stiff frame 12 is obtained. By filling the empty spaces 36 and 37 of the double wall with a neutron-absorbing resin, a frame whose baseplate 12a and the separating median wall 12b are capable of absorbing neutron flux produced by fuel assemblies placed in the housings 13a and 13b of the frame 12 is obtained. FIG. 9 shows the right hand door 14a of the fuel assembly reception and support unit. The door 14a (and likewise the second door 14b) is formed by sheet-metal elements folded into an L-shape which are connected to one another at the ends of the branches of the L by extensions of one of the branches, the articulating parts 23 and the locking lugs 24. Furthermore, between the two metal sheets forming the L-shaped door, spacers 38 are placed at a certain distance from one another over the length of the door 14a. Each of the spacers 38 has, as can be seen in FIG. 10, two L-shaped plates spaced out from one another in the longitudinal direction of the door and fixed at their ends to an articulating part 25 and to a locking lug 24, respectively. A fuel assembly clamping device, placed in the housing defined by the door, is fixed to each of the branches of the L at each spacer 38 between the two L-shaped plates forming the spacer, ensuring the fuel assembly is held in a transverse direction. As can be seen in FIG. 10, each of the clamping devices 39 has a flat pad 40 which can be manoeuvred from the outside of the door by a screw 41, in order to move it in a direction perpendicular to the branch of the L of the door in which the locking device 39 is mounted. At each of the spacers 38, the door 14a has two clamping devices 39 intended to come into contact with two external faces of a spacer-grid of a fuel assembly positioned in the housing defined by the door 14a. In this way, the fuel assembly is clamped into its housing, on two mutually perpendicular sides. FIG. 11 shows a longitudinal end of the door 14a which is closed by an L-shaped plate 41 to which are fixed, projecting towards the outside, pegs 27 for fixing the door 14a to the end wall 21. As can be seen on the cutaway part of FIG. 11, a blocking rod 42 is mounted so that it slides in aligned openings in the upper horizontal wall of the door 14a and between the pegs 27a. Furthermore, the rod 42 is manoeuvrable from the outside of the door 14a. When the door 14a is in the closed position and the end plate 20 (or 21) is pulled down to the closed position of the longitudinal ends of the housings of the fuel assembly reception and support unit, the rod 42 can be introduced into aligned openings passing through the external parts of the plate 20 (or 21) between the slots 28, in the transverse direction and the openings between the pegs 27a placed in alignment with the openings of the plates 20 (or 21). In this way the end closure plates 20 and 21 are locked on the end parts of the door 14a. Of course, each of the ends of the door 14a having pegs 27a and 27xe2x80x2a can be locked in an identical fashion. The same locking rod 42 can lock the second door 14b by being introduced into the openings of the plate 20 (or 21) and the pegs 27b (or 27xe2x80x2b). The empty space between the two elements of the L-shaped wall of the doors 14a and 14b is filled with a neutron-absorbing resin, in order to absorb any neutron flux originating from a fuel assembly and directed towards the outside of the fuel assembly reception and support unit. The resin, which has high density (density from 1.5 to 2), and the spacers ensure the mechanical integrity of the doors. The internal structure of the container according to the invention defines two housings for two fuel assemblies which are completely closed and inside which the fuel assemblies are held laterally and in the axial or longitudinal direction. As the housings are completely closed, if any shock should cause partial destruction of a fuel assembly, parts of the fuel assembly are incapable of escaping from the internal structure which ensures the containment of the fuel assembly. The pieces of the fuel assembly are therefore incapable of spreading inside the external envelope of the container. Furthermore, the fuel assemblies are separated from each other inside the internal structure of the container, by a neutron-absorbing wall. The fuel assembly housings defined by the internal structure also have a neutron-absorbing wall closing the housings on the outside, i.e. towards the internal surface of the external envelope of the container. Improved mechanical protection of the fuel assemblies during their transport inside the container is therefore obtained at the same time as a reduction in the risks of achieving criticality during transport of more than one fuel assembly. The invention is not limited to the embodiment which has been described. In this way, the internal structure of the container may have a different shape to that which has been described and may have elements other than a T shaped frame and tilting doors. The shape of the housings in the internal structure of the container depends on the shape of the fuel assemblies being transported. In all cases, the internal structure has walls assembled to each other defining at least one completely closed fuel assembly reception and holding housing. The invention is applicable to the transport of any nuclear fuel assembly having a right prismatic shape. The container according to the invention can be used not only for the transport of new fuel assemblies but also for the transport of used fuel assemblies having low activity.
	summary
	abstract	A detection apparatus is usable to detect the neutron absorption capability of a control element of a nuclear installation and includes a neutron radiograph apparatus and a robot apparatus. The neutron radiograph apparatus includes a neutron emission source of variable strength, a detector array, a mask apparatus and a positioning robot all under the control of a central processor and data acquisition unit. The neutron emission source is advantageously switchable between an ON state and OFF state with variable source strength in the ON state, which avoids any need for shielding beyond placing the neutron emission source in an inspection pool at the nuclear plant site including but not limited to the spent fuel or shipping cask laydown pools. The neutron emission source is situated at one side of a wing of the control element and generates a neutron stream, the detector array is situated on an opposite side of a wing, and the neutron emission source and detector array are robotically advanced along the wing. The detector array is monitored in real time, and various masks of the mask apparatus can be positioned between the neutron emission source and the detector array to more specifically identify the position on the blade where the neutrons are passing through.
050230468	summary	BACKGROUND OF THE INVENTION 1. Field of the Invention The invention is generally related to the inspection of nuclear fuel rods and in particular to a drive unit for driving and rotating a fuel rod in a test tank for ultrasonic inspection of the end closure weld. 2. General Background During fabrication of nuclear fuel rods, ultrasonic (UT) inspection is performed on all fuel rod closure welds. The end of the fuel rod to be inspected is inserted horizontally through a seal into a test tank filled with water. The inspection is performed by rotating the fuel rod about its longitudinal axis and scanning the weld with the UT transducer during rotation. Equipment commonly used in such inspections utilizes a push/pull unit located at the end of the fuel rod away from the test tank and a rotation unit located adjacent to the test tank to position and manipulate the fuel rod for the UT inspection. The carriage mounted push/pull unit travels on linear ball bushings to insert the fuel rod in the test tank and incorporates an air cylinder operated clamp to withdraw the fuel rod at the completion of the inspection cycle. Rotation of the fuel rod is provided by a 3-jaw lathe chuck which is turned by an electric motor through a timing belt drive. An electrically operated actuator opens and closes the chuck jaws. The push/pull unit pushes the leading end of the fuel rod through the electric actuator and chuck into the test tank against a retractable stop. The actuator then closes the chuck to grip the fuel rod and the motor rotates the chuck and fuel rod for the UT inspection cycle. Upon completion of the test cycle, the motor is shut off and the actuator opens the chuck jaws to release the fuel rod. The fuel rod is then drawn back from the test tank and through the chuck and actuator by the push/pull unit. During the return stroke, and after the fuel rod has cleared the chuck and actuator, the clamp on the push/pull unit opens to release the fuel rod which is left on the handling rack for transfer to the next station. One problem presented by such equipment is the requirement of frequent maintenance due to water on the fuel rod being left on the internals of the chuck and actuator during withdrawal of the fuel rod from the test tank. Particularly after periods of disuse, rusting can cause these parts to seize, requiring disassembly and rebuilding. Another problem is that the clamping action of the 3-jaw chuck on the fuel rod causes axial displacement of the fuel rod in excess of the 0.001 inch tolerance necessary for the new seal welds. The clamping action of the 3-jaw chuck also increases the potential for damage to the fuel rod itself in the form of the fuel rod cladding being scratched and removed from the exterior of the rod or the relatively thin wall of the fuel rod being dented. The metallic cladding is important in retaining fission gas products within the fuel rod and protecting the fuel rod from the corrosive environment while dents in the fuel rod wall may limit movement of fuel pellets in the rod as they tend to change shape due to irradiation during their operational life span. Inspection devices which applicant is aware of include the following. U.S. Pat. No. 4,368,644 entitled "Tool For Inspecting Defects In Irregular Weld Bodies" discloses a tool which is actuated to move an ultrasonic inspection transducer over the surface of an irregular weld body such as the intersection of the nozzle of a reactor vessel to the vessel. U.S. Pat. No. 4,660,419 entitled "Reference Standard For Calibration Of Ultrasonic Arrays" discloses the use of an array of ultrasonic transducers disposed along a circular arc for examining tubular objects moving therethrough and a calibrating cylinder having calibrating flaws therein. U.S. Pat. No. 4,681,730 entitled "Process And Device For Detecting Leaking Nuclear Fuel Elements In A Nuclear Assembly" discloses the use of an ultrasonic transducer carried by rod means for insertion in the assembly between an end piece and the caps of the element. U.S. Pat. No. 4,096,757 entitled "Method And Apparatus For Examining Weld Defects In Vertical Pipes By Supersonic Waves" discloses a supersonic probe on a rotary shaft revolved along the inner surface of a pipe to be examined. U.S. Pat. No. 4,686,078 entitled "Method And Apparatus For Displacing A Reactor Weld Scanner Assembly By Variable Buoyancy" discloses the use of buoyancy chambers to move a scanner and telescopic boom into position relative to a reactor vessel. It is seen that the known art does not address the above problems of axial displacement tolerance and water deposition on the drive unit. SUMMARY OF THE INVENTION The present invention addresses the aforementioned problems in a straightforward manner. What is provided is a push/pull/rotation device that is part of the automated system of transfer racks and equipment designed to feed fuel rods to and from the fuel rod end cap seal weld ultrasonic inspection system. A gripping and rotating device has a motor mounted thereon and a spindle directly coupled to the motor. A rubber gripping mechanism is accepted at one end of the spindle and has a flanged port through which air pressure is applied for gripping the end of a fuel rod to be tested. Actuation of the motor causes rotation of the spindle and fuel rod held by the gripping mechanism. The housing of the gripping and rotating device is attached to a mounting plate connected to an assembly for pushing the fuel rod into the test tanks for inspection and pulling the fuel rod out of the test tanks when the inspection is completed.
047724307	claims	1. A process for compacting and solidifying solid waste materials, comprising shredding solid waste materials having a moisture content of up to 30% by weight to obtain a shredded mass, and compacting and solidifying said shredded mass at a temperature of 120.degree. to 260.degree. C. by extrusion together with at least one thermoplastic resin as a solidifying agent for said waste materials, said extrusion being carried out under compression in an extrusion molder with a screw to obtain a rod-like mass simultaneously with generating friction heat by friction of the thus obtained rod-like mass with the inner wall of said extrusion molder to melt the thermoplastic resin contained in the peripheral portion of said rod-like mass in sliding relation to said inner wall and said screw, said thermoplastic resin being present in an amount by weight of at least 10% of said solid waste materials whereby a solid mixture of said waste materials and said thermoplastic resin is obtained in the form of a rod-like mass. 2. A process for compacting and solidifying solid waste materials which contain used ion exchange resins which consists of shredding said solid waste materials of moisture content up to 30% by weight to obtain a shredded mass and compacting and solidifying said shredded mass at a temperature of 100.degree.-190.degree. C. by extrusion together with at least one thermoplastic resin as a solidifying agent for said waste materials, said extrusion being carried out under compression in an extrusion molder with a screw to obtain a rod-like mass simultaneously with generating friction heat by friction of the thus obtained rod-like mass with the inner wall of said extrusion molder to melt the thermoplastic resin contained in the peripheral portion of said rod-like mass in sliding relation to said inner wall and said screw, said thermoplastic resin being present in an amount by weight of at least 10% of said solid waste materials whereby a solid mixture of said waste materials and said thermoplastic resin is obtained in the form of a rod-like mass. 3. The process according to claim 1 wherein said extrusion molder is an extruder, said extruder has a die, and said waste material is subdivided to a particle size corresponding to the size of the diameter of said die.
041587790	claims	1. Shielding device for the protection of human embryo cells from the effects of radiation during examination by X-rays comprising a radiant absorbing protective screen including a central radiant absorbing protective plate of predetermined configuration, a plurality of additional radiant absorbing protective displaceable sheets and adjusting means to displace the sheets relative to each other to vary the covering surface area of the screen. 2. Device according to claim 1 including a supporting frame to which the protective screen is rotatably secured for rotation of the screen within its own plane. 3. Device according to claim 2 wherein the plate and sheets are composed of lead. 4. Device according to claim 3 wherein the plurality of sheets comprises pairs of first lead sheets movable at right angles for enlarging or reducing the covering area of the protective screen and further including second lead sheets mounted in pairs and being rotatable operatively on one pair of the first sheets, pressure means for correspondingly rotatably biasing the second sheets, and cam members disposed on another pair of the first sheets on which the second sheets are freely and slidably supported. 5. Device according to claim 4 including holders associated with the adjusting means to which the pairs of sheets are secured, the pairs of sheets being synchronously and oppositely displaceable. 6. Device according to claim 5 wherein the holders comprise slides having a rack-like extension provided with teeth, a pinion for meshing with such teeth for displacement of all the sheets and a knob for rotating the pinion to displace the sheets. 7. Device according to claim 6 wherein the associated pairings of sheets are arranged in various movement planes, and all of the sheets are displaceable relative to each other in the direction of the corresponding planes of the sheets. 8. Device according to claim 1 including a closed casing completely enclosing the protective screen, the casing comprising an upper part and a lower part and having on one such part a conical surface inclined in the direction of radiation. 9. Device according to claim 8 wherein the casing comprises two identical parts of plate-like shape, the parts being joinable by an annular contact surface provided on the edges of such parts. 10. Device according to claim 8 wherein the casing is composed of transparent material. 11. Shielding device for the protection of human embryo cells from the effects of radiation during examination by X-rays comprising a radiant absorbing protective screen including a central radiant absorbing protective plate of predetermined configuration, a plurality of additional radiant absorbing protective displaceable sheets and adjusting means to displace the sheets to vary the covering surface area of the screen, the central plate and displaceable sheets being arranged to provide a protective screen of elliptical shape which shape is retained over the entire range of adjustment of the adjusting means. 12. Device according to claim 11 including an adjusting ring for adjusting the entire protective screen comprised of the central plate and displaceable sheets between two alternative adjustments which deviate from each other by an angular displacement of 90 degrees. 13. Device according to claim 12 including a supporting arm to which the protective screen is connected and a support column for the supporting arm on which arm is vertically adjustable. 14. Device according to claim 13 wherein the supporting arm is a telescopically adjustable arm. 15. Device according to claim 12 including a pivot arm support for supporting the protective screen as a parallelogram guide.
RE0367605	summary	FIELD OF THE INVENTION This invention relates generally to material modification, and more specifically to a class of techniques whereby a thin layer of material on the surface of a body can be rapidly heated, followed by a rapid quench as the heat energy is conducted into the body. The invention also relates to other types of surface treatment using ion beams to convey energy to a near-surface region. BACKGROUND The mechanical and chemical properties of surfaces are an important factor in almost all materials applications. Numerous techniques have been developed to enhance these properties for particular applications. These techniques range from widely used techniques such as galvanizing and heat treatment of alloys to specialty techniques such as ion implantation and laser glazing which are expensive and not scaleable to common industrial use. The use of surface treatments to improve properties such as surface hardness, wear resistance, corrosion resistance, and fatigue lifetime add significant value to a wide range of products in industries including automobile manufacture, aerospace, microelectronics, tool and die manufacture, power generation, and the production of steel, aluminum, ceramics, and plastics. Thermal treatment to alter the surface properties of materials has been a standard industrial process since the early smiths developed techniques for pack carburization of cast iron in their forges to produce a material more suited to the fabrication of durable swords. Such treatments, principally metallurgical in nature, have thus formed a part of manufacturing technology for more than a thousand years. Two primary driving forces are available through the use of thermal treatments. The first involves the use of high temperature to overcome kinetic barriers which keep something from happening. The second is based on the rapid quenching of hot material to preserve, in some degree, the microstructure of the hot material or that of a metastable structure encountered in the evolution toward the quenched material. The metallurgy of sword blades, although not understood at the time and developed through empirical research, represents the pinnacle of materials engineering during the Dark Ages. It also serves to demonstrate various surface treatment modalities which are still of primary importance. In the Dark Ages, iron was generally produced as wrought iron, i.e., as a matrix of nearly pure iron with a relatively high density of slag inclusions. (Slag is the combination of impurities and the flux used in the reduction of iron ore.) Wrought iron is reasonably tough, but is very soft, and hence not well suited to the production of weapons. It is possible to harden the material by cold-working in the course of shaping, thus producing a work-hardened material, but this material is brittle as well as hard. Thus, a method that hardened only the surface layers would produce a superior weapon, one which would both hold an edge and take an impact without fracture. The process most often used to this end was pack carburization. A sword was forged (hammered) into shape, producing a hard but brittle implement. The sword was then packed in a mixture of carbon (charcoal or coke) and other organic materials. The pack was heated to the carburizing temperature (about 950.degree. C.), whereupon carbon was transferred to the steel by decomposition of carbon monoxide at the iron surface. The carbon diffused into the iron, forming a crude form of hypereuctectoid iron surrounding a wrought iron core. This is a thermodynamically favorable process which requires heating to yield significant diffusion in a reasonable amount of time. The sword was then heated to above the transformation temperature (723.degree. C.). As a result, the iron core was left unchanged, whereas the steel tegument transformed into a mixed phase of austenite and cementite. (Austenite is the .gamma.-phase of iron with dissolved some carbon, while cementite is Fe.sub.3 C, a hard brittle substance.) The sword was then quenched abruptly to prevent the austenite-cementite mixed phase of the tegument from transforming back into the original material. (This step uses the second of the primary thermal treatments, using rapid quenching to preserve a high-temperature structure.) This quenching was usually performed by plunging the hot sword into oil, but some texts insisted that the finest swords resulted from quenching by thrusting the hot sword through the body of a young boy or a virgin maiden. The popularity of these latter practices is not recorded. The structure resulting from the above steps was a strong, tough iron core surrounded by a thin (<1 mm) layer of fine-grain high-carbon steel having high hardness but low toughness. The high carbon-steel tegument of the sword can now be sharpened to a fine edge, which will last through much use, while the soft iron core will allow deformation on impact, thus avoiding brittle fracture of the sword blade. The relative quality of local surface treatments of wrought iron was an important factor in determining the outcome of many early political disagreements. Surface treatment of a wide range of materials is still extremely important in modern manufacturing. Considering only steel for a moment, carburizing is still widely used to produce a hard surface on steels, but has been joined by a wide range of related techniques including nitriding (including iron nitriding), carbonitriding, cyaniding, and liquid carburizing. Thermal treatment alone is also used to harden the surfaces of medium and high-carbon steels, where there is already enough carbon to form the austenite-cementite mixed phase using heating alone. However, in these cases only the surface must be heated, leading to the development of numerous techniques for heating only the surface of a body, including pulsed inductive heating, where heat is generated by induction of eddy currents, which are confined primarily to the near-surface region in a conducting body, direct heating by a flame followed by quenching quickly enough that the inner regions of the body are not heated past the transformation temperature, and laser hardening, in which a very thin layer of the surface is heated quickly enough that quenching is accomplished by simple thermal conduction into the body of the part being hardened. Laser hardening is an early example of the type of process of interest in the present application, i.e., a rapid heat-rapid quench process. Finally, a work hardened surface can also be formed by heating a thin surface layer sufficiently that it ablates from the bulk of the material, creating shock waves which create dislocations in the near-surface regions. Other purposes than hardening can be served by surface treatment of materials. For example, in steels, the chromizing process (diffusing chromium from an external source) increases the corrosion resistance of the material by turning the surface region into a form of stainless steel which may or may not be harder than the initial material, but will resist the action of oxidants and other common sources of corrosion. However, many other materials, including metals, alloys, semiconductors, ceramics, and other nonmetals are subject to corrosive effects. These include galvanic corrosion, in which a current is generated between two dissimilar metals in electrical contact, resulting in a soluble species being generated from one of the metals. Another form of corrosion is pitting, in which a passivation (or protecting) layer is breached in a small area (a pit). The material underneath, which is usually protected by the passivation layer, is then exposed to the corrosive environment, and dissolves inward from the position of the pit. (Note that a passivation layer may be intrinsic, as in the oxide layer on aluminium, or extrinsic, as in the surface layer of zinc in galvanized steel.) Corrosion can also be stimulated by tensile stress, biological organisms, and atmospheric contamination. A side effect of corrosion mechanisms can be hydrogen embrittlement, in which the hydrogen generated by a corrosive process enters into a metal or alloy, reducing the ductility of the material, which is then weakened, allowing surface flaws to grow under a stress, thus increasing the susceptibility of the material to further corrosion and eventual failure under stress. Many methods exist to reduce the pernicious effects of corrosion, but it is estimated that some 4% of the gross national product is still lost to corrosive effects. One type of surface treatment to improve corrosion resistance which is consistent with the class of treatments under discussion here, i.e., rapid heat-rapid quench techniques, is the formation of surface alloy layers by forming a layer of a second material on the object to be protected, and then melting the surface layers to form a homogeneous liquid material. If the quenching is then rapid enough, a solid solution results even if an equilibrium alloy cannot be formed at a low temperature. The solid solution may be amorphous nanocrystalline, microcrystalline, or actual precipitates may form, depending on the rate of quenching. A small amount of precipitation will still allow corrosion resistance to be improved, and may increase the hardness of the treated surface layer. Surface treatments are also important in the control of wear, which constitutes the primary reason why the artifacts of society become useless and have to be replaced. Wear is simply the removal of material from a solid surface as a result of sliding action. Wear is occasionally a useful process (e.g., writing with pencil and paper), but more often is deleterious to both the structure and the operation of mechanisms. There are four primary types of wear, adhesive, abrasive, corrosive, and fatigue wear. Adhesive wear arises from the formation, during sliding, of regions (called junctions) of adhesive bonding on a microscopic scale. If the junctions do not break along their original interfaces upon further sliding, then a chunk from one of the surfaces will have been transferred to the other surface. Such particles constitute wear in their formation, and may also add to abrasive wear. Abrasive wear is produced by a hard object being dragged along a softer one, thereby digging out a groove. The abrasive agent may be one of the surfaces, particles removed from the surfaces by other wear mechanisms, or external particles, such as sand in a bearing. Corrosive wear occurs when sliding action takes place in a corrosive environment, the pieces are nominally protected by a passivating layer, and the sliding action continuously removes the passivating layer, thus exposing fresh surfaces to the action of the corrosive. Fatigue wear occurs as cracks form and grow as the result of fatigue, especially in rolling systems. A crack forms below the surface, and grows to intersect the surface, thereby lifting a large particle out of the surface. The various forms of wear are often synergistic, resulting in a form of degradation which is nearly universal in any mechanism or device having moving parts. Adhesive wear is the most fundamental, existing in any sliding or rotating contact in which two surfaces touch. The primary line of defense against wear is the use of lubricants, which act to prevent contact of surfaces in relative motion, thus reducing wear by as much as a million times the dry value. However, unless the relative velocities of the surfaces is high enough that the surfaces `surf` on a continuous film of lubricant, there will still be contact and adhesive wear will occur. The condition of the sliding surfaces helps to determine the rate of wear. The friction between a pair of sliding surfaces having large surface roughness will be approximately the same as that between similar surfaces having smoother surfaces. The friction in the first case is the work done in tearing apart a few large junctions, whereas in the second case the friction results from tearing apart many small junctions, but the total surface area of the junctions (and hence the work required to tear them asunder) is determined primarily by the amount of deformation of the surfaces caused by the force normal to the surfaces, which is the same in both cases. From the point of view of the wear occurring during sliding, however, the difference in surface roughness makes several important contributions. Take .lambda. as a length characterizing the surface roughness. (In some cases the appropriate length will be the grain size rather than the surface roughness, but the arguments below still hold.) It is clear that .lambda. will also characterize the size of the particles torn off by adhesive wear. That is, if .lambda. is 10 times larger, the size of the detached particles will also be 10 times larger. Smaller particles will contribute less to abrasive wear mechanisms. The difference in size of the detached particles resulting from differing degrees of surface roughness yields another advantage. A simple scaling argument will illustrate this. Assume .lambda..sub.R for the rough surface is 10 times larger than .lambda..sub.S for the smoother surface. Under equal external loading, the areas of contact will be the same. As each junction now has an area on the order of .lambda..sup.2 there must be 100 times as many junctions on the smooth surface as on the tough surface. Further assume, as above, that the characteristic size of the detached particles is .about..lambda.. The volume of material detached by adhesive wear is then .about.N.lambda..sup.2, where N.about..lambda..sup.2 is the number of particles formed by adhesive wear. The total volume of material removed by adhesive wear in a given sliding process is thus proportional to the size of the surface roughness. In summary, smoother surfaces, despite producing the same friction, result in less direct adhesive wear because of the square-cube scaling law above, and also reduce the amount of self-abrasive wear by reducing the size of the abrasive particles generated during adhesive wear. Such surfaces can be fabricated by melting the surface layers of the body and allowing the heat energy to dissipate into the body, thus obtaining rapid quenching of the liquid. Fatigue wear can also be affected by surface morphology, but not primarily by the length scale of the surface roughness. More important here is the presence of abrupt structures, such as cracks, ledges, overhangs, etc., which offer sites for stress concentration, and the earlier material failure accompanying such concentration. Such stress concentration does not depend on the length scale of the defect, but rather on its shape. This type of wear will be reduced if the surface can be treated to have more gradual changes in surface morphology. Again, surface morphology can be altered to provide less opportunity for stress concentration by rapidly melting the surface layers and quenching the heat energy into the body of the material in question. Such smoother surfaces will also serve to limit various mechanisms for corrosion. The potential applications for rapidly heating (and perhaps melting) a thin surface layer which is then self-quenched via thermal conduction into the body of an object are very broad. Beyond those described in detail above, one may alter the surface layers of a material. This may be done in a number of ways, but the main route toward such alterations is the ability of rapid quenching to produce non-equilibrium structures, such as amorphous or nanocrystalline surface layers. Metastable surface alloys can be produced by rapid melting and quenching. This requires a material system in which a thin layer of material A is formed on a substrate B. The phase diagram of these materials is such that they are immiscible when solid, but form a single-phase liquid when molten. (Heating above the melting point may be required.) If this material is then rapidly quenched, an amorphous alloy composed of the two components will result. If the quenching process is somewhat slower, nanoscale precipitates will form. The size of these precipitates depends on the cooling rate. Note that these materials need not be metals. A coating of gold on a germanium substrate melted and rapidly quenched will form such as amorphous alloy. Compounds including members of the metalloids are well known as helpful in formation of amorphous material. When the pulse energy is much greater than that required to melt the heated surface layer, the surface layer will ablate. This can have three desirable effects. One is to serve as a source of pure material for an associated deposition process analogous to sputtering, but providing greatly enhanced purity and smoothness of the deposited material. It is also possible to obtain unique surface structures, which have, for example, altered electron emission characteristics, by ablating a surface layer from a substrate. Some of the ablated material will redeposit on the substrate, forming the aforementioned unique structures. Third, the shock wave created in the substrate by such ablation produces work-hardening effects far into the material (perhaps several hundred microns) through formation of dislocation structures below the heated surface layer. Another application of a rapid heating-rapid quenching cycle is to clean a surface without altering the properties of the surface in any manner. This would be possible when the contaminant will desorb from the surface at a temperature lower than the melting point of the surface. Properly done, only a very thin (<<.mu.m) layer of the surface would have to be heated to remove contaminants compatible with this method without the use of solvents or other chemicals. Such a process could replace many cleaning steps presently required in machining and semiconductor manufacture, to name only two possibilities. This is an important consideration in these days of heightened ecological awareness and regulation. The surfaces of porous and/or highly defective materials such as ceramics can be smoothed and rendered resistant to crack nucleation by forming a surface layer of glass using the same type of rapid heat-rapid anneal treatment. A similar smoothing of surfaces was described in the discussion of wear above. Further applications include `polishing` of machined parts. Precision machined parts will commonly retain machining marks on the order of 10 .mu.m in size. Surface melting can allow the surface tension of the material to induce material reflow, smoothing the surface. Such techniques could also find application in the final polishing of diamond-turned optics, thereby totally avoiding conventional optical surface generation techniques. Finally, smooth surfaces offer fewer flaws to initiate corrosive processes. This class of surface finishing techniques will thus reduce the initial rate of corrosion significantly, beyond any changes in surface chemistry which may also be accomplished. Having established that rapid-heat rapid-quench processes are potentially of great industrial use, one must naturally ask why they are not presently being applied in standard industrial practice. There are numerous reasons why previous laboratory-scale attempts to apply such processes failed to be accepted in the market. Consider the conditions required to melt a thin layer (1.about.1 .mu.m) of a steel surface. How much energy is required to melt the surface assuming that no thermal conduction into the bulk of the steel occurs? The melting point T.sub.m of steel is about 1530.degree. C., the density pis .about.7000 kg/m.sup.3, and the specific heat c is .about.3Nk where N is the number of atoms per kg (.about.1.07.times.10.sup.25), and Boltzmann's constant k is 1.38.times.10.sup.-23 J/.degree.K. The energy per m.sup.2 in a 1 .mu.m thick layer of molten steel is roughly .rho.cT.sub.m .DELTA.1, or some 5600 joules per square meter. As the present interest is in industrial-scale processes, treatment of a square meter of material at a time is not unreasonable, at least for discussion. Now the time required for quenching of this energy into the bulk of the steel must be estimated. The rate of power flow out of the molten layer is roughly kT.sub.m /21 joules per square meter per second. Combining this with the earlier result, and assuming that k is not a function of temperature, ##EQU1## where .tau. us the characteristic time for the heat energy to leave the molten layer through conduction into the bulk of the steel. K is .about.100 watts per meter per .degree.K, so the characteristic time for this situation is roughly 30 nanoseconds. Note that this estimate gives a cooling rate of .about.5.times.10.sup.10 .degree.K per second, a remarkably large value compared to .about.10.sup.6 .degree.K-sec.sup.-1 for techniques such as splat quenching or planar flow casting techniques. Thicker surface layers will require longer cooling periods: for example a 10 .mu.m melted layer on a steel body will cool at .about.5.times.10.sup.8 .degree.K-sec.sup.-1, a value still associated with non-equilibrium effects. In the primary mode of operation, the energy of the thin molten layer must be deposited in a period of time shorter than .tau. so that the deposited energy efficiently heats only the desired surface layer, and not the underlying material. Accordingly, the deposited power P must be greater than the energy deposited divided by the characteristic time period, or EQU P>0.18 terawatts per square meter. A secondary mode of operation is also available, in which the beam energy is deposited in a thin surface layer on a time scale much longer than the characteristic thermal diffusion time for the surface layer being heated. This mode of operation is analogous to flame annealing. Such a mode is useful for annealing the surface layer, to induce grain growth or to produce a thin nanocrystalline surface layer on an amorphous material. In this mode substantial temperature increases will be experienced by much more of the material than the surface layer being directly heated, an effect which must be accounted for when carrying out the ion beam surface treatment. (A worker skilled in the art can use the equations given in this specification to predict thermal profiles (i.e., temperature increase vs. depth vs. time) for a given set of production conditions.) Making this class of techniques even marginally practical (larger molten thickness would be desirable in most cases) requires a source unit that can deliver .about.10.sup.4 joules into the surface of a body in a 30-3000 nanosecond pulse. Further, unless rapid cycling (>>1 Hz) of the source is possible, the amount of material that can be treated per source unit is too small to have an impact on any but specialty items. High process efficiency is also required, as otherwise removing the waste heat from the source unit will become a difficult task, as will providing the total power required. Consider a more definite case. The source unit is to be a pulsed laser. (The difficulties surrounding the problem of depth and consistency of power absorption will be ignored for a moment.) To get 10.sup.4 joules output, our 1% efficient laser will require 10.sup.6 joules input. To provide 10 pulses per second, a minimal speed for practical applications of this technology, the source unit must receive some 10 megawatts continuous input and have a cooling system capable of removing and disposing of nearly that much power continuously. The cost in wasted electricity alone is about $10 million dollars per year of operation. The low power efficiency of laser systems which provide short enough pulses of sufficient energy to treat large areas of a surface is clearly a problem. Lasers present other problems when considered for this class of applications. A pulsed laser system with the required level of power has been developed for antiballistic missile systems, but the physical size and capital cost of each system is enormous. In addition, the lifetime of certain critical components is quite short (<10.sup.3 pulses), requiring enormous downtime for maintenance in an industrial situation. Further, the depth of power deposition is limited to an optical skin depth. As this is much less than a micron for any suitable laser system now available acting on metals, one of two situations will develop. The total energy will be delivered suddenly to one skin depth of surface, which will then vaporize and ablate from the surface. Alternately, the energy can be slowly fed into the surface through the bottleneck presented by the requirement that the outer few nanometers of the surface not vaporize, thus requiring longer pulses with lower power. This option results in long heating periods, and substantial heating of the material underlying the desired surface heating region. Such a situation is non-optimal. Finally, in order to use a laser as a source unit for this class of manufacturing applications, the surface condition of the material presented must be carefully controlled so that the power absorption is uniform throughout the material being treated. Such control in a general industrial manufacturing environment would prove difficult. Another possible source is an ion beam generator. Such generators are able to deposit their energy with reasonable uniformity down to depths of many microns, depending on the energy and species of the ions used, offering some promise for application to the present class of manufacturing processes. It is important to note that the ion beam generator is not being used for ion implantation in the usual sense. There is a great deal of information on alteration of surface and near-surface regions by ion implantation, in which a rapid thermal effect is not the operative driver, but rather the gross changes in chemistry caused by implantation of the ions or the localized lattice damage resulting from slowing of individual ions. The point is that in conventional ion implantation the rate of implantation (i.e., the beam current per unit area) is of little importance as long as the ions eventually are implanted. In the present applications, the thermal effects caused by the extremely high current of ions impacting the surface are primarily responsible for the favorable surface modifications. In most cases the total dose of ions will be small enough to leave the surface composition essentially undisturbed. This point will be discussed in a more quantitative manner below. For the moment the problem of making a suitable ion beam generator will be ignored, and attention placed on the characteristics such a generator must have to function in the modalities described above. Two problems present themselves. First, for given species of ion and target, how much beam energy is required to penetrate a given distance into the target, thus heating the target surface to that depth? Second, what total dosage is required to melt the affected area? The answers to these questions will determine the characteristics required by an ion beam generator useful for surface treatment in a manufacturing environment. The rate as which energy is lost to electronic collisions (the primary mode of energy loss in the relevant regime) by an ion of mass M.sub.1 and atomic number Z.sub.1 while traversing an amorphous (or polycrystalline) target consisting of atoms of mass M.sub.2 and atomic number Z.sub.2 can be expressed in dimensionless units for length (.rho.) and energy (.epsilon.) within the LSS theory (Ion Implantation in Semiconductors, by J. W. Mayer et al., Academic Press, 1970, pgs 21-26) as ##EQU2## The dimensionless parameters are given by EQU .rho.=4100dM.sub.1 /[(M.sub.1 +M.sub.2).sup.2 (Z.sub.1.sup.2/3 -Z.sub.2.sup.2/3)]R(.mu.m), EQU .epsilon.=9500M.sub.2 /[Z.sub.1 Z.sub.2 (M.sub.1 +M.sub.2)]E(MeV), and EQU k=0.0793Z.sub.1.sup.2/3 Z.sub.2.sup.1/2 (M.sub.1 +M.sub.2).sup.3/2 /[(Z.sub.1.sup.2/3 +Z.sub.2.sup.2/3).sup.3/4 M.sub.1.sup.3/2 M.sub.2.sup.1/2 ]. In the above equations, R is distance in microns, E is the ion energy in megaelectron volts, and d is the density of the target material in grams per cubic centimeter. The energy loss equation can be solved for energy remaining after a given distance of travel in the target by substituting .rho.=.eta..sup.2, and integrating to find EQU .epsilon.(.rho.)=k.sup.2 .rho..sup.2 /4-k.rho..epsilon..sub.o +.epsilon..sub.o, where .epsilon..sub.o is the initial dimensionless energy of the ion. Given this equation, the range of an ion in the target material is found by setting .epsilon.(.rho.)=0, and solving the resulting binomial equation for the total range .rho..sub.t to give EQU .rho..sub.t =2.epsilon..sub.o.sup.1/2 /k. Both theory and experiment agree that the energy of the ions is distributed relatively uniformly throughout a volume starting at the surface and proceeding .about..rho..sub.t inward. To give a feel for the above equations, consider a specific example. Carbon ions (Z=6, M=12) having an energy of 1 MeV are incident on an iron (Z=26, M56) surface. The dimensionless energy .epsilon. is equal to 50 E(MeV), the dimensionless range .rho. is equal to 6.2 R/(.mu.m), and k=0.37. (All numerical values are approximate.) By the above range equation, .rho..sub.t =2.times.50.sup.1/2 /0.37=38.2. Solving for the actual distance R.sub.t (.mu.m)=.rho./6.2=38.2/6.2=6.2 .mu.m range. The energy distance relationship calculated here is by no means universal, but serves to illustrate that when thermal heating of a surface layer having a thickness of several microns is desired, the ion energy required to likely to be on the order of 1 MeV. The question of the total dosage required to melt a surface layer can now be illustrated. Continuing with the above example, the earlier estimate that 5600 J/m.sup.2 is required to melt a 1 .mu.m layer of steel shows that some 3500 J/m.sup.2 is required to be deposited to melt the 6.2 .mu.m surface layer heated by a 1 MeV carbon beam. This amount of energy is also equal to 2.2.times.10.sup.17 MeV. The process of melting the 6.2 .mu.m surface layer thus requires an addition of carbon ions amounting to about 2% of a monolayer. The affected region is some 10.sup.4 S of monolayers thick, so in this example the contamination of the surface layers by carbon is on the order of 1 part in 10.sup.6, an amount negligible to the chemistry of most surface modification processes. This demonstrates that the effect of high energy pulsed ion beams is due almost totally to thermal heating of the surface layers, a process made very different from ion implantation by the time scales involved. Finally, the beam current required can be estimated. A pulse of 1 MeV carbon ions must consist of 2.2.times.10.sup.17 ions if a square meter is to be treated in a single pulse. This amount of ions must be transmitted in no more than a few hundred nanoseconds. (The timescale is longer because of the increased thickness of the melted zone.) The resulting rate is about 10.sup.25 ions/sec, representing a current of about 1.6 megaamperes. The pulse must therefore carry a power of some 1.6 terawatts per square meter of surface treated. The size of this number explains why so few experimental studies of surface modification using the thermal effects of ion beam treatment have been made. Given that the use of ion beam generators for surface modification can be carried out as described above, why is there currently so little penetration of commercial markets? The use of ion beams for thermally altering the near surface characteristics of a material has been fraught with substantial problems. Most notable of the limitations with existing ion beam technologies have been: 1) high costs per area treated; 2) the inability to generate a large number of pulses without the costly replacement of ion beam generator components; 3) low repetition rates; 4) low average power; and 5) the inability to reliably produce a uniform ion beam of a single selectable ion species. Typical ion beam generators use dielectric surface arcing on an anode as a source of ions and thereafter magnetically or geometrically direct and focus the generated ion beam onto the material of interest. This surface arcing (also called "flashover") destroys the anode surface in less than 100 pulses, and produces a mixed species of ions that cannot be adjusted. Other difficulties arising from flashover include: production of large quantities of neutral gas that makes high repetition rate difficult, generation of debris which can contaminate surfaces being treated, and non-uniformity and irreproducibility of the beam in some cases due to the localized and difficult to control nature of flashover. State-of-the-art ion beam generators are typically "one shot" devices, i.e., they operate at low repetition rate (<<1 Hz). Existing ion beam generators cannot be operated at high repetition rates (<<1 Hz) for a number of reasons. First, existing pulsed power supplies are no able to generate electrical pulses at high repetition rates having the voltage, pulse width (i.e., normal temporal duration), and power required to generate the ion beams needed (i.e., consistent with the discussion above) for the various beneficial applications described herein. This limitation renders commercial exploitation impractical. Second, the design of existing ion beam generators does not allow repetitive operation for an extended number of operating cycles (<<10.sup.3) without replacement of major components. This limitation would require a maintenance time--manufacturing time ratio incompatible with routine manufacturing operations. Fourth, existing ion beam generators generally operate with electrical efficiencies<5%, thus presenting major challenges to the pulsed power supply and the cooling system of the generator. These limitations and others have made it impossible to routinely utilize the ion beam technology described above for surface treating materials. The present invention generates high energy, repetitive ion beams which overcome the limitations of existing ion beam generators and provides a cost-effective processing technology for thermally altering the near surface characteristics of materials. SUMMARY OF THE INVENTION The present invention provides a system for generating a high energy, pulsed ion beam repetitively over an extended number of operating cycles. In particular, the present invention provides an ion beam generator capable of repetitive operation over an extended operating cycle suitable for thermally treating large surface areas of a material at low cost. This ion beam generator comprises a high voltage, high current pulsed power system and a pulsed ion beam source, both capable of high repetition rates and both having the capability for an extended operating life. This new technology enables the cost effective, commercial scale use of ion beam surface treatment. High energy, repetitively pulsed ions beams produced according to the present invention can produce surface treatments ranging from high temperature anneals through melting to ablation, followed by rapid thermal quenching to ambient temperatures. The control variables for the ion beam treatment include ion species, kinetic energy of the ions, temporal width of the ion pulses, and total dose of ions (which total dose may be delivered in a number of discrete pulses). Variation of these parameters allows uniform heating, to a desired temperature, of a surface region of any material whose thickness approximately corresponds to the range of the ions used in the material being heated. The present invention enables a commercially practical method allowing surface treatment of a material for numerous purposes described herein using pulsed ion beams, comprising the steps of: generating a repetitively pulsed ion beam; and irradiating a surface of the material with the repetitively pulsed ion beam to heat said surface to a predetermined degree to a predetermined depth, wherein the pulsed ion beam has a duration less than about 1 .mu.s and beam energy greater than or equal to 0.05 MeV. Further detail concerning the makeup of the present invention and its scope of applicability will become apparent from the detailed discussion of specific embodiments presented below. However, it should be understood that the detailed description and specific examples, while representing the class of embodiments of the invention and its uses, are given by way of illustration only since various changes and modifications within the spirit and scope of the invention will be apparent to those skilled in the art from this detailed description.
	abstract	A method and apparatus satisfying growing demands for improving the precision of angle of incidence of implanting ions that impact a semiconductor wafer and the precision of ribbon ion beams for uniform doping of wafers as they pass under an ion beam. The method and apparatus are directed to the design and combination together of novel magnetic ion-optical transport elements for implantation purposes. The design of the optical elements makes possible: (1) Broad-range adjustment of the width of a ribbon beam at the work piece; (2) Correction of inaccuracies in the intensity distribution across the width of a ribbon beam; (3) Independent steering about both X and Y axes; (4) Angle of incidence correction at the work piece; and (5) Approximate compensation for the beam expansion effects arising from space charge. In a practical situation, combinations of the elements allow ribbon beam expansion between source and work piece to 350 millimeter, with good uniformity and angular accuracy. Also, the method and apparatus may be used for introducing quadrupole fields along a beam line.
	claims	1. A laser fusion device comprising:a target shell that generates a nuclear fusion reaction;a chamber in which the target shell generates the nuclear fusion reaction;a target shell supply device that supplies the target shell to a reference point on an inner side of the chamber;a target shell monitoring device that monitors a state of the target shell supplied to the chamber by the target shell supply device;a compression laser output device that outputs a compression laser light for compressing the target shell to the target shell;a heating laser output device that outputs a heating laser light for heating the target shell to the target shell; anda control device that controls the target shell supply device, the compression laser output device, and the heating laser output device, whereinthe target shell has a hollow spherical shell shape, includes, on an inner side of the target shell, a spherical space demarcated by an inner surface of the target shell, includes at least one through hole connecting an outer side of the target shell and the space on the inner side of the target shell, and includes, on an outer surface of the target shell, an irradiation area to be irradiated with a compression laser light, andthe compression laser output device includesa compression laser that outputs the compression laser light for compressing the target shell; anda condensing optical device for compression, that condenses the compression laser light, which is output from the compression laser, toward the irradiation area of the target shell, andthe heating laser output device includesa heating laser that outputs a heating laser light for heating the target shell; anda condensing optical device for heating, that condenses a heating laser light, which is output from the heating laser, toward the through hole of the target shell, andthe control device includesa target shell supply unit that controls the target shell supply device so as to supply the target shell to the chamber,a timing calculation unit that calculates an arrival timing, at which the target shell supplied by the target shell supply unit arrives at the reference point, based on a monitoring result by the target shell monitoring device, and calculate, based on the arrival timing, a compression laser light output timing at which the compression laser outputs the compression laser light and a heating laser light output timing at which the heating laser outputs the heating laser light following the compression laser light output timing,a light condensing unit that controls the condensing optical device for compression based on the monitoring result by the target shell monitoring device so as to condense the compression laser light, which is output from the compression laser, toward the irradiation area of the target shell at the compression laser light output timing calculated by the timing calculation unit, and control the condensing optical device for heating based on the monitoring result by the target shell monitoring device so as to condense the heating laser light, which is output from the heating laser at the heating laser light output timing calculated by the timing calculation unit, toward the through hole of the target shell; andan output unit that controls the compression laser so as to output a compression laser light at the compression laser light output timing calculated by the timing calculation unit, and control the heating laser so as to output a heating laser light at the heating laser light output timing calculated by the timing calculation unit, after the condensing optical device for compression and the condensing optical device for heating are controlled by the light condensing unit. 2. The laser fusion device according to claim 1, wherein the target shell includes the two through holes, and the two through holes extend along a central axis passing through a center of the target shell and opposes each other across the center. 3. The laser fusion device according to claim 1, wherein the target shell is a polymer including deuterium or tritium. 4. A nuclear fusion generating method comprising:a supply process that supplies a target shell, which generates a nuclear fusion reaction, to a chamber;a preparation process that prepares for, after the supply process, irradiation with a compression laser light for compressing the target shell and a heating laser light for heating the target shell after compression; andan irradiation process that irradiates the target shell with the compression laser light and the heating laser light in sequence after the preparation process, whereinthe target shell has a hollow spherical shell shape, includes, on an inner side of the target shell, a spherical space demarcated by an inner surface of the target shell, includes at least one through hole connecting an outer side of the target shell and the space on the inner side of the target shell, and includes, on an outer surface of the target shell, an irradiation area to be irradiated with a compression laser light, andthe preparation process includesa calculation process that calculates the compression laser light output timing at which a compression laser light is output and the heating laser light output timing at which a heating laser light is output based on a monitoring result while monitoring a state of the target shell supplied to the chamber in the supply process; anda light condensing process that condenses, based on the monitoring result, the compression laser light toward the irradiation area of the target shell at the compression laser light output timing and the heating laser light toward the through hole of the target shell at the heating laser light output timing, andin the irradiation process, the compression laser light is irradiated to the irradiation area of the target shell at the compression laser light output timing, and the heating laser light is irradiated to the through hole of the target shell at the heating laser light output timing following the irradiation with the compression laser light. 5. The nuclear fusion generating method according to claim 4, wherein the target shell includes the two through holes, and the two through holes extend along a central axis passing through a center of the target shell and opposes each other across the center. 6. The nuclear fusion generating method according to claim 4, wherein the target shell is a polymer including deuterium or tritium.
	summary
	claims	1. An apparatus for separating radioactive nuclides from a waste salt and recovering a refined salt, comprising:a vaporization chamber having a reaction container installed therein for accommodating a waste salt;a first top cover provided with a stirrer configured to agitate the waste salt accommodated in the reaction container of the vaporization chamber;a second top cover provided with a first electric heater configured to distill the waste salt under a reduced pressure and mutually replaceable with the first top cover provided with the stirrer when a chemical conversion reaction of the waste salt using the stirrer is completed;a condensation chamber in which salt steam produced by heating the vaporization chamber is condensed and liquefied;a decompression device configured to decompress inner parts of the vaporization chamber and the condensation chamber to a predetermined pressure;a recovery container which is disposed at the bottom of the condensation chamber and in which the salt vapor liquefied at the condensation chamber precipitates to be recovered; anda bottom cover switchably installed at the bottom of the condensation chamber to unload the recovery container. 2. The apparatus of claim 1, wherein a baffle coupled to the stirrer is installed in the reaction container to be adjacent to an inner wall of the reaction container. 3. The apparatus of claim 2, further comprising an upward/downward driving device driven to lift the stirrer and the baffle upward at the same time. 4. The apparatus of claim 1, further comprising a top cover opening/closing device configured to automatically open and close the first top cover or the second top cover. 5. The apparatus of claim 1, further comprising a bottom cover opening/closing device configured to automatically open and close the bottom cover. 6. The apparatus of claim 1, further comprising:a second electric heater configured to heat the vaporization chamber; anda third electric heater and a fourth electric heater configured to heat upper and lower portions of the condensation chamber, respectively,wherein the heating by the second electric heater, the third electric heater and the fourth electric heater is controlled in sequentially decreasing temperatures so that a temperature gradient in the vaporization chamber and the condensation chamber is formed to facilitate the flow of salt vapor. 7. The apparatus of claim 6, wherein the vaporization chamber is decompressed to a predetermined pressure by means of the decompression device while the waste salt is heated by the second electric heater to a predetermined temperature at which the waste salt is able to be smoothly vaporized. 8. The apparatus of claim 1, wherein a cooling water circulation passage is formed at the bottom of the recovery container.
	description	This application is a continuation-in-part of U.S. Non-Provisional application Ser. No. 13/826,293 filed Mar. 14, 2013, now U.S. Pat. No. 9,396,822, which claims priority to U.S. Provisional Application No. 61/611,585 filed Mar. 16, 2012, both hereby incorporated by reference as if submitted in their entirety. In a nuclear power generation facility, a nuclear reactor core (a/k/a, reactor vessel) contains nuclear fuel rods, and is equipped to initiate, control, and sustain nuclear chain reaction in the nuclear fuel rods to generate heat. The heat generated by the nuclear reaction is absorbed by a circulating primary coolant into which the fuel rods are immersed, maintaining a stable operating temperature. The circulating coolant limits the operating temperature and thus keeps the nuclear reaction in control; it also carries away the heat generated by the controlled nuclear reaction which is in turn used to produce pressurized steam that drives a turbine. The turbine in turn drives a power generator to produce electricity. The most common types of nuclear reactors use closed-loop circulating purified water as the primary coolant, which in Boiling Water Reactors is boiled into steam by the nuclear reaction to drive a turbine, and then condensed back into liquid phase to be cooled with separate cycling cooling water drawn from a large body of external supply, such as a sea, a river, or a lake. An older version of nuclear power reactor design, the Pressured Water Reactor (PWR) further separates the reactor vessel coolant from the heated water that generates steam to drive a turbine. The nuclear fuel material is contained in a tube-like rod made with radiation-neutral zirconium alloy. Such an assembly is called a fuel rod. During the reactor operation in the power generating mode, the surface temperature of the fuel rod cladding is normally kept at approximately 280 degrees Celsius. The nuclear reaction is further moderated and controlled by control rods inserted between the fuel rods to absorb neutrons generated by nuclear reactions in the fuel rods. The zirconium alloy is transparent to neutrons, which is the reason it is used as the cladding material for the nuclear fuel. Unfortunately, when heated to 550 degrees Celsius or above, zirconium reacts with steam and generates hydrogen, which is highly explosive at high temperature and the presence of oxygen. Explosions in and around a nuclear reactor in such a situation are certain to spew extremely dangerous radioactive material into the environment. Furthermore, the disintegration of the fuel cladding causes the nuclear fuel material to fall to the bottom of the reactor vessel to continue the out of control nuclear reaction and the continued elevation of temperature, which is called a melt-down. Even when the reactor is shut down and the stimulated chain reaction is stopped, the nuclear fuel will continue its intrinsic decay and reaction, with the generated heat spontaneously increasing the rate of reaction until the remaining reactive material is entirely spent. Cooling the nuclear material therefore is critical to keep the reaction under control and below a threshold rate that can cause spontaneous acceleration. Above that threshold the spontaneous acceleration of reactive processes will lead to out-of-control conditions which may result in harmful radiation with the accompanying radioactive by-products to be released into the environment. Therefore, whether the fuel rods remain in the reactor vessel, or are kept in storage outside the reactor vessel, the continual cooling of the fuel rods is required at all times. Even spent nuclear fuel is typically stored in cold water pools and needs to be continually cooled for several years before the spent fuel rods can be safely removed to dry and permanent storage. When an accident, equipment malfunction, loss of power, or operator error causes the reactor to lose cooling, which is conventionally facilitated only by electrically powered pumps circulating the primary and secondary coolants immersing the fuel rods, the fuel rods will rapidly heat up resulting in a self-propelling cycle of increased heating and accelerated nuclear activity, soon reaching the critical temperature of 550 degrees Celsius and higher, where the zirconium cladding will react with steam. In the presence of water vapor in the vicinity of the fuel rods, the zirconium and steam react to immediately generate copious amounts of explosive hot hydrogen gas. When brought into contact with any form of oxygen in the environment, disastrous explosions result until the zirconium and the nuclear reactive material are both exhausted, the environment totally destroyed, or safe cooling is installed and the temperature is brought under [email protected]=Zr(OH).sub.4-4H.sub.2 In addition, when zirconium alloy casing disintegrates during its reaction with steam, it allows the nuclear fuel pallets to drop to the bottom of the steel reactor vessel, out of reach of all other conventional nuclear activity control mechanisms that may still be functioning. The fuel temperature in that case would continue to rise even more rapidly until the fuel melts and forms a pool at the bottom of the reactor vessel which can burn through the vessel wall into the floor of containment chamber, and even melt through the containment chamber floor, and expose the molten nuclear fuel, its continual nuclear reaction, and massive radioactive by-products into the environment. This is called a nuclear meltdown. In the case of the recent Fukushima Nuclear Reactor crisis in Japan, the electrically powered cooling system failed through the earthquake, and the backup electrical power generators failed due to the tsunami flooding. The fuel rods in the six reactors and the cold water pools that store the spent fuel lost cooling. After the tsunami passed, the Japanese Government and TEPCO (Tokyo Electric Power Corporation) operators used portable generators and pumps to pump sea water into the power plants and reactor vessel to cool the overheated fuel rods. Concerned that the sea water is highly corrosive, under Japanese Government's request for assistance, US ships shipped a vast amount of purified water to the Fukushima site for cooling the reactor vessels. Unfortunately, the vast amount of steam thus generated, whether from sea water or from purified water, interacted with the overheated zirconium cladding of the fuel rods (at and above 550 degree C.) and produced copious amounts of hydrogen gas, resulting in repeated explosions. The figures and descriptions provided herein may have been simplified to illustrate aspects that are relevant for a clear understanding of the herein described devices, systems, and methods, while eliminating, for the purpose of clarity, other aspects that may be found in typical devices, systems, and methods. Those of ordinary skill may recognize that other elements and/or operations may be desirable and/or necessary to implement the devices, systems, and methods described herein. Because such elements and operations are well known in the art, and because they do not facilitate a better understanding of the present disclosure, a discussion of such elements and operations may not be provided herein. However, the present disclosure is deemed to inherently include all such elements, variations, and modifications to the described aspects that would be known to those of ordinary skill in the art. References in the specification to “one embodiment,” “an embodiment,” “an illustrative embodiment,” etc., indicate that the embodiment described may include a particular feature, structure, or characteristic, but every embodiment may or may not necessarily include that particular feature, structure, or characteristic. Moreover, such phrases are not necessarily referring to the same embodiment. Further, when a particular feature, structure, or characteristic is described in connection with an embodiment, it is submitted that it is within the knowledge of one skilled in the art to affect such feature, structure, or characteristic in connection with other embodiments whether or not explicitly described. Additionally, it should be appreciated that items included in a list in the form of “at least one A, B, and C” can mean (A); (B); (C); (A and B); (A and C); (B and C); or (A, B, and C). Similarly, items listed in the form of “at least one of A, B, or C” can mean (A); (B); (C); (A and B); (A and C); (Band C); or (A, B, and C). In the drawings, some structural or method features may be shown in specific arrangements and/or orderings. However, it should be appreciated that such specific arrangements and/or orderings may not be required. Rather, in some embodiments, such features may be arranged in a different manner and/or order than shown in the illustrative figures. Additionally, the inclusion of a structural or method feature in a particular figure is not meant to imply that such feature is required in all embodiments and, in some embodiments, may not be included or may be combined with other features. 1. Reactor Chamber, or Reactor Pressure Vessel (RPV) 2. Nuclear fuel elements (rods) 3. Control rods 4. Circulation pumps 5. Control rod motors 6. Steam 7. Feed-water 8. High pressure turbine (HPT) 9. Low pressure turbine 10. Generator 11. Exciter 12. Condenser 13. Coolant 14. Pre-heater 15. Feed-water pump 16. Cold water pump 17. Concrete enclosure, the Containment Chamber 18. Connection to electricity grid 19. Pressure Release Outlet and Valve 20. Coolant Intake or Outlet with control valve 21. Coolant Intake or Outlet with control valve. One or more embodiments use liquid nitrogen, the densest and highly transportable form of nitrogen, and the cold nitrogen gas it produces when released from its container, for emergency cooling of fuel rods and nuclear reaction chambers. An embodiment includes a liquid nitrogen back-up cooling system to be installed in addition to the existing water cooling systems in today's nuclear power plants to improve nuclear power reactor safety. Nitrogen gas is inert. Under atmospheric pressure, its boiling point is minus-196 degrees Celsius. Liquid nitrogen is kept in a well-insulated and pressurized container to help retain its liquid phase. No electricity or pumping action is required to release liquid nitrogen from its container into a warmer or less pressurized environment, making it a highly effective, safe, and rapid coolant when no electrical power is available, or when the nuclear reactor and fuel of concern is already over heated. Furthermore, Boron powder can be incorporated with liquid nitrogen or compressed cold nitrogen gas to absorb neutrons released from nuclear fuel to further dampen and slow the nuclear reaction. Liquid nitrogen is an extremely compact, extremely cold, and readily transportable source of the inert nitrogen gas. Further, its temperature at atmospheric pressure being minus 196 Degrees C., far below the freezing point of water, makes it extremely useful in a wide range of applications, including as an open-cycle refrigerant or coolant. Liquid nitrogen and nitrogen gas are both chemically inert and do not react with other substances. The large quantity of nitrogen gas evolved from evaporating liquid nitrogen as it comes in contact with high temperature objects, dilutes and displaces other gases present, such as hydrogen or oxygen, making it an asphyxiate gas, capable of putting out fire, and maybe hazardous to breathing due to the diluted oxygen content. Liquid nitrogen is self-pressurized in its container. When released from its container, liquid nitrogen ejects with force. The higher the container is pressurized, the higher the ejection force and its reach. In an embodiment, for emergency cooling after an event that caused power loss and over-heating of the reactor chamber and fuel rods is suspected, one can spray the exterior of the reactor chamber immediately to help reduce the temperature within the chamber through convection. Simultaneously, preparation to inject liquid nitrogen into the reactor chamber should be taken, and injection directly inside the reactor chamber toward the fuel rods, with the goal to reach the fuel rods should ensue as quickly as possible. Both measures require no pumping action and no electricity. One can connect the outlet of an adequately large, filled and pressurized liquid nitrogen container to an insulated flexible hose. The hose can be directed to the spraying action for the first measure, and to a water/coolant inlet of the reactor chamber for the second measure, and open the outlet valve of the liquid nitrogen container. If the inlet valve of the coolant inlet of the reactor chamber is not already open, it should also be opened. After a nuclear accident, when radioactive material may already be in the area, these activities should be accomplished by using a robot, or using a trained operator with adequate protective gears. When the outlet valve is opened, pressurized liquid nitrogen in its container automatically ejects out of the LN storage container into the environment, or into the reactor chamber the hose is connected to. As stated above, neither pumping action nor electricity is needed to achieve this ejection. One can similarly fill the containment chamber surrounding the reactor chamber with the cold nitrogen gas produced by releasing liquid nitrogen from its container into the containment chamber. Alternatively, compressed nitrogen gas tanks can be used for this purpose. Liquid nitrogen is widely available, and contains a far-higher quantity of nitrogen gas within the volume of a given container; thus, it is far more efficient to transport than is compressed nitrogen gas. The injection of nitrogen into an environment where oxygen and hydrogen are present greatly dilutes the density and proximity of the oxygen and hydrogen, and may thus prevent an explosion caused by the chemical interaction between hydrogen and oxygen. The pressurized liquid nitrogen injected into the reactor chamber and the extremely cold nitrogen vapor evolved from liquid nitrogen will quickly quench the fuel rods, condense the steam, and freeze the remaining water in the chamber. This cooling method practically reduces the explosion risk of any hydrogen that may have already gathered in the chamber before the injection of liquid nitrogen to near zero, and prevents further generation of hydrogen. The nitrogen vapor produced from the liquid-phase nitrogen during this fuel rod and chamber cooling process is also very cold, and when released into the atmospheric environment outside of the reactor chamber, the nitrogen vapor dilutes and displaces the oxygen in the environment, greatly reducing the explosion risk of any hydrogen present. Furthermore, boron powder or other neutron absorbing materials or compounds can be placed inside the liquid nitrogen container before filling it with liquid nitrogen. Using the ejection force of liquid nitrogen, the boron powder is mixed with liquid nitrogen and injected into the reactor chamber. Boron slows or halts the nuclear reaction by absorbing neutrons generated by the spontaneous decay and reaction of the nuclear fuel. One method, in an embodiment, is to install valves and inlet/outlet at both ends of an LN container intended for use at a nuclear power plant. Subsequently, insert boron powder into such an empty LN container before filling it with LN. Connect one inlet/outlet to a large liquid nitrogen source, and the other inlet/outlet to the target nuclear reactor chamber. First open the valve of the LN container outlet and the inlet to the target nuclear reactor chamber. Then open the inlet of LN container to allow LN from a large LN container to flow into the double inlet/outlet smaller LN container filled with Boron powder. The pressurized ejection of LN from the large source container through the boron filled container, and then into the reactor chamber will spray the boron powder into the reactor chamber and coat objects within. In an embodiment, one may partially fill an LN container with boron powder before filling it with LN, and rotate the container about an axis thereof to combine gravity feed with the ejection force of the LN to also eject the boron powder along with the ejecting LN into the reaction chamber to coat the fuel rods. This process can be repeated by replacing emptied out LN/Boron container with a full one, and releasing nitrogen along with boron as long as needed, until the fuel rods are cooled to a sufficiently cool temperature, and the nuclear activity is sufficiently low. Another embodiment of the invention is to install an automated liquid nitrogen backup cooling system to a nuclear power reactor chamber, wherein the liquid nitrogen container is built with or connected to a temperature and/or pressure activated valve which opens at a predetermined threshold temperature or pressure to allow LN to inject into the reactor vessel. The LN container build for this purpose may also contain two ports. One port may be connected to a nuclear reactor chamber or a containment chamber and equipped with said temperature and/or pressure activated valve pre-set at a threshold temperature or pressure. The 2nd port may be connected to a refilling large on-site LN storage tank. This port can be an open connection to the refill large tank, or equipped with a pressure-activated valve, which opens when the pressure in the smaller LN container drops to indicate a significant ejection of the LN content into the reactor or containment chamber. The wall of the reactor vessel may be optionally coated with a ceramic thermal insulator such as boron nitride to insulate the nitrogen injected into the reactor chamber from external ambient temperature to help maintain its low temperature. Another embodiment may have the pre-installed LN insulated containers (Dewar) prefilled with a desired amount of boron powder. Another embodiment may have an additional onsite large LN storage tank that automatically feeds into the LN container directly hooked to the reactor chamber, as its content LN is emptying into the reactor chamber. Another embodiment of the invention may include connecting a filled LN container to a containment chamber, through a temperature gauge. An additional onsite large LN storage tank can be connected to the LN container that is connected to the containment chamber, and automatically feeds LN into the LN Dewar hooked to the containment chamber, as its LN content is emptying into the containment chamber. Another embodiment may have one onsite large LN storage tank that automatically pressure feeds the LN Dewar hooked to the reactor vessel and the LN Dewar connected to the containment chamber. The walls of pools containing spent fuel cells may also be coated with thermal insulating material, and an insulated pool cover installed. When rapid cooling is needed, liquid nitrogen or boron infused LN can be injected into the pool to cool the spent fuel. Even without thermal insulation coating on the walls and the pool cover, a spent fuel pool still may have a temperature activated backup LN nitrogen cooling equipment installed. Another embodiment may be to install an LN fabrication plant that automatically manufactures and fills the large LN storage tank and the smaller LN containers that are directly connected to fill reactor chambers, containment chambers, and spent fuel pools. In case of power outage, the LN fabrication plant can be powered by back up generators and batteries located at higher levels. This is particularly of interest in locations with multiple reactors, such as the Fukushima site. Regardless of what cooling method is used (even with water cooling), there will be significant amounts of gas that require release. The Fukushima explosions were caused by the hot hydrogen gas released from the reactor chamber reacting with oxygen in the environment. A bank of “scrubbers” that take out radioactive components and dangerous gases such as hydrogen from the exhaust and pressure-relief outlet may be employed in an embodiment. Scrubbers are used in the semiconductor industry to clean exhaust, and are highly effective in eliminating, or at least limiting dangerous outgas released into the environment. While this may not completely eliminate the dangerous outgas problem, it at least will reduce it significantly. At Fukushima, they just vented to the atmosphere. Venting through a scrubber would have been much preferred. However, it is necessary that a bank of scrubbers installed prior an accident. It is difficult to install scrubbers after a disaster has occurred, and dangerous radioactive material along with dangerous Hydrogen and other gases are spewing. Another embodiment of the invention is to use liquid nitrogen to extinguish fire and prevent explosions in other circumstances that using water is not desirable, either due to presence of material that can be irreparably damaged by water, or the presence of chemicals which will react with water, or there is simply no access to water. Ejecting liquid nitrogen toward a heated target area produces a very large quantity of nitrogen gas, starving the targeted area of oxygen while also cooling the area and objects within, thus extinguishing fire and preventing explosions. This embodiment may be implemented by fitting a fire truck with a very large insulated liquid nitrogen container similar to trucks used in transporting a large quantity of liquid nitrogen, pressure gauges and valves for controlling the ejection of liquid nitrogen to target, and adequate length of thermally insulated hose to conduct the ejecting liquid nitrogen to target. Museums, luxury homes containing fine furnishing and art, chemical plants, science labs can be fitted with sprinkling system that ejects liquid nitrogen instead of water. The liquid nitrogen storage that connects to the sprinkling system is monitored and refilled periodically as needed. A heating element can be added inside the liquid nitrogen storage container to assist controlled increase of liquid nitrogen rejection rate. For fighting forest fires, liquid nitrogen containers designed with thermally activated release mechanism and pre-filled with liquid nitrogen can be air-dropped, or mechanically propelled into the site of fire. The released liquid nitrogen meeting fire and heat would immediately convert to gaseous nitrogen and starve the fire of oxygen to put out the fire. One of the mechanical propulsion methods can be as simple as sling cannon. One of the simplest disposable release mechanisms is a welded joint that joins the lid of the liquid nitrogen container with the rest of the container using a material that melts or separates at above a desired pre-set temperature. FIG. 1 illustrates conceptual schematics of a Boiled Water Nuclear (BWN) Reactor and its cooling system. Items 1-18 as labeled are common components in a conventional BWN reactor. Vent 19 and valve controlled additional coolant inlet/outlet 20 and 21 may or may not exist in a conventional BWN reactor chamber. If exist, 20 and/or 21 and can be used for injecting liquid nitrogen and the cold nitrogen gas evolves from the liquid nitrogen into the reactor chamber to cool the chamber and the fuel rods within. If 19, 20, and 21 do not exist, steam outlet 6 can be disconnected from the Turbine and used as an emergency over-pressure relieve outlet, and coolant water inlet 7 can be used to inject liquid nitrogen and cold nitrogen gas into the reactor chamber. The same principle and practice can be used in all other types of nuclear reactors. In a conventional nuclear power plant, back up cooling is driven by backup power connected to its regular cooling system when regular power source is lost. FIG. 2: During emergency cooling, a liquid nitrogen Container 22 is brought to site and connected to a Coolant Inlet/Outlet 20 of the Reactor Chamber. If feasible, a second LN Container 23 can be connected to Inlet/Outlet 21. Alternatively, a vacuumed container can be connected to Outlet 21, and used to receive over pressurized nitrogen gas from the reactor chamber. FIG. 3: In a pre-installed back-up liquid nitrogen cooling system, at least one filled liquid nitrogen container 24 is connected to a nuclear reactor chamber through a thermally activated valve 25. When ambient temperature in the Chamber rises to a predetermined safety threshold temperature, the liquid nitrogen in the container is released into the reactor chamber to quickly bring down the chamber and fuel rod temperatures. The liquid nitrogen containers may also be partially prefilled with boron powder to help slow down the nuclear chain reaction. FIG. 4 shows a very large liquid nitrogen storage container 41, linked to at least one smaller liquid nitrogen container 40 that is connected to a nuclear reactor chamber or a containment chamber through a thermally activated valve 25. The large storage container 41 automatically fills the smaller LN container(s) 40 connected to the reactor chamber or containment chamber as its LN content is released into the reactor chamber or the containment chamber. FIG. 5 shows a LN fabrication facility 50 which produces liquid nitrogen on-site to fill at least one very large liquid nitrogen storage container 51, which is linked to at least one liquid nitrogen container 52 that is connected to a nuclear reactor chamber or a containment chamber through a thermally activated valve 25. The fabricator automatically fills the large storage container, which automatically fills the smaller LN container connected to the reactor chamber or containment chamber as its LN content is released into the reactor chamber or the containment chamber. FIG. 6 shows a container, with a lid and a welding, that can be propelled by a sling cannon to a fire. A moveable vehicle such as a fire truck shown in FIG. 6 can transport liquid nitrogen to the fire. The fire can be a chemical fire. Another container shown in FIG. 6 includes a temperature activated valve and a port system. The port system gauges remaining liquid nitrogen in the liquid nitrogen container and activates refilling of the container in response to the port system determining remaining liquid nitrogen is below a predetermined threshold. A sprinkling system shown in FIG. 6 has a heat sensor controlled valve. While the present invention has been described in terms of preferred embodiments, it will be appreciated by one of ordinary skill that the spirit and scope of the invention is not limited to those embodiments, but extend to the various modifications and equivalents as defined in the appended claims.
	description	This application claims priority to U.S. Provisional Application Ser. No. 60/988,348 filed on Nov. 15, 2007, the specification of which is herein incorporated by reference in its entirety. The invention relates to a cooling system for a nuclear reactor. In nuclear reactors designed with passive operating systems, the laws of physics are employed to ensure that safe operation of the nuclear reactor is maintained during normal operation or even in an emergency condition without operator intervention or supervision, at least for some predefined period of time. A Multi-Application Small Light Water Reactor project conducted with the assistance of the Idaho National Engineering and Environmental Laboratory, NEXANT and the Nuclear Engineering Department of Oregon State University sought to develop a safe and economical natural light water reactor. FIG. 1 illustrates a nuclear reactor design 5 that resulted from this project. The nuclear reactor design 5 includes a reactor core 6 surrounded by a reactor vessel 2. Water 10 in the reactor vessel 2 surrounds the reactor core 6. The reactor core 6 is further located in a shroud 22 which surround the reactor core 6 about its sides. When the water 10 is heated by the reactor core 6 as a result of fission events, the water 10 is directed from the shroud 22 and out of a riser 24. This results in further water 10 being drawn into and heated by the reactor core 6 which draws yet more water 10 into the shroud 22. The water 10 that emerges from the riser 24 is cooled down and directed towards the annulus 23 and then returns to the bottom of the reactor vessel 2 through natural circulation. Pressurized steam 11 is produced in the reactor vessel 2 as the water 10 is heated. A heat exchanger 35 circulates feedwater and steam in a secondary cooling system 30 in order to generate electricity with a turbine 32 and generator 34. The feedwater passes through the heat exchanger 35 and becomes super heated steam. The secondary cooling system 30 includes a condenser 36 and feedwater pump 38. The steam and feedwater in the secondary cooling system 30 are isolated from the water 10 in the reactor vessel 2, such that they are not allowed to mix or come into direct contact with each other. The reactor vessel 2 is surrounded by a containment vessel 4. The containment vessel 4 is placed in a pool of water 16. The pool of water 16 and the containment vessel 4 are below ground 9 in a reactor bay 7. The containment vessel 4 is designed so that water or steam from the reactor vessel 2 is not allowed to escape into the pool of water 16 or the surrounding environment. A steam valve 8 is provided to vent steam 11 from the reactor vessel 2 into an upper half 14 of the containment vessel 4. A submerged blowdown valve 18 is provided to release the water 10 into suppression pool 12 containing sub-cooled water. During a loss of feedwater flow, the nuclear reactor 5 is designed to respond by scramming the reactor core 6, flooding the containment vessel 4 or depressurizing the reactor vessel 2. The latter two of these responses result in the nuclear reactor 5 being shut down and unable to generate electricity for an extended period of time. The present invention addresses these and other problems. A power module assembly is herein disclosed as comprising a reactor vessel containing a reactor core surrounded by a primary coolant. A containment vessel is adapted to be submerged in a containment cooling pool and to prohibit a release of the primary coolant outside of the containment vessel. A secondary cooling system is configured to remove heat generated by the reactor core, wherein the heat is removed by circulating liquid from the containment cooling pool through the primary coolant. A cooling system for a nuclear reactor is herein disclosed as comprising an inlet line configured to deliver coolant to a heat exchanger, and an outlet line connected to the heat exchanger. The outlet line is configured to remove heat from the nuclear reactor. The cooling system further comprises an inlet port adapted to supply emergency feedwater to the inlet line, and an outlet port connected to the outlet line and configured to vent the emergency feedwater. The cooling system is configured to circulate the emergency feedwater through the heat exchanger by natural circulation. A method of cooling a nuclear reactor is herein disclosed. The method comprises detecting a loss of feedwater condition and replacing a feedwater flow from a secondary cooling system with an emergency feedwater supply. The emergency feedwater is circulated through a heat exchanger to remove heat from the nuclear reactor. The emergency feedwater is circulated through natural circulation. Conventional nuclear facilities are expensive to license and build, with significant upfront investment costs and delayed return of profits. In addition to energy cost considerations, efficiency requirements, and reliability concerns, today's nuclear reactor designs must also take into account issues of nuclear proliferation, terrorist activities, and a heightened awareness of environmental stewardship. Developing countries that could otherwise greatly benefit from nuclear power are frequently left to resort to other energy sources such as coal, gas or hydroelectric power generators that produce significant amounts of pollution or have other detrimental environmental impact. These developing countries may not have the technological or natural resources that enable them to build a nuclear power plant. Countries that have already developed nuclear power may be hesitant to introduce these technologies into the developing countries out of concern of the loss of control of the nuclear materials or technology. Passively safe nuclear power systems help address some of these concerns. Further system improvements and innovative designs are expected to usher in a new era of nuclear power as a globally viable primary energy source. In the Multi-Application Small Light Water Reactor (MASLWR) design, a loss of feedwater flow requires actuation of the long-term cooling mode of operation to provide for core cooling. For example, scramming the reactor core, flooding the containment vessel or depressurizing the reactor vessel. The latter two of these responses result in the nuclear reactor being shut down and unable to generate electricity for an extended period of time. Furthermore, no provision was provided in the MASLWR design for decay heat removal during a loss of site power. FIG. 2 illustrates a novel power module assembly 25 including a secondary cooling system 50. The power module assembly 25 includes an internally dry containment vessel 44. The containment vessel 44 is cylindrical in shape, and has spherical upper and lower ends. The entire power module assembly 25 may be submerged in a containment cooling pool 46 which serves as an ultimate heat sink. The containment vessel 44 may be welded or otherwise sealed to the environment, such that liquids and gas do not escape from, or enter, the power module assembly 25. The containment vessel 44 may be bottom supported, top supported or supported about its center. Supporting the containment vessel 44 at the top may facilitate maintenance and removal of the power module assembly 25 from the containment cooling pool 46. A reactor vessel 42 is located or mounted inside the containment vessel 44. An inner surface of the reactor vessel 42 may be exposed to a wet environment including a primary coolant 100 or liquid, such as water, and an outer surface may be exposed to a dry environment such as air. The reactor vessel 42 may be made of stainless steel or carbon steel, may include cladding, and may be supported within the containment vessel 44. The power module assembly 25 may be sized so that it can be transported on a rail car. For example, the containment vessel 44 may be constructed to be approximately 4.3 meters in diameter and 17.7 meters in height (length). By completely sealing the containment vessel 44, access to the reactor core 6 may be restricted. Any unauthorized access or tampering may be monitored. Furthermore, the subterranean profile of a nuclear power system makes it less visible and easier to conceal. The containment cooling pool 46 may be covered with a protective shield (not shown) to further isolate the power module assembly 25 from outside threats or airborne objects such as planes or missiles. The containment vessel 44 encapsulates and, in some conditions, cools the reactor core 6. It is relatively small, has a high strength and may be capable of withstanding six or seven times the pressure of conventional containment designs in part due to its smaller overall dimensions. Given a break in the primary cooling system of the power module assembly 25 no fission products are released into the environment. The primary coolant 100 remains entirely contained in the containment vessel 44. The reactor core 6 is illustrated as being submerged or immersed in a primary coolant 100, such as water. The reactor vessel 42 houses the primary coolant 100 and the reactor core 6. A shroud 22 surrounds the reactor core 6 about its sides and serves to direct the primary coolant 100 up through an annulus 23 and out a riser 24 located in the upper half of the reactor vessel 42 as a result of natural circulation of the primary coolant 100. In one embodiment, the reactor vessel 42 is approximately 2.7 meters in diameter and includes an overall height (length) of 13.7 meters. The reactor vessel 42 may include a predominately cylindrical shape with spherical upper and lower ends. The reactor vessel 42 is normally at operating pressure and temperature. In one embodiment, the containment vessel 44 is internally dry and may operate at atmospheric pressure with wall temperatures at or near the temperature of the containment cooling pool 46. During normal operation, thermal energy from the fission events in the reactor core 6 causes the primary coolant 100 to heat. As the primary coolant 100 heats up, it becomes less dense and tends to rise up through the riser 24. As the primary coolant 100 cools down, it becomes relatively denser than the heated coolant and is circulated around the outside of the annulus 23, down to the bottom of the reactor vessel 42 and up through the shroud 22 to once again be heated by the reactor core 6. This natural circulation causes the primary coolant 100 to cycle through the reactor core 6, transferring heat to a secondary cooling system 50 to generate electricity at a generator, such as generator 34 of FIG. 1. The secondary cooling system 50 comprises an inlet line 53 configured to deliver a secondary coolant to a heat exchanger 55 that operates as a heat sink for the power module assembly 25. An outlet line 52 is connected to the heat exchanger 55 and is configured to remove heat from the reactor core 6 by circulating the secondary coolant through the primary coolant contained in the reactor vessel 42. The inlet line 53 and outlet line 52 therefore serve as a means of delivery of the secondary coolant for the secondary cooling system 50. Relatively cool secondary coolant is transported to the heat exchanger 55 via the inlet line 53, whereas as relatively hot or superheated coolant is transported from the heat exchanger 55 to the generator via the outlet line 52. The secondary cooling system 50 further comprises one or more inlet ports 54 adapted to supply emergency feedwater to the inlet line 53. In one embodiment, the emergency feedwater supply is contained in the containment cooling pool 46. The containment cooling pool 46 may contain water or some other liquid coolant. One or more outlet valves 58 may be connected to the outlet line 52 and configured to vent the emergency feedwater after it has circulated through the heat exchanger 55. A check valve 56 may be provided between the inlet port 54 and the inlet line 53 to restrict the flow of emergency feedwater or secondary coolant in one direction. The inlet port 54 may include an inlet screen to filter contaminants in the containment cooling pool 46. The secondary cooling system 50 may be configured to circulate the emergency feedwater through the heat exchanger 55 by natural circulation. The natural circulation may be accomplished due to a temperature difference between the liquid in the containment cooling pool 46 and the primary coolant. The liquid, or emergency feedwater, undergoes a temperature change as it passes through the heat exchanger 55. The natural circulation may further be accomplished or augmented as a result of the temperature change of the emergency feedwater and an elevation difference of the inlet ports 54 and the outlet valves 58. In one embodiment, the one or more inlet ports 54 are located near the bottom of the containment cooling pool 46. As previously described, the inlet line 53 may be connected to a feedwater pump, such as pump 38 of FIG. 1, and the outlet line 52 may be connected to a steam turbine, such as turbine 32 of FIG. 1. The secondary cooling system 50 is able to provide natural circulation of the emergency feedwater without the assistance of a feedwater pump or external power source. During a loss of feedwater flow scenario, a reactor scram and turbine trip are initiated by an actuation signal. As steam pressure in the secondary cooling system 50 increases, the one or more outlet valves 58 are opened in a staged manner to depressurize the steam generator. The one or more outlet valves 58 may include redundant fail safe vent valves that discharge the secondary coolant to a set of submerged sparger nozzles located below the surface of the containment cooling pool 46. After a preset time delay from actuation of the outlet valves 58, a second set of valves associated with the one or more inlet ports 54 may be opened in such a way as to align the feedwater supply piping with the bottom of the containment cooling pool 46. As the steam generator depressurizes, a natural circulation flowpath is established from the inlet ports 54 through the heat exchanger 55 and out the outlet valves 58. Liquid from the containment cooling pool 46 provides a makeup source of emergency feedwater. Steam produced in the heat exchanger 55 and released through the outlet valves 58 is condensed in the containment cooling pool 46. The outlet valves 58 may be positioned just below the surface of the containment cooling pool 46. The secondary cooling system 50 may be configured to remove the liquid from the containment cooling pool 46 during one or more types of emergency operation. The emergency operation may include a loss of coolant accident, a loss of secondary coolant flow, a loss of on-site power, among others. FIG. 3 illustrates an embodiment of a novel cooling system 60. The cooling system may be configured to operate with a nuclear reactor such as the power module assembly 25 of FIG. 2. The cooling system 60 comprises inlet line 53 and outlet line 52 configured to remove heat from a reactor core by circulating a coolant through heat exchanger 55. Relatively cool coolant is transported to the heat exchanger 55 via the inlet line 52, whereas as relatively hot or superheated coolant is transported from the heat exchanger 55 to the generator via the outlet line 52. The cooling system 60 further comprises one or more inlet ports 54 adapted to supply emergency feedwater to the inlet line 53. In one embodiment, the emergency feedwater supply is contained in a cooling pool 46 (FIG. 2), and comprises water or some other coolant. A cooling pipe 57 connecting the inlet port 54 to the inlet line 53 may be insulated to reduce an amount of heating of the emergency feedwater traveling through the cooling pipe 57 that might otherwise occur due to a temperature difference of the coolant located at the top and bottom of the cooling pool. One or more outlet pots 58 may be connected to the outlet line 52 to vent the emergency feedwater after it has circulated through the heat exchanger 55. The emergency cooling system 60 further comprises one or more accumulator tanks 70 configured to inject coolant into the inlet line 53 when a loss of feedwater flow is detected. The one or more accumulator tanks 70 provide coolant to the heat exchanger 55 until natural circulation of the emergency feedwater is established via the inlet ports 54 and outlet ports 58. The one or more accumulator tanks 70 may be partially filled with water. The accumulator tanks 70 may be pressurized with a non-condensable gas, such as nitrogen. In one embodiment, a bladder 71 is provided with, or in, the accumulator tank 70 to prevent the release of the non-condensable gas (e.g. nitrogen) into the heat exchanger 55. During a loss of feedwater flow scenario, the accumulator tanks 70 inject the water into the inlet line 53. The injection of water serves to subcool the water in the reactor vessel while natural circulation is established in the cooling system. Operation Example operations of various embodiments is now provided, making reference to the secondary cooling system 50 of FIG. 2 and the emergency cooling system 60 of FIG. 3. The reactor core 6 undergoes a hot shutdown condition with control rods inserted. The shutdown condition may result from an emergency operation of the power module assembly 25 or the secondary cooling system. A normal steam flow through the outlet line 52 and a feedwater flow through the inlet line 53 are isolated. The inlet port 54 and outlet valve 58 are opened to the containment cooling pool 46. This creates a natural circulation flow path driven by the coolant density difference and elevation difference between the cold water at the inlet port 54 and the outlet valve 58. Cold water from the containment cooling pool 46 is drawn into the heat exchanger 55 where it is heated and vented into the containment cooling pool 46. The heat exchanger 55 removes heat from fluid, for example the coolant 100, in the annulus 23 creating a density difference between the fluid inside the riser 24 and the fluid in the annulus 23. Because the reactor core 6 is located at an elevation below the heat exchanger 55, a buoyancy force is created that drives warm fluid up through the shroud 22 and riser 24 and drives cold fluid down through the annulus 23 into the lower plenum 51. This creates a natural circulation flow through the reactor core 6 that removes the decay heat. The check valve 56 limits fluid flow in one direction. During a loss of main feedwater flow, a low discharge pressure of the Low Main Feedwater Pump 72 or a low water level of the Steam Generator 74 may result in a reactor trip. After a time delay from the reactor trip, the main steam vent valves (outlet valves 58) are opened in a staged manner. When a low level of the Accumulator tank 70 and a low discharge pressure of the Low Main Feedwater Pump 72 are detected, the Main Feedwater Stop Valve 76 is closed. Then, the inlet valve (inlet ports 54) is opened in order to align the cooling flow to the containment cooling pool 46. A check valve 66 prevents reverse flow from the Steam Generator 74 and a high point vent 68 provides for the initial purging of air in the emergency feedwater line. In various embodiments, the same operations as discussed for the above loss of main feedwater flow condition may be followed during a station blackout or loss of power to the on-site facility. During a loss of coolant accident, a low water level of the Low Pressurizer 73, a low pressure of the coolant system, or a high pressure of the containment vessel 44 may result in a reactor trip. The Main Steam Isolation Valve 78 and Main Feedwater Stop Valve 76 are closed. If the steam generator pressure 77 does not increase (e.g. there is no steam generator tube rupture) and the steam generator tube band pressures are equal, then the outlet valves 58 may be opened in a staged manner. For example, the Main Feedwater Stop Valve 76 is closed when a low level of the Accumulator tank 70 and a low discharge pressure of the Low Main Feedwater Pump 72 are detected. Additionally, the inlet port 54 may be opened to align the cooling flow to the containment cooling pool 46. The outlet valves 58 may also be opened in a staged manner. A reactor sump valve may be opened when a low differential pressure is detected between the containment vessel 44 and the reactor vessel 42. FIG. 4 illustrates a novel method of cooling a nuclear reactor. At operation 410, a loss of feedwater condition is detected. The loss of feedwater may be due to a loss of coolant accident, a loss of feedwater pressure, a failed feedwater pump, or a loss of on-site power, such as a station blackout. At operation 420 a feedwater flow from a secondary cooling system is replaced with an emergency feedwater supply. At operation 430 the emergency feedwater is circulated through a heat exchanger to remove heat from the nuclear reactor. The emergency feedwater is circulated through the heat exchanger through natural circulation. The natural circulation is due to a difference in temperature between the emergency feedwater supply and the emergency feedwater circulating through the heat exchanger. In one embodiment, the emergency feedwater supply comprises a containment cooling pool surrounding the nuclear reactor. At operation 440, the emergency feedwater is vented into the containment cooling pool. An elevation difference between an outlet port and an inlet port submerged in the containment cooling pool may provide for sustainable natural circulation of the emergency feedwater for more than three days. Depending on the size of the containment cooling pool, in one embodiment, the natural circulation may be maintained upwards of 90 days. Loss of feedwater flow and decay heat removal is resolved by the various embodiments disclosed herein. The novel systems add significant capability to the MASLWR reactor design by providing a passive means of cooling the nuclear core after a control rod insertion without the need for external power. Various embodiments are able to provide emergency feedwater to the steam generator in the event of a loss of normal feedwater flow, and provide reactor core decay heat removal of the reactor core subsequent to a reactor control rod insertion. The containment cooling pool serves as a source of makeup feedwater and as a heat sink for decay heat removal. Various embodiments are able to remove core decay heat by directing coolant from the large containment cooling pool through the helical coil heat exchanger tubes located in the reactor vessel annulus. A natural circulation flow path is established as hot water and steam are vented into the containment pool and cold water is drawn into the inlet port. Various embodiments disclosed herein provide alternate long-term cooling mode of operation and an indefinite heat removal of the reactor core without operator action. Very little mass is lost form the containment pool. Various embodiments can be actuated manually to remove decay heat for maintenance, and the systems are relatively simple, having few if any moving parts. Passive systems do not require on-site power to operate, instead relying on the principles of natural circulation. Furthermore, various embodiments provide for quick restart of a reactor module, providing less operating downtime and increased protection of investor capital. Although the embodiments provided herein have primarily described a pressurized water reactor, it should be apparent to one skilled in the art that the embodiments may be applied to other types of nuclear power systems as described or with some obvious modification. For example, the embodiments or variations thereof may also be made operable with a boiling water reactor. A boiling water reactor may require larger vessels to produce the same energy output. Having described and illustrated the principles of the invention in a preferred embodiment thereof, it should be apparent that the invention may be modified in arrangement and detail without departing from such principles. We claim all modifications and variation coming within the spirit and scope of the following claims.
	description	This application is a U.S. National Stage Application of PCT Application No. PCT/AT2005/000454, filed Nov. 15, 2005, which claims priority from Austrian Patent Application No. A 1927/2004, filed on Nov. 17, 2004. The invention relates to improvements of pattern-lock methods for use in a particle-beam processing apparatus. More in detail, the invention relates to a method of controlling the positioning of an image of a pattern in a particle-beam projection processing apparatus during exposure of a target with the image; the pattern is formed in a pattern defining means and is, by means of a beam of electrically charged particles, imaged through a particle-optical projecting system onto the target, in the following conceptual steps: imaging a number of reference marks provided in the pattern defining means through the same projecting system onto a corresponding number of registering means positioned at locations of desired positions of images of the reference marks, detecting images of the reference marks at the registering means generating corresponding signals in dependence of the fine position of the respective images, processing these signals in a control means and deriving control signals for adjusting the imaging properties of the projecting system and bringing the reference mark images into coincidence with the mentioned desired positions. Methods and devices for so-called pattern-locking, i.e. holding fixed the position of the imaged structure pattern on a target, such as a wafer substrate, in a particle projection system are well-known and are discussed in detail in U.S. Pat. No. 4,823,011 (=EP 0 250 391) and U.S. Pat. No. 4,967,088 (=EP 0 294 363) of the applicant (assignee). One important application of processing apparatus of this kind is in the field of particle-beam lithography used in semiconductor technology, as a lithography apparatus, wherein, in order to define a desired pattern on a substrate surface, the wafer is covered with a layer of a radiation-sensitive photoresist, a desired structure is imaged onto the photoresist by means of a lithography apparatus which is then patterned by partial removal according to the pattern defined by the previous exposure step and then used as a mask for further structuring processes such as etching. Another important application of processing apparatus of this kind is in the field of nano-scale patterning, by direct ion beam material modification or ion beam induced etching and/or deposition, used for the fabrication or functionalization of nano-scale devices, particularly having sub-100 nm feature sizes. In the process of forming a desired pattern on a wafer (or like target) it is a regular requirement to keep the position of the image accurately on the wafer during exposure; this task is commonly referred to as pattern-locking. Pattern-locking is done, e.g., with respect to the X and Y translation of the image in the wafer plane, rotation angle θ and magnification MX and MY with respect to the lateral scales, respectively, as well as trapezoid distortion ΔX and ΔY. During exposure, variations of these parameters may occur by voltage fluctuations (magnification M) or by slight movements of the target station with respect to the projection system, both in lateral directions X,Y and in Z direction, the latter causing also change in magnification M; another major source of imaging variations are external magnetic and/or electromagnetic fields. Since the position of the pattern image in the patterned beam cannot be measured directly (that would interfere with the image), additional reference beams, so-called ‘reference beamlets’, are employed for the purpose of alignment. The beamlets are defined in the device patterning the beam (mask or other pattern definition device), namely, by reference apertures positioned next to but separate from the pattern field, and are imaged towards the target plane in the same way as the beam portion which bears the pattern information of the image to be formed on the target. Thus, the beamlets surround a central beam portion which bears the pattern information and is also referred to as ‘dye beam’. In contrast to the dye beam, the reference beamlets do not reach the target (unlike the dye beam, they shall not be imaged onto the target), but they are registered in a corresponding number of detection units which serve as registering means measuring the positions of the individual beamlets, and the positions thus measured are compared with nominal positions relative to a stable reference plane in order to derive correction signals for the optical system. The detection units are contained in a reference apparatus, which is positioned in front of the target as seen in the direction of the beam (i.e., upstream). The reference apparatus together with the system to derive the correction signals from the measuring signals of the reference apparatus are referred to as pattern-lock system. (Sometimes in this term also those optical elements of the imaging system which are controlled by said correction signals are included.) In usual state-of-the-art setups, the reference apparatus comprises a scanner block and a reference plate, wherein the scanner block is positioned in front of the reference plate. The reference plate is provided with registration marks for detecting incident beamlets in cooperation with detectors housed in the scanner block, which also comprises deflector means for the beamlets. Each beamlet is associated with a respective deflector means through which it passes, a registration mark and a detector. Upon incidence with a registration mark, the beamlet produces secondary electrons which are detected by the respective detector. By means of the deflector means, each beamlet is scanned across the registration mark; the frequency of the scanning movement is typically in the 100 Hz range. From the time-dependent detector signal, taking into account the time-dependent deflection voltage signal, the position of the beamlet (that is, with respect to its non-deflected position) is derived. A state-of-the-art scanner block is a device surrounding an opening for the dye beam, for instance as a ring-shaped unit with a circular opening in order to affect the electrical fields guiding the beam as little as possible. On the other hand, the reference beamlets are situated outside of the pattern beam region, in order to pass through the detector means situated in the scanner block. The geometry of the reference beamlets is reflected in the layout of the pattern definition device, in which the reference marks are situated in a (usually ring-shaped) area surrounding the pattern field of the aperture pattern. The beam illuminates the pattern field as well as the reference marks, consequently it must be sufficiently defined (with respect to homogeneity and optical aberrations) over the total area including the reference mark area. This layout of the reference device brings about space problems, in particular in reducing optics since the image produced at the target is small and the final optical elements do not allow sufficient room for space-consuming devices such as a scanner block. Therefore, new ways are sought to resolve the space problems inferred from a pattern-lock device. This task is solved by means of the invention through several aspects. A first aspect of the invention is applicable in a particle-beam apparatus wherein the imaging of the pattern is done by means of at least two consecutive projector stages of the projecting system, namely, at least one non-final projector stage and one final projector stage, the latter producing the image of the pattern. In contrast to the pattern image, the detection of reference marks images is done at the location of an intermediary image of a non-final projector stage, the registering means being positioned at locations of nominal positions of an intermediary imaging plane. This solution allows to perform the pattern-lock at a place with considerably alleviated space restrictions; moreover, with reducing imaging systems, the intermediate image is at a lower degree of demagnification, so the detection devices can be formed with relieved miniaturiziation. According to a second aspect of the invention, the scanning movement of the reference beamlets is generated already at the place of the pattern definition device. Thus, reference beamlets formed by at least one of the reference marks are shifted laterally by means of deflector means provided in the pattern defining means in dependence of a time-dependent electric voltage applied to the deflector means, resulting in a corresponding time-dependent scanning movement of the respective reference mark image(s). In one preferred embodiment, the lateral shift of the reference beamlets may be generated by two (or more) sets of electrodes, which are positioned in series as seen along the direction of the beamlets, for instance electrode multipoles. Another aspect of the invention uses the exposure pauses between two exposure steps, such as the pausing for movement of the beam to another die field on a wafer. During exposure pauses a lateral movement of the reference mark images is applied by means of the projecting system—in particular, by means of a deflection means such as an electrostatic multipole means—and the resulting varying signals from the image detection are recorded, and based on these signal data a suitable operation point for the position of the reference mark image with regard to the respective registering means is selected. This procedure allows repeated recalibration of the reference system, improving the qualtiy of the pattern-lock control loop. In order to obtain an improved resolution of the relative position of a reference beam, a reference mark image may be detected by means of a registering means comprising a multitude of detective areas arranged beside each other and detection signals produced by the detective areas when hit by radiation from the reference mark are added to a sum signal, wherein the detective areas yield different contributions to the detection signal in relation to a unit of impingent radiation, respectively, resulting in a sum signal varying in dependence of the position of the reference mark image with respect to the detection means. In one suitable relaization of such a detection means, the detective areas are alternatively connected to the positive and negative input of a subtracting unit. Alternatively, the detective areas may be realized as portions which are alternatively elevated and lowered with regard to the direction of the radiation by a height difference which should be not smaller than the lateral width of the portions. For the explanations in this disclosure, a Cartesian coordinate system is used with Z denoting the axial coordinate along the optical axis and X, Y denoting the two perpendicular directions. The preferred embodiment of the invention discussed in the following is based on the pattern definition (PD) system disclosed in the U.S. Pat. No. 6,768,125 (=GB 2 389 454 A) of the applicant (assignee) whose teaching is herewith incorporated into this disclosure. In the following, the technical background of the PD system, as far as relevant to the invention, is first discussed with reference to FIGS. 1 to 4 (which correspond to those of the U.S. Pat. No. 6,768,125), then embodiments of the invention in the PD system are discussed. It should be appreciated that the invention is not restricted to the following embodiments nor the PD system, which merely represent one of the possible implementations of the invention; rather, the invention is suitable for other types of particle-beam exposure systems as well. Pattern Definition Device and PML2 System The PD system as described in the U.S. Pat. No. 6,768,125 realizes a multi-beam direct write concept, dubbed PML2 (short for “Projection Mask-Less Lithography #2”). It employs a PD device comprising a number of plates stacked on top of the other, among them an aperture array means (aperture plate) and a blanking means (blanking plate). These separate plates are mounted together at defined distances, for instance in a casing. The aperture array means has a plurality of apertures of identical shape defining the shape of beamlets permeating said apertures, wherein the apertures are arranged within a PD field composed of a plurality of staggered lines of apertures, wherein the apertures are spaced apart within said lines by a first integer multiple of the width of an aperture and are offset between neighboring lines by a fraction of said integer multiple width. The blanking means has a plurality of blanking openings arranged in an arrangement corresponding to the apertures of the aperture array means, in particular having corresponding staggered lines of blanking openings. The teaching of the U.S. Pat. No. 6,768,125 with regard to the architecture and operation of the PD device, and in particular the architecture of its blanking plate, are hereby included as part of this disclosure. As already mentioned, an electron beam generated by an illumination system is used in the PML2 system. It illuminates the PD means in order to define a beam pattern to be projected on a target surface. 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. The beam permeating the aperture array (or more exactly, through the switched-on apertures of the array) forms a patterned particle beam bearing a 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 continuously 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 scanning stage). An overview of a maskless particle-beam exposure apparatus PML2 employing the preferred embodiment of 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. The main components of the apparatus 100 are—corresponding to 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 substrate 41. The whole apparatus 100 is contained in a vacuum housing 105 held at high vacuum to ensure an unimpeded propagation of the beam lb, pb along the optical axis cx of the apparatus. The particle-optical systems 101, 103 are realized using electrostatic or electromagnetic lenses. The illumination system comprises, for instance, an electron gun 11, an extraction system 12 as well as a condenser lens system 13. It should, however, be noted that in place of electrons, in general, other electrically charged particles can be used as well. Apart from electrons (emitted from an electron gun) these 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. The ion source 11 emits energetic ions, i.e., having a defined (kinetic) energy of typically several keV, e.g. 10 keV. By means of an electro-optical condenser lens system 13, the ions emitted from the source 11 are formed into a wide, substantially telecentric ion beam serving as lithography beam lb. The lithography beam lb then irradiates a PD device 20 which, together with the devices needed to keep its position, form the PD system 102. The PD device 20 is held at a specific position in the path of the lithography beam lb, which thus irradiates a plurality of apertures 21 (see FIG. 2). Some of the apertures are “switched on” or “open” so as to be transparent to the incident beam; the other apertures are “switched off” or “closed”, i.e. non-transparent (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 (in FIG. 1, below the device 20). The pattern as represented by the patterned beam pb is then projected by means of an electro-optical projection system 103 onto the substrate 41 where it forms an image of the switched-on mask apertures 21. This is in contrast to, e.g., Arai et al., where rather than the apertures, images of the particle source are formed on the substrate. The projection system 103 implements a demagnification of, for instance, 200× with two crossovers c1,c2. The substrate 41 is, for instance, a silicon wafer covered with a photo-resist layer. The wafer 41 is held and positioned by a wafer stage 40 of the target station 104. In the embodiment of the invention shown in FIG. 1, the projection system 103 is composed of two consecutive electro-optical projector stages 31,32. The electrostatic lenses used to realize the projectors 31,32 are shown in FIG. 1 in symbolic form only as technical realizations of electrostatic imaging systems are well known in the prior art, such as, for instance, the U.S. Pat. No. 4,985,634 (=EP 0 344 646) of the applicant. The first projector stage 31 images the plane of the apertures of the device 20 into an intermediate plane el which in turn is imaged onto the substrate surface by means of the second projector stage 32. In the intermediate plane el the beam is substantially telecentric as it is at the position of the PD device. Both stages 31,32 employ a demagnifying imaging through crossovers c1,c2; thus, while the intermediate image in the plane e1 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 elevate problems of miniaturization in the PD device. After the first stage 31 the beam width is well reduced—for instance, from an initial width of the PD field pf (FIG. 2) of L=60 mm to about 4 mm at the intermediate plane e1. As a consequence, since the dimensions of the electro-optical components of the second stage 32 need not be reduced to the same scale as the beam width, the lens elements can be realized larger with respect to the beam, which allows for an easier treatment of lens defects and imaging aberrations. For example, with a total source-substrate length of about 2 m, the focal length of the final lens after the second stage crossover c2 can be as small as about 20 mm. This allows for high ion currents that can be treated, for instance of the order of 4 to 10 μA, because space charge correlation have only little influence. In both projector stages the respective lens system is well compensated with respect to chromatic and geometric aberrations; furthermore, a residual chromatic aberration of the first stage 31 can be compensated by suitable fine correction of the electrode potentials in the second stage 32. By virtue of the chromatic compensation, the energy of the ions (or in general, charged particles) emitted from the source 11 are allowed to have a comparatively high energy blur of up to ΔE=6 eV. This allows to use sources with less stringent requirements for quality and, therefore, emitting higher currents. Furthermore, the effect of stochastic errors, which are due to particle interactions mainly in the crossovers c1,c2, is reduced as the stochastic errors of the first stage are demagnified in the second stage, and stochastic errors in the second stage have little influence due to the small distance of the second crossover c2 from the substrate plane. As a means to shift the image laterally, i.e. along a direction perpendicular to the optical axis cx, deflection means 315,325 are provided in one or both of the projector stages. The deflection means can be realized as, for instance, a multipole electrode system which is either positioned near to the crossover, as shown in FIG. 1 with the first stage deflection means 315, or after the final lens of the respective projector, as is the case with the second stage deflection means 325 in FIG. 1. In this apparatus, a multipole electrode is used as deflection means both for shifting the image in relation to the stage motion and for correction of the imaging system in conjunction with the pattern-lock system explained in detail below; additionally, a magnetic coil 62 may be used to generate a rotation of the pattern in the substrate plane where needed. FIG. 2 shows a plan view of the arrangement of apertures in the PD device 20. A plurality of square-shaped apertures 21 is provided which are arranged within a PD field pf in a regular array in which the apertures 21 are aligned along adjacent lines pl, wherein in each of the lines pl the same number of apertures is present. The PD field pf may be segmented into stripes containing apertures, referred to as apertures fields af, and regions sf clear of apertures. In the aperture fields af, seen along the direction perpendicular to the lines pI, the apertures form a sequence of rows r1,r2,r3; in the embodiment shown, the rows r1-r3 are not adjacent but spaced apart. The apertures are arranged in the aperture fields af according to a skewed regular arrangement: The apertures of every third row align (n=3; in other embodiments, n may also assume other values >2) as the pitch pn between neighboring rows is three times the width w of an aperture (i.e., pn=n×w), and the offset pm between neighboring rows is 4 times the width of an aperture (i.e., pm=m×w with m=4; m may also assume other values >2). Within a line pl, the offset of apertures is n·pm=12. Thus, the apertures cover only 1/(n×m)= 1/12 of the area of the field af and, at a time, only one out of n×m=12 image elements can be exposed as shown in FIG. 3; the other elements are exposed in subsequent steps by means of moving the substrate along the “scanning direction” sd relative to the image of the apertures. For the spatial arrangement of the apertures 21 and the layout of circuitry to control the apertures it is referred to U.S. Pat. No. 6,768,125. FIG. 3 illustrates the image field mf produced on the substrate; for the sake of clarity it is assumed that all apertures are switched on in this figure. The width fw of the image field is the width L of the PD field pf reduced by the demagnification factor of the projection system. The image field is composed of a plurality of image elements mx (also referred to as pixels). For a given position of the image field on the substrate, each of the apertures 21 of the aperture array corresponds to an image element mx, but as the apertures only cover a fraction of the PD field area, only a corresponding fraction of the number of image elements (shown hatched in FIG. 3) can be exposed at a time. In order to expose also the other image elements, the substrate is moved under the beam so as to shift the image field on the substrate. FIG. 3a illustrates the exposure of pixels in subsequent positions of the motion of the substrate through the possible 12 (=n×m) positions; the pixels are accordingly referenced with letters a to 1 (the pixels shown hatched are position a). The whole image field mf is moved over the surface of the photoresist-covered wafer serving as substrate 41 so as to cover the total area of the substrate surface. As shown in FIG. 4, in order to minimize the path length of the movement, the scanning direction sd alternates from one scan line sI to the next (boustrophedonal motion). In the example discussed here, the minimum feature is 50 nm, for instance, and the smallest spot to be illuminated on the wafer, here defined as the pixel width x, is 25 nm. The image field width fw is 300 μm; in order to produce this image field in connection with a two-stage 200× demagnification projection system as mentioned, the square-shaped PD field has a width L=60 mm. It should be noted that the parameters n, m, w etc. used in the example illustrated in FIGS. 2 to 4 can also take other respective values instead of those described above. Beam Control The task of controlling the position of the dye beam on the target can be split into the several tasks which are attributed to different control systems according to the respective speed of control: 1. Wafer Alignment. Any target to be processed, in particular wafers containing structures from previous processing steps, have to be properly aligned with respect to the lithography beam to ensure proper image overlay. This procedure has to be done once for every wafer to be processed. For wafer alignment, methods are well-known from prior art and are not further discussed here. 2. Footprint calibration. The associated rate is well below 0.1 Hz, typically about 0.01 Hz. In order to compensate for long-term drifts of the system, the complete optical column is recalibrated at predetermined intervals. For this, the position of the lithography beam with respect to the stage system 40 and the beam footprint is measured. For footprint calibration, methods are well-known from prior art and not further discussed here. 3. Pattern-Lock. This takes place at a rate of the order of 1 kHz to be able to compensate for disturbances in the order of 100 Hz. The beam position is measured by means of a registering device at real-time and locked at the position of the registering device through a closed control loop. Pattern-locking serves to compensate short-term drifts in the particle-beam system and is discussed in detail below. 4. Beam Tracking. This happens at a rate of the order of 1 kHz. Deviations of the stage system 40 with regard to a nominal position are corrected by a beam tracking system. Beam tracking systems are well-known from prior art and are close related to write-on-the-fly techniques. 5. Toggle Correction. The pattern to be imaged onto the target is moved through the lines of apertures by a rate of typically 1 MHz or more. This discrete movement is adapted to the continuous movement of the target stage 40. This is suitably achieved by means of the reflections means 325 which supply a correcting shift (saw-tooth-like over time) to the image on the target. This disclosure focuses on the pattern-lock system for particle-beam apparatus such as the PML2 system. The pattern lock is required to compensate electromagnetic, in particular magnetic, disturbances in the regime around 100 Hz. Pattern Lock System A pattern-lock system is provided in order to stabilize the position of the image of the mask apertures (image field mf) on the substrate with respect to the ion-optical system. The principles of a pattern-lock system are described in U.S. Pat. No. 4,823,011 and U.S. Pat. No. 4,967,088. The entire contents of these two patent documents are hereby incorporated through reference. According to the pattern-lock concept, a number of reference beams rb are used which are formed in the PD system by reference marks 26 at the side of the PD field pf (FIG. 6). The reference beamlets propagate through the optical column together with the dye beam, thus experiencing substantially the same treatment (by the optical elements) and disturbances (by other sources) as the dye beam. A typical layout of the reference marks 26 is, for instance, a number of slots in a field of size 0.5 mm×0.5 mm, arranged at a distance of several mm from the PD field pf (FIG. 6a). A beam-tracking system can furthermore ensure compensation for image placement errors which are due to deviations in the stage movement, using a real-time feedback controller in conjunction with a multipole pattern correction. For instance, correction of image position and distortion can be done by means of a multipole electrode 325 (FIG. 1); additionally, a magnetic coil 62 can be used to generate a rotation of the pattern in the substrate plane. In FIG. 1 the position of a reference apparatus 60 according to the state of the art is shown, while the layout of a pattern-lock system according to the invention is discussed in the following with FIGS. 5ff; where not denoted otherwise in the following, the layout of the particle beam apparatus of FIG. 1 is not changed. FIG. 5 shows an overview of a particle-beam apparatus according to the invention with emphasis on the pertinent pattern-lock system with the paths of the reference beamlets rb. As already mentioned, the reference beamlets rb are defined by means of reference structures formed in the PD device 20 from the illuminating beam lb; after the PD device, the beamlets rb are imaged together with the patterned beam proper pb by the same optical elements; in particular, they are imaged through the first crossover c1 into an image of the reference structures at the intermediate plane e1. At that place, a registering device 601 surrounds the patterned beam pb and intercepts the reference beamlets rb. (In contrast, the patterned beam is further imaged through the second and final stage of the projection system onto the target.) A detail of the arrangement at the intermediate plane is shown in the detail of FIG. 5a. The registering device 601 comprises a number of detection units 61 (one for each beamlet rb) which serve as registering means for the respective reference beamlet, and are positioned at the place where the intermediate image is formed. This layout dispenses with the conventional requirement that the registering of the reference beams is done just in front of the target, where space problems are significant due to the strongly reduced size of the features and sensitivity of the image to influences on the imaging system. On the other hand, just because of the reduced size of the second stage, the optical system after the intermediary image i1 is easily shielded from external influences since it has a short length and the beam is comparatively little sensitive to external influences due to small diameter and high particle energy (e.g., 100 keV). On the other hand, before the intermediary image i1, the beam is difficult to shield from the surroundings and is ‘soft’ due to its comparatively large dimension and the low particle energy. PD System FIG. 6 shows a PD system 102 adapted for use with the invention, namely, in FIG. 6a a top view and in FIG. 6b a cross-sectional view. The PD system not only serves to define the pattern in the lithography beam, but also to form the reference beamlets rb. FIG. 7 shows the configuration of one single aperture of the PD system 102 in detail, and FIG. 8 shows the configuration of a reference mark in the PD system in a like detail. The system 102 comprises a number of plates 22 mounted in a stacked configuration. The PD system 102 according to the invention is realized as a composite device whose components serve respective functions. Each of the plates 22 is realized as a semiconductor (in particular silicon) wafer in which the structures were formed by microstructuring techniques known in the art. The lithography beam traverses the plates through an array of apertures in the PD field pf (FIG. 6). Each aperture corresponds to a set of openings 210,220,230,250 which are defined in the plates 22 (FIG. 7). In the following, the PD system 102 and its component plates 22 are discussed as far as necessary to disclose the invention; for more details the reader is referred to the U.S. Pat. No. 6,768,125. The thickness of each of the plates 22 is about 500 μm to 50 μm in the area of the apertures; their mutual distance is in the order of 10 μm to 1 mm. It should be noted that in FIGS. 6b and 7, the dimensions in the longitudinal axis (z-axis parallel to the optical axis of the apparatus) are enlarged and not to size. The plates 22 are cooled by thermal radiation from cooling elements 28, 29 (FIG. 1) provided behind and in front of the PD system 102, e.g. by means of cooled shield elements and/or using cooled electrodes of the first projector stage or the condenser lens. The blanking of the beamlets is controlled by means of a blanking means realized as a blanking plate 202 which comprises an array of openings 220 (“blanking openings”) of width w2, each corresponding to an aperture; the blanking openings, thus, form a blanking field bf in a manner analogous to the arrangement of apertures in the PD field pf. Each opening 220 comprises a set of beam blanking electrodes 221 as well as the circuitry 222 for controlling the electrodes 221. The width w2 of the openings 220 is larger than the dimension of the beamlet bm does not impinge on the electrodes 221. The PD device 102 further comprises an aperture array means which serves to define the beamlet laterally and which is here realized as an aperture array plate 203 (in the following in short ‘aperture plate’) with an array of openings having a width w3. In the embodiment shown the aperture plate is situated after the blanking plate 202 as the last of the plates 22 (as seen along the direction of the beam). It should be noted that it is the aperture 230 of width w3 which defines the lateral shape of the beamlet emerging from the system 102 (corresponding to the width w of an aperture in FIG. 2). The beamlet bm transgresses the subsequent openings of the plates 22 along the path p1 and is then imaged in the imaging system (FIG. 1), provided the blanking electrodes 221 are not energized; this corresponds to the “switched-on” state of the aperture (with respect to the switching state, no distinction is made between the blanking opening, the aperture associated with it respectively or the beamlet defined by that aperture). A “switched-off” aperture is realized by energizing the electrodes, applying a transverse voltage. In this state, the blanking electrodes 221 deflect the beamlet bm off the path p1 to a deviating path p0 so the beamlet will be absorbed, for instance at the stop plate 204 (FIG. 1) positioned at some place after the PD device. (It should be noted that it is sufficient in the switched-off state to deflect the beamlet bm by a small angle only, where the beamlet may still pass through the aperture 230; provided the deflection is sufficient to bring about a lateral deviation of the beam at a later position that is easy to block off the switched-off beamlet.) In front of the other plates (as seen in the direction of the lithography beam) a cover means realized as a cover plate 201 is provided in order to protect the other plates, in particular the blanking plate 202 and its delicate circuitry, from irradiation damage. The cover plate 201 takes up the majority of the impingent lithography beam lb; the particles can only pass through the openings 210, formed in an array corresponding to that of the blanking plate, which openings make up only a small fraction of the total area of the blanking field bf. As can be seen from FIG. 6b, the cover plate 201 and the aperture plate 203 are, respectively, the first and the last of the plates 22 of the PD device. This is a remarkable feature as it is only these two plates that come into contact with the radiation of the lithography beam. Therefore, only these two plates will have to be replaced at periodic intervals whereas for the other plates a provision of replacement is not necessary. (In a variant, the aperture plate 203 may be positioned as the first plate; it then fulfils the task of the cover plate. In this case only one plate comes in contact with the radiation of the lithography beam.) The plates 22 are held by chucks 23 which are positioned with respect to each other by means of actuators 24, 25 realized as piezoactuators or nanopositioning elements of known type. The vertical actuators 25 may also be left off in order to save space; then the positioning between the plates may be defined by the height of the chucks 23 themselves which then are simply stacked on each other. It is possible to omit the horizontal actuators provided the relative position of the plates are defined in such a way to ensure that the plates do not move more than a few 100 nm. One of the chucks, in FIG. 6 for instance the chuck of the aperture plate, may be formed as a cup 233 so as to facilitate lateral positioning of the other chucks. Preferably, the plates 22 and chucks 23 are produced from the same material, e.g. silicon, or materials having the same thermal expansion behavior in the operating temperature range. The chucks also provide for the electric supply of the blanking plate 202 and correction means 205 (if present); for the sake of clarity, the electric lines are not shown in the figures. In the plates 22 openings 26 are provided for the definition of reference beams rb as explained further below. The reference beams rb and the patterned beam pb are then imaged towards the substrate plane; in contrast to the patterned beam pb, however, the reference beams rb do not reach the substrate 41 but are measured in a pattern-lock system which is explained further below. The chucks 23 further have alignment openings 236 which serve as alignment markers for relative positioning of the chucks 23 and the plates 22 they hold. Furthermore, one or more additional adjustment units may be provided in the PD device. In the embodiment shown here, one adjustment unit 501 is positioned between the cover and blanking plates 201, 202; in other embodiments of the invention, any suitable combination of adjustment units could be implemented. The adjustment unit serves to (statically) control the path of the beamlets bm as they travel along the set of openings 210, 220, 230, 250 of the PD system 102, thus taking into account the effects of, and correcting for, possibly non-perfect matching of the components of the apparatus 100 with the PD device and/or the alignment of components (plates) of the PD device to each other, as well as dimensional deviations of the PD device components, in particular those that will occur due to fabrication and mounting tolerances. Also, some optical aberrations like the image distortion and field curvature can be reduced as well as the image distortion caused by the global space charge effect. Preferably, an adjustment unit 501 is positioned immediately after the first plate of the PD system, i.e., the cover plate 201, as shown in FIGS. 6 and 7. The adjustment unit 501 helps to reduce the requirements imposed on the illumination system 101 and makes the alignment of the PD plates easier. The adjustment unit 501 can be realized as explained in the following with reference to FIG. 7. An adjustment unit is composed of two deflector plates 50a, 50b, which each have conductor planes 51a, 51b comprising electrodes ea1,ea2,eb1 and the pertinent feeding lines (not shown) at the “bottom” side (the side facing towards the target). Static electric voltages can be applied to the electrodes of the adjustment unit which will correct the path geometry in a suitable way. The deflector plates 50a, 50b are aligned and fixed to each other by bonding or vacuum-compatible gluing. The electric contacts between the different conductor planes 51a,51b are made by, for instance, wire bonding. The deflector plates are provided with an array of openings matching the apertures of the PD system, but having a width w5 which is well greater than the width w1 of the beamlet as defined by the cover plate. Reference Mark and Beamlet Deflection FIG. 8 shows the configuration of a reference mark in the PD system of FIG. 6 in detail. In the first plate of the PD system, i.e. the cover plate 201, a number of parallel slots, in this example five slots, are formed. This set of slots functions as reference mark 26 which defines a reference beamlet rb consisting of a corresponding number of sub-beamlets. The reference beamlet then passes through openings 260 formed in the other plates 501,202,203. The openings 260 are sufficiently wide so as to ensure unimpeded passage of the reference beamlet. The reference beamlet thus formed is then imaged as explained above with reference to FIG. 6, forming an image of the reference mark 26 on the respective detection unit 61. A deflection unit 56 is provided in the PD system for each reference beamlet rb as means 56 to deflect the beamlet rb. Its purpose is to provide a—small but sufficient—angular deflection θr of the direction of the beamlet rb. The angular deflection Or serves to adjust the path of the reference beamlet so that it passes through the exposure device without being blocked by an aperture or other beam limiting device (e.g., stop plate 204) and correctly arrives at the respective detection unit. Such an angular deflection may be necessary due to higher-order aberration effects of the illumination system. In the example shown in FIG. 8, the deflection unit 56 is mounted on the adjustment unit 501. It is worthwhile to note that the adjustment unit may also have adjustment electrodes of the kind explained with FIG. 7 for each of the reference beamlets; then, the angular deflection θr produced by the electrodes 561 superimposes on the (static) deflection exerted by the adjustment electrodes. The longitudinal section of FIG. 9a and the cross-section of FIG. 9b illustrate a suitable realization of the deflection unit 56. At least one set of electrode multipoles 57 is mounted within a housing tube 562 surrounding the electrodes and containing the electric field produced by them. In the regime of small angles employed here, the angular deflection is a linear function of the strength of the electric field held in the space between the electrodes 57. Preferably, the set of deflecting electrodes may be divided into two segments 571,572 in axial direction, as illustrated in FIG. 9a. This is, for instance, suitable for applications where it is required to move the reference beamlet rb across the detection unit in a manner independent from the other reference beamlets and/or the dye beam. In particular, if the two multipole segments apply exactly opposite electric fields to the passing beamlet rb, the beam will be shifted by an amount proportional to the field strength along the direction of the electric field, resulting in a shift of the position of the image of the reference mark 26 on the detection unit, but no angular deflection is imparted to the beamlet. An additional angular deflection is applied if the two electric fields add to a net value different from zero. The two segments 571,572 are controlled individually, which allows to adjust the angular deflection θr and beamlet shift independently from each other. By means of the angular deflection θr the reference beamlet can be deflected, in particular in radial direction with respect to the optical axis. This ensures that the reference beamlets, despite being located at a considerable distance to the optical axis cx and thus being subject to possibly substantial off-axis aberrations of the optics, can be guided through the optical column and in particular the crossover c1 without a collision with any beam limiting devices (e.g., stop plate). Furthermore, by virtue of introducing a beamlet shift, the deflection units can further be used for fine correction of the hit position of the reference beamlets at the detection units. Consequentially, the hit position of the reference beamlets is adjusted by means of the deflection units in a way giving maximal signal response for positional deviations of the reference beamlets. Detection Unit As already mentioned, each of the reference marks 26 is imaged onto a respective detection unit 61 comprised in the registering device 601. FIGS. 10 to 14 show suitable realizations of a detection unit 61. In these examples, it is assumed that the reference beamlet consists of three sub-beamlets; this case can easily be extended to other reference beamlet configurations. Referring to the detection unit 610 of FIG. 10, six register pads 611,612,613,614,615,616 are provided which are alternatively connected with the positive and negative input of a differential amplifier 619. The pads 611-616 are formed as adjacent stripes of uniform, preferably oblong, shape and made of a electrically conductive material isolated from each other and against the surrounding. The material is absorptive to the electrically charged radiation. Consequentially, when a field is hit by the radiation of a reference beam, it will be electrically charged and the resulting electrical current is measured through the connected differential amplifier 619. Alternatively, the pads may be made of a material with a high coefficient for the production of secondary electrons, in which case the impingent radiation would cause a charging effect mainly due to the production of secondary electrons. In both cases, the electrical current is proportional to the amount of radiation on a pad. The register pads are arranged within a field corresponding to the dimensions of the reference mark 26; in the embodiment discussed here, they are formed as stripes within a field of about 36 μm side length. If six pads are realized within this field, each pad will be about 6 μm wide; if the number of pads is higher, the width will be reduced accordingly. FIG. 11 illustrates the register pads 611-616 being irradiated by the sub-beamlets 61s of the reference beam. In the case shown, the same amount of irradiation is imparted to the positive pads 611,613,615 and the negative pads 612,614,616; if the reference beam moved toward the right, then the positive pads would produce correspondingly more electric current while the current of the negative pads would likewise diminish, resulting in a distinct difference signal at the output of the differential amplifier 619. FIG. 12 shows a variant detection unit for a reference beam consisting of three sub-beamlets, realizing a groove detector 63. The detector 63 is realized as a piece of electrically conducting material which is isolated against the surrounding and connected to the input of an amplifier 639 which serves to measure the current against ground (surroundings). Several grooves 637, in this case shaped as three stripes so as to form a comb-like structure, are provided with the depth d6 of the grooves being greater than their width w6. Of course, the structure of the grooves corresponds with the structure of the reference beam (sub-beamlets). Referring to FIG. 13, irradiation by a reference beam of electron radiation causes emission of secondary electrons from the surface of the detector material. The secondary electron will radiate off into all directions. Within the groove, the secondary electrons are absorbed by the walls, like in a Faraday cup, and a resulting negative current will be measured corresponding to the impingent electron beam; those beam portions which hit the ridge regions, however, will cause secondary electrons that can leave the detector freely, so a net effect of electron lack will result, i.e., a positive current. Thus, the sign and value of the resulting current varies according to its position with respect to the groove-ridge boundary: negative if the sub-beamlets predominantly fall into the grooves and positive if the major portion hits the ridges. The same effect will be observed for a beam of negative charged ions. For positive charged ions, the resulting current will be positive always, but with the amplitude depending on the position. FIG. 14 illustrates a “multi-detection unit” 64 with a two-dimensional arrangement of register pads. Four sets of pads 64a,64b,64c,64d of the kind shown in FIG. 10 are positioned in parallel, but slightly offset. Thus, the reference beam will cause a different detector signal with each of the pad sets at the output of the individual amplifiers 649a-649d. FIG. 15 shows a linear variant 65 of a “multi-detection unit”. In this example, five pad sets are arranged in one line, each set consisting of two pairs of pads 651,652,653,654 connected to differential amplifiers 659. The offset of the pads of same sign is not quite the same as the offset of the sub-beamlets 65s, so the individual pad sets will be hit in different manners by the reference sub-beamlets resulting in different net output signals of the amplifiers 659. It is worthwhile to note that the configuration of the register pads need not be exactly corresponding to the configuration of the sub-beamlets of the reference beam; in fact it will be useful if the shape of the register pads suitably differs from the reference sub-beamlets as already exemplified in FIG. 15. Furthermore, referring to FIGS. 16 and 17 it may be advantageous if the structure of the pads and sub-beamlets is not uniformly periodic. FIG. 16 shows the profile 67 of a non-periodic structure of a groove detector (cf. the discussion of FIG. 12 above) wherein the width of the grooves varies: the width of selected grooves 677 is, for instance, doubled. In this example, the configuration of the sub-beamlets of the reference beam is likewise modified so as to correspond with the detector groove structure. Grace to the structure with suitably modified groove widths the intensity of the resulting detector current Ir will display a complex dependence upon the relative position Δx of the reference beamlet, as shown in FIG. 17. One distinct feature is the occurrence of the high-contrast central main maximum m1; another advantageous feature is the presence of secondary maxima m2 which come about through a proper choice of distance of structure features, in a symmetric arrangement, and allow to measure the width of the reference scanning movement. Pattern-Lock Control Loop Referring again to FIG. 5, the signals produced by the detection units, such as the detector currents Ir, are collected as input signals s1 and led to a pattern-lock control unit PCU for evaluation in order to obtain positional information for each reference beamlet. From the positional information the relative distortion state of the dye beam in terms of X/Y displacement, rotational displacement, magnification distortion, trapezoidal distortion, etc. can be determined. From these data, in turn, the control unit PCU calculates the amount of deviation from desired nominal values and determines corresponding correction signals s2. The signals s2 are fed back to the optical column by means of adjustment elements 315 of the optical column (e.g., electrostatic multipoles and/or magnetic elements as mentioned above) using a PID closed loop controller, in order to minimize the displacement/distortion signal and thus reducing the corresponding displacement/distortion state of the dye beam relative to some reference plane. The measurement of the detector signal is done with a measuring rate of, for instance, 1 kHz. This allows to determine and compensate mis-positioning influences which affect the reference beams (and consequently, the dye beam) within a frequency range below the measuring rate, i.e., in this case up to several 100 Hz. It should be noted that the patternlock will only keep a stable displacement/distortion state relative to some reference plane, while the displacement/distortion state control with respect to a the target (e.g., wafer) must be maintained by additional procedures/devices probably very similar to patternlock itself, usually termed as alignment. In order to perform the position detection for the reference beamlets, suitable voltage variations are applied to the electrodes of each deflection unit 56 such that the corresponding reference beamlets, and consequently the reference mark images, are scanned over the detection units thus giving a characteristic signal which in turn can be evaluated to give the desired beamlet position. Simplified Beamlet Deflection A simplified layout of a deflection unit is shown in FIGS. 18 and 19. In this case, a reference beamlet passes through a deflection unit 56′ which comprises a pair of deflection plates 561 only. The deflection plates are oriented in a manner that their main axis mx (the horizontal axis in FIG. 19b) intersects with the optical axis cx of the exposure apparatus thus allowing angular deflection of the beamlets for the purpose of guiding them through the column without a collision with any beam limiting devices. This angular deflection in combination with the present off axis aberrations further provide a possibility for fine adjustment of the position of the image of the reference beamlets formed in the reference plane. This simplified layout in combination with a recalibration scheme as discussed below with regard to FIG. 20 has the advantage that the deflection units 56′ are much simpler than those discussed above, and further that scanning of beamlets is not required any more. Moreover, if it is possible to illuminate the PD device 20 in a way that allows the reference beamlet to pass all apertures in the beamline without correction in the PD device, the deflection unit may be omitted. In this case the fine adjustment of the hit position of the reference beamlets at the detection units can be done by other means, e.g. suitable adjustment of the optical elements of the first projection stage and/or selection of a suitable portion of the detection unit in the registering device, such as one of the pad sets of the “multi-detection units” 64,65 shown in FIGS. 14 and 15. In this case, in order to perform the position detection for the reference beamlets, a scanning of beamlets should be avoided so the reference beamlets will nominally have a fixed position, thus giving a fixed signal level at the detector unit. Variations of the reference beamlets positions due to disturbances will result in variation of signal level on the corresponding detector, this signal level variations in turn are used for position determination of the reference beamlets. Recalibration Scheme FIG. 20 which shows an example of a detection unit signal Ir as a function over time t, illustrates a preferred method combining, in a cycle, recalibration of the reference beamlet position and control of pattern-lock. This method exploits the fact that in a manufacturing process of a processing apparatus 100, exposure of a target takes place during defined intervals of time Texp which are separated by periods where the target is not exposed; during these exposure pauses, the dye beam is usually switched off or otherwise blanked out. According to the method proposed, during an exposure pause recalibration of the reference beamlets is preformed. For this, the reference beamlets (together with the dye beam) are actively scanned over the detector range by means of the deflection units 315. This active scanning is performed in a time interval Tcal at the beginning of an exposure pause. FIG. 20 illustrates that during recalibration two extremal peaks are detected which mark the two ends of the operable regime, and towards the end of the calibration interval Tcal an operating point I0 is adopted in the center between these two extremals (in FIG. 20 only a simplified single ramp between both extremals is shown rather than the whole complex characteristic of a detection unit for clarity). After an operating point has been chosen and the calibration interval Tcal finished, patternlock control is resumed using the signal Ir for pattern-lock feedback as described. The patternlock control interval Treg fully embeds the time of one or more consecutive dye exposure steps Texp. During a dye exposure step, the reference beamlets hit the detectors without active scanning. By this measure, the reference beamlets are rigid with respect to the dye beam, and their position is used as indication of the position of the dye beam. This method allows to determine the complete signal behavior over the detector range and recalibrate the detector and amplifier, in particular with regard to the base position of the detector range. By this method, the base position can be adjusted to the center of the signal range which can then be used during the next exposure step(s). Suitably, an operating point I0 is chosen at which the slope of the signal is sufficiently high (for a high resolution) and allows sufficient space for control to either side. In order to get reliable signal levels from the detection units periodic recalibration of the individual detector response functions are necessary. By means of the recalibration any drifts in the detection units can be compensated. In a PML2 system this recalibration can advantageous be done in every writing pause necessary for the stage movement reversal action after every written image stripe. Although the invention is illustrated and described herein with reference to specific embodiments, the invention is not intended to be limited to the details shown. Rather, various modifications may be made in the details within the scope and range of equivalents of the claims and without departing from the invention.
	claims	1. A device configured for reconstructing axial measurement values in a nuclear fuel, the device calculating an axial measurement distribution by reconstructing a plurality of measurement values measured by a plurality of detectors that are disposed at predetermined intervals in the nuclear fuel along the axial direction of the nuclear fuel, the device comprising:a processor that includesa reconstruction parameter generator configured to generate a reconstruction parameter on the basis of core design data, or core analysis data, and a data adjustment factor; andan axial measurement distribution generator configured to calculate an axial measurement distribution in the nuclear fuel on the basis of the measurement values that are measured by the detectors and the reconstruction parameter that is generated by the reconstruction parameter generator,wherein the data adjustment factors are an inclination adjustment factor, an axial distribution adjustment factor, and an integral value adjustment factor. 2. The device for reconstructing axial measurement values in a nuclear fuel according to claim 1, whereinthe axial measurement distribution generator calculates the axial measurement distribution in the nuclear fuel by correcting the reconstruction parameter in accordance with the measurement values such that the deviation between the measurement values and the reconstruction parameter is at minimum. 3. The device for reconstructing axial measurement values in a nuclear fuel according to claim 1 whereinthe axial distribution adjustment factor is comprised of a plurality of adjustment factors for making adjustment with different periods. 4. A method of reconstructing axial measurement values in a nuclear fuel, the method being a method of calculating an axial measurement distribution by reconstructing a plurality of measurement values measured by a plurality of detectors that are disposed at predetermined intervals in the nuclear fuel along the axial direction of the nuclear fuel, the method comprising:a step of generating a reconstruction parameter on the basis of core design data, or core analysis data, and data adjustment factors; anda step of calculating an axial measurement distribution in the nuclear fuel on the basis of the measurement values that are measured by the detectors and the generated reconstruction parameter,wherein the reconstruction parameter is generated on the basis of an inclination adjustment factor, an axial distribution adjustment factor, and an integral value adjustment factor that serve as the data adjustment factors.
	claims	1. An imaging system for generating three dimensional image data using X-ray backscattering from a first side of a structure, said imaging system comprising:a first X-ray source and a second X-ray source;a first rotating collimator coupled to said first X-ray source and to a second rotating collimator coupled to said second X-ray source;a first shutter coupled to said first rotating collimator and a second shutter coupled to said second rotating collimator;an X-ray detector; anda controller coupled to said first and second X-ray sources, said first and second rotating collimators, and said X-ray detector, said controller is configured to:emit a first X-ray beam through said first rotating collimator at a first point in the structure through the first side of the structure while the second shutter is closed;detect backscattered X-rays from the first X-ray beam from the first side of the structure;emit a second X-ray beam through said second rotating collimator at the first point through the first side of the structure while said first shutter is closed; anddetect backscattered X-rays from the second X-ray beam from the first side of the structure; andgenerate three dimensional image data of the structure based on the detected backscattered X-rays from the first side of the structure. 2. The imaging system of claim 1, further configured to:emit the first X-ray beam at a first point in the structure through the first side of the structure from a first angle;detect first backscattered X-rays from the first X-ray beam from the first side of the structure;generate first image data from the first backscattered X-rays;emit the second X-ray beam at the first point in the structure through the first side of the structure from a second angle;detect second backscattered X-rays from the second X-ray beam from the first side of the structure;generate second image data from the second backscattered X-rays; andcombine the first image data with the second image data. 3. The imaging system of claim 1, further comprising a platform that is coupled to the structure, wherein said imaging system is further configured to rotate relative to the first side of the structure while emitting the X-rays and detecting the backscattered X-rays. 4. The imaging system of claim 1, further configured to generate parallel slices of image data corresponding to a plurality of depths within the structure. 5. The imaging system of claim 1, further configured to combine image data from two angles along a plane that is perpendicular to the structure. 6. The imaging system of claim 1, further configured to generate first image data for a plurality of first points at a first depth, then generate second image data for a plurality of second points at a second depth in the structure. 7. The imaging system of claim 1, further configured to generate first image data at a first plurality of depths along a first plane within the structure, then generate second image data at a second plurality of depths along a second plane that is adjacent to the first plane. 8. A method for generating three dimensional image data using X-ray backscattering from a first side of a structure, said method comprising:emitting a first X-ray beam through a first rotating collimator at a first point in the structure through the first side of the structure while a first shutter of the first rotating collimator is open and a second shutter of a second rotating collimator is closed;detecting backscattered X-rays from the first X-ray beam from the first side of the structure;emitting a second X-ray beam through the second rotating collimator at the first point through the first side of the structure while the second shutter of the second rotating collimator is open and the first shutter of the first rotating collimator is closed; anddetecting backscattered X-rays from the second X-ray beam from the first side of the structure; andgenerating three dimensional image data of the structure based on the detected backscattered X-rays from the first side of the structure. 9. The method of claim 8, further comprising:emitting the first X-ray beam at a first point in the structure through the first side of the structure from a first angle;detecting first backscattered X-rays from the first X-ray beam from the first side of the structure;generating first image data from the first backscattered X-rays;emitting the second X-ray beam at the first point in the structure through the first side of the structure from a second angle;detecting second backscattered X-rays from the second X-ray beam from the first side of the structure;generating second image data from the second backscattered X-rays; andcombining the first image data with the second image data. 10. The method of claim 8, wherein the first and second rotating collimators, the first and second X-ray sources, and the X-ray detector are included in an imaging system, said method further comprising rotating the imaging system relative to the first side of the structure while emitting the X-rays and detecting the backscattered X-rays. 11. The method of claim 8, further comprising generating parallel slices of image data corresponding to a plurality of depths within the structure. 12. The method of claim 8, further comprising combining image data from two angles along a plane that is perpendicular to the structure. 13. The method of claim 8, further comprising:generating first image data for a plurality of first points at a first depth; andsubsequently generating second image data for a plurality of second points at a second depth in the structure. 14. The method of claim 8, further comprising:generating first image data at a first plurality of depths along a first plane defined within the structure; andsubsequently generating second image data at a second plurality of depths along a second plane defined within the structure that is adjacent to the first plane.
052672913	description	DESCRIPTION OF THE PREFERRED EMBODIMENT Referring to FIG. 1, a perspective of a nuclear reactor core of a boiling water nuclear reactor is illustrated at four adjacent fuel bundles B1-B4. The four fuel bundles B1-B4 rest on fuel support S at the lower end and extend upwardly to a top guide G. Sealed fuel rods R extend from lower tie plate L above core support S to upper tie plate U. As can be seen lower tie plate L, fuel rods R, and upper tie plate U are all surrounded by channels C. In the case of fuel bundle B3, channel C and rods R are shown broken away without the upper tie plate U. Sufficient breaking away has occurred to expose one spacer S. Flow within the fuel bundles B1-B4 is easy to understand. Water flows into the fuel bundles B1-B4 from below core plate P at support S where the flow is distributed by fuel support S to equal flow among fuel bundles B1-B4. Flow through the lower tie plate of water occurs. The water passes upwardly confined by channels C and finally out of up tie plate U. During this flow the channels C separate flow interior of the fuel bundle from core bypass region R between the fuel bundles. It will be noted that fuel bundles B1-B4 are square in section and held in spaced apart relation with respect to one another. This being the case, it will be understood that the respective fuel bundles define a cruciform interstices of the core bypass region R. In the sketch here, a control blade N is shown partially inserted for control of the reaction. During normal operation, the control blade N is withdrawn, core bypass region R flooded, and reaction continues aided by the moderating effect of water moderator in region R. Flow to core bypass region R is well known and will not be specifically illustrated here. It will suffice to say that certain metering apertures at the bottom of fuel bundles B1-B4 and other reactor leakage paths maintain this region flooded with liquid moderator. Sufficient flow is provided to the region to prevent flashing of the contained water moderator to steam. Referring to FIG. 2, bundle B1 is illustrated in section adjacent spacer S. The fuel bundle comprises a 9 by 9 matrix of fuel rods R with a central water rod W for supplying a high moderator fraction to the central portion of fuel bundle B1. It will be understood that there is a flow region for the generation of steam interior of the fuel bundles while the surrounding region R is flooded with water. Attention can be devoted to the construction of spacer S. The illustrated spacer S is a so-called ferrule type spacer S have a matrix of interconnected ferrules F each surrounding a rod R at the elevation of the spacer. The spacers as a group are surrounded by band 14. FIG. 2 is a cross-section that illustrates the prior art by showing over emphasized the difficulties of the prior art. Specifically, spacer S has moved to the left opening a large gap 18 with respect to one wall of channel C at small gap 20 and closing small gap 20 with respect to the remaining channel wall. As has heretofore been stated, this migration of spacer S as a group causes rods R adjacent channel C to be even more proximate moderator within region R (and become more reactive and generate more heat) while at the same time restricting the flow of moderator within the fuel bundle B1 adjacent the inside wall of channel C at small gap 20. This will cause the critical power of the peripheral fuel rods R at this location to decrease. Once the critical power decreases at one bundle location, the remainder of the fuel bundle B1 is required to be restricted in heat out put so that the critical power limit at a fuel rod R is no where exceeded within a fuel bundle. Referring to FIGS. 3, 4, and 5, modification to band 14 is illustrated to prevent the prior art condition of FIG. 2. Band 14 (of a spacer S--the rest of the spacer being omitted for clarity) is shown with respective near sides 31, 32 and far sides 33, 34. Referring to far sides 33, 34, it will be seen that side 33 is formed with protrusions 45, 46 while side 34 includes formed protrusions 47, 48. Referring to the detail of FIG. 4, it can be seen that these respective protrusions are "bath tub" like indentations impressed to the metal of band B and relatively thick--especially when compared to protrusions 41-44 described hereafter. It is the purpose of the four protrusions 45-48 to occupy a full centering interval between a channel C and a spacer S at band B (See FIG. 6). Referring to near sides 31, 32, it will be seen that side 31 is formed with protrusions 41, 42 while side 32 includes formed protrusions 43, 44. Referring to the detail of FIG. 5, it can be seen that these respective protrusions are again "bath tub" like indentations impressed to the metal of band B and relative thin--especially when compared to protrusions 45-48. It is the purpose of the four protrusions 41-44 to occupy less than a full centering interval between a channel C and a spacer S at band B (See FIG. 6). This interval defined by the protrusions 41-44 with respect to channel C is chosen with two purposes in mind. First, clearance is required to fit the channel C over the remainder of the assembled fuel bundle B during assembly. Accordingly, protrusions 41-44 are less than protrusions 45-48 by a sufficient interval to define this assembly clearance. Second, a minimum interval is defined from the fuel rods R adjacent channel C to define a "worst case" critical power limit. Accordingly, protrusions 41-44 are dimensioned to provide this "worst case" critical power dimension. Finally, it will be seen that near sides 31, 32 include leaf springs 61-64. It will suffice to discuss leaf spring 64 as illustrated in FIG. 5. It will be understood that springs 61-63 are identically constructed. Referring to FIG. 5, leaf spring 64 is attached by some suitable means to band 14 at 65. Leaf spring 64 has an essentially U-shaped configuration imparted by bends at 65, 66, 67 and 68. A large raised central portion 70 resiliently bears against a channel C while the distal end 69 of spring 64 from attached portion 65 fits in sliding relation over the exterior of band 14. Preferably, springs 61-64 are located adjacent the corners of channels C. In this location, they bear on channel C where the channels have maximum strength. Function of springs 61-64 is plain to see in FIG. 6. Simply stated, springs 61-64 bias spacer S at band 14 on sides 31, 32 away from channel C. This causes registration of protrusions 45-48 to channel C with protrusions 41-44 being spaced apart from channel C. Centering of the confined fuel rods R within channel C occurs at the optimum interval established by protrusions 45-48 and the dimension of spacer S with respect to channel C. We prefer the leaf springs 61-64 vertically disposed. Similarly, protrusions formed from "bath tub" like projections in the metal are shown. The reader will understand that protrusions and springs can include a wide variety of equivalents.
	description	This application is the U.S. National Phase of PCT/US2015/011752, filed Jan. 16, 2015, which claims the benefit of priority from U.S. Provisional Application No. 61/928,301, filed Jan. 16, 2014, titled “System and Method For Improving Sensitivity Of A Sipping System,” the entire contents of which are hereby incorporated by reference herein. The present invention relates generally to vacuum canister sipping systems that are designed to detect failures in nuclear fuel assemblies. Vacuum canister sipping detects fuel failures in nuclear fuel assemblies by isolating a fuel assembly in a chamber/canister and reducing the ambient pressure in the chamber to draw out fission gas through the cladding failures in the fuel elements of the assembly. The gas bubbles are collected at the top of the chamber and passed through a radiation detector (e.g., scintillation detector) to detect the presence of fission gas. Assemblies that give off higher than expected fission gas activity are identified as failed and leaking in this manner. Known vacuum canister sipping systems are described, for example, in Westinghouse Product Fly Sheet, “Vacuum Canister Sipping Services,” NS-FS-0052 (75370) April 2009. U.S. Pat. Nos. 5,457,720 and 5,546,435 disclose other known methods for detecting failures in fuel assemblies. A limitation of existing vacuum sipping systems is that the application of vacuum during the sipping cycle causes dissolved gasses, radioactive or otherwise, in the canister water to come out of solution and contribute to the radioactivity of the gas sample being detected. It can therefore be challenging to detect faint fuel leak signatures under pool water conditions of high and/or fluctuating ambient dissolved fission gas content. One or more embodiments provide a method and/or apparatus that extracts dissolved fission gas from the water used in the sipping canister prior to the application of vacuum to the fuel, thus improving the detection sensitivity of the system. The extraction of fission gasses from the canister water prior to the start of the detection cycle can be performed for the purpose of reducing background fission gas concentration and improving the signal to noise ratio of the detection system. Alternatively, extraction of the fission gasses can be performed for the purpose of regularly measuring the fission gas concentration in the water, thus ensuring a reliable standard of “background” for each sipping cycle. The background gas extraction may be accomplished so as not to cause fission gas escape from a leaking fuel rod within an assembly being tested. The subsequent application of vacuum during the conventional sipping cycle steps draws out fission gas through fuel leaks and is detectable as a separate indication from that of the background fission gas. One or more embodiments provide a method for detecting a failure in a nuclear fuel assembly in a fuel pool, the method including: placing the nuclear fuel assembly into a canister of a canister vacuum sipping device disposed in the fuel pool; removing fission gases from fuel pool water to create degassed water; and after said placing of the nuclear fuel assembly into the canister, undertaking a vacuum sipping procedure using the canister sipping device to detect a failure in the fuel assembly by detecting fission gas leaking from the fuel assembly while the degassed water is disposed in the canister. One or more of these embodiments may further include flushing the degassed water through the canister after said placing of the nuclear fuel assembly into the canister and before said undertaking, thereby displacing at least a portion of fuel pool water disposed within the canister with the degassed water, wherein said flushing reduces a concentration of fission gases in water disposed within the canister. According to one or more of these embodiments, during the flushing, a volume of the degassed water being flushed into the canister is at least twice a volume of the canister. According to one or more of these methods, the removing includes filtering the fuel pool water through a gas transfer membrane device that extracts fission gases out of the fuel pool water. According to one or more of these embodiments, the removing includes applying vacuum pressure to a strip gas being used with the gas transfer membrane device such that the gas transfer membrane device filters both fission gas and dissolved air out of the fuel pool water. According to one or more of these embodiments, the removing includes introducing sparge gas into the canister such that dissolved fission gasses in fuel pool water disposed within the canister diffuse into the introduced sparge gas, and removing at least a portion of the introduced sparge gas from the canister. According to one or more of these embodiments, the removing of the fission gases from fuel pool water may occur after the placing of the nuclear fuel assembly into the canister and before the undertaking of the vacuum sipping procedure. One or more of these methods may also include maintaining a pressure within the canister at local ambient fuel pool pressure conditions during the introducing of the sparge gas into the canister. According to one or more of these embodiments, a portion of the sparge gas that was introduced into the canister is collected from the canister and its radioactivity measured to define a background level for comparison against the radioactivity measured during the vacuum sipping procedure. According to one or more of these embodiments, a supply of sparge gas used during the introducing has a lower fraction of fission gas than a dissolved gas mixture in fuel pool water disposed in the canister prior to the introducing. One or more embodiments provide a canister sipping system that includes: a canister vacuum sipping device including a canister, the canister vacuum sipping device being configured to detect a failure in a nuclear fuel assembly disposed in the canister by detecting fission gas leaking from the fuel assembly during a vacuum sipping procedure of the canister vacuum sipping device; and a water treatment device connected to the canister and configured to remove fission gases from fuel pool water to create degassed water, wherein the water treatment device is configured to provide the degassed water within the canister prior to undertaking of the vacuum sipping procedure. According to one or more of these embodiments, the water treatment device includes: a gas transfer membrane device that is configured to extract fission gases out of fuel pool water that has been extracted from a fuel pool to form the degassed water, and a canister flushing mechanism that is configured to flush the degassed water through the canister after the nuclear fuel assembly is placed into the canister, thereby displacing at least a portion of the fuel pool water disposed within the canister with the degassed water. According to one or more of these embodiments, the canister flushing mechanism is configured such that a volume of the degassed water being flushed into the canister is at least twice a volume of the canister. According to one or more of these embodiments, the water treatment device includes: a sparge gas supply and inlet connected to the canister, the water treatment device being configured to introduce sparge gas into the canister via the inlet such that dissolved fission gasses in fuel pool water disposed within the canister diffuse into the introduced sparge gas, and a sparge gas outlet connected to the canister, the outlet being configured to allow the introduced sparge gas to exit the canister, wherein the water treatment device is configured to remove the fission gases from fuel pool water after the nuclear fuel assembly is placed into the canister and before the undertaking of the vacuum sipping procedure. According to one or more of these embodiments, the system is configured to maintain pressure within the canister at a local ambient fuel pool pressure during the introducing of the sparge gas into the canister. According to one or more of these embodiments, the canister vacuum sipping device is configured to collect a portion of the sparge gas that was introduced into the canister, and measure a radioactivity of collected sparge gas to define a background level for comparison against a radioactivity measured during the vacuum sipping procedure. According to one or more of these embodiments, the sparge gas supply has a lower fraction of fission gas than a dissolved gas mixture in fuel pool water disposed in the canister prior to operation of the water treatment device. One or more embodiments provide a method for detecting a failure in a nuclear fuel assembly in a fuel pool, the method including: placing the nuclear fuel assembly into a canister of a canister vacuum sipping device disposed in the fuel pool; recirculating a carrier gas through fuel pool water disposed in the canister, thereby causing fission gasses in the fuel pool water in the canister to diffuse into the carrier gas being recirculated; and undertaking a vacuum sipping procedure using the canister sipping device to detect a failure in the fuel assembly by detecting fission gas leaking from the fuel assembly. According to one or more of these embodiments, the method also includes, before the undertaking, detecting with a radiation detector of the canister vacuum sipping device, a radiation level in the carrier gas to define a baseline radiation level, wherein the using of the canister sipping device to detect a failure in the fuel assembly includes comparing the baseline radiation level to a radiation level detected during the vacuum sipping procedure. According to one or more of these embodiments, the recirculating continues until a rate of increase in radiation level in the recirculating carrier gas falls below a predetermined threshold, and the undertaking occurs after the rate of increase in radiation level in the recirculating gas falls below the predetermined threshold. One or more embodiments provide a canister sipping system that includes: a canister vacuum sipping device including a canister and a radiation detector, the canister vacuum sipping device being configured to detect a failure in a nuclear fuel assembly disposed in the canister by detecting fission gas leaking from the fuel assembly during a vacuum sipping procedure of the canister vacuum sipping device; a gas recirculation device configured to recirculate a carrier gas through a loop that includes the canister, a pump, and the detector of the canister vacuum sipping device, wherein the pump is configured to recirculate gas through the loop; and a carrier gas supply connected to the loop and configured to supply carrier gas to the loop. These and other aspects of various embodiments of the present invention, as well as the methods of operation and functions of the related elements of structure and the combination of parts and economies of manufacture, will become more apparent upon consideration of the following description and the appended claims with reference to the accompanying drawings, all of which form a part of this specification, wherein like reference numerals designate corresponding parts in the various figures. In one embodiment of the invention, the structural components illustrated herein are drawn to scale. It is to be expressly understood, however, that the drawings are for the purpose of illustration and description only and are not intended as a definition of the limits of the invention. In addition, it should be appreciated that structural features shown or described in any one embodiment herein can be used in other embodiments as well. As used in the specification and in the claims, the singular form of “a”, “an”, and “the” include plural referents unless the context clearly dictates otherwise. All closed-ended (e.g., between A and B) and open-ended (greater than C) ranges of values disclosed herein explicitly include all ranges that fall within or nest within such ranges. For example, a disclosed range of 1-10 is understood as also disclosing, among other ranged, 2-10, 1-9, 3-9, etc. Canister sipping equipment can detect a leaking nuclear fuel assembly only when the level of activity released by the assembly is noticeably higher than the variation in background radiation. There are multiple sources of background noise that can degrade the minimum detection threshold of the equipment (e.g. presence of dissolved fission gas in the fuel pool and sipping canister, external gamma radiation to the radiation detector, contamination of the radiation detector, etc.). For example, during vacuum canister sipping procedures, both the fuel assembly being tested and the vacuum canister itself are disposed in a fuel pool at the nuclear plant. The water in the fuel pool typically contains some fission gases (e.g., from failed/leaking fuel assemblies in the pool). When such fission-gas-containing fuel pool water is present in the vacuum canister during the sipping process, the ambient fission gas reduces the sensitivity of the system to detecting fission gas that is leaking from the tested fuel assembly, as opposed to fission gas that is present in the ambient fuel pool environment. The ambient fission gas in the fuel pool may therefore impair the accuracy of failure detection or the ability to detect small failures in the fuel assembly being tested. For some types of reactors (e.g., PWR reactors), the fuel pool water is typically borated, while in other types of reactors (e.g., BWR reactors), the fuel pool water is not borated. Various embodiments of the present invention are applicable to reactors that use borated and/or unborated water in the fuel pool. As used herein, the term “fuel pool” and its derivatives includes similar elements of a nuclear plant where equipment can be installed and that are flooded with water to enable fuel movement. For example, the refueling cavity, transfer canal, cask loading pit, etc. are all considered to be a “fuel pool” as that term is used herein. Accordingly, the term “fuel pool water” includes the water disposed in any such “fuel pool.” To increase failure detection sensitivity, the measurement noise sources may be reduced. According to one or more embodiments, dissolved fission gas is filtered out of the pool water that is used in the canister to reduce this significant source of measurement noise. Reduction of dissolved fission gas in the canister, which is especially high after processing a leaking assembly, can be accomplished by pumping the water out of the canister so that it is replaced by water from the pool. This flushing reduces the amount of dissolved fission gas in the chamber back to the normal background of dissolved fission gasses in the bulk pool inventory. However, it is possible for the level of dissolved fission gasses in the bulk pool to be very high (e.g. if there are many leaking assemblies during an outage), which results in elevated background levels even after flushing with water from the pool. FIG. 1 illustrates a vacuum canister sipping system 10 according to an embodiment of the present invention. The system 10 includes a water treatment device 20 connected to a sipping device 30. As shown in FIG. 1, the water treatment device 20 includes a recirculating pump 50, one or more gas transfer membrane modules 70 (see also FIG. 2), a vacuum pump 80, a control system 90, an inlet water passageway 100 that leads to an inlet of the pump 50, an intermediate water passageway 110 connecting an outlet of the pump 50 to an aqueous stream inlet 70a of the module 70, a water outlet passageway 120 connecting an aqueous stream outlet 70b of the module 70 to an inlet of the sipping device 30, a strip gas inlet passage or port 70c connecting to a strip gas inlet of the module 70, an intermediate gas passageway 130 connecting a strip gas outlet 70d of the module 70 to an inlet of the vacuum pump 80, and a strip gas outlet passageway 140 connected to an outlet of the pump 80. The discharged air (including the fission gasses) from the vacuum pump 80 is directed either to a plant ventilation return or other location. For example, the plant may wish to direct the fission-gas-laden strip gas from the passageway 140 to a waste gas treatment system. According to various embodiments, the relative positions of components of the water treatment device 20 may be switched without deviating from the scope of the present invention. For example, the pump 50 may be disposed anywhere along the passageways 100, 110, 120 and/or be disposed downstream from the module 70. The pump 50 could alternatively be dispose downstream from the canister 500, for example downstream from the valve 520, and suction water through the system 10. Similarly, the pump 80 could alternatively be disposed upstream from the module 70 and pump strip gas to the inlet 70c. As shown in FIG. 1, many or all of the components of the water treatment device 20 may be mounted to a common frame 85 such as a skid or palette that provides for easy movement of the device 20. To use the device 20, the device and frame 85 may be positioned on a floor 95 of the nuclear power plant next to the fuel pool 160 to facilitate easy connection of the passageway 100 with the fuel pool 160 and connection of the passageway 120 with the sipping device 30. According to various alternative embodiments, the device 20 may alternatively be disposed on multiple frames without deviating from the scope of the present invention. Additional parts of the system 10 such as components of the sipping device 30 (e.g., the pump 570, detector 580, the control system 590) may also be mounted to the frame 85 to provide an easily accessible, compact system 10. As shown in FIG. 1, during operation of the water treatment device 20, the pump 50 moves ambient water from the fuel pool 160 sequentially through the passageway 100, the pump 50, the passageway 110, the module 70, and the passageway 120 to the sipping device 30. At the same time, the pump 80 sucks strip gas (e.g., ambient/local air around the fuel pool or air from another source) through the inlet 70c, the module 70, the passageway 130, the pump 80, and the outlet passageway 140. Turning to the operation of the module 70, as shown in FIG. 2, the module 70 comprises an enclosed housing 200 through which the water flows from the aqueous inlet 70a to the aqueous outlet 70b. A distribution tube 210 distributes the water flow outwardly from the inlet 70a. The water then passes around a central baffle 220, and back to a collection tube 230 that leads to the outlet 70b. The strip gas flows from the strip gas inlet 70c through a plurality of hollow fiber membranes 250 and out through the outlet 70d. As pool water flows around the membranes 250 inside the module 70 and strip gas passes through the membranes 250, the membranes 250 allow exchange of gases between the water and strip gas, resulting in a discharge dissolved gas composition in the water that closely matches the composition of the incoming strip gas (i.e., nearly free of fission gas). As a result, fission gas in the incoming pool water is filtered out of the water and into the strip gas stream. Consequently, the module 70 removes fission gas from the incoming pool water from the passageway 110 and discharges clean/degassed water with less fission gas into the passageway 120. The removed fission gas combines with the strip gas in the module 70 and is expelled through the outlet 70d. According to various embodiments, the module 70 and membranes 250 remove fission gases, but do not remove boron from borated fuel pool water (if the fuel pool water is borated according to various embodiments). Accordingly, the clean/degassed water being used to flush the canister 500 remains borated (in embodiments in which the fuel pool water was already borated), and does not impact the boron concentration in the fuel pool. In embodiments in which the fuel pool water is borated, if fresh clean water, rather than borated pool water, were used to flush a canister 500, the water would either be borated before use, or used in an unborated state. Unborated water is preferably not used in fuel pools for certain types of reactors due to significant criticality control issues. In the illustrated embodiment, the tubes 210, 230, baffle 220, and membranes 250 are part of a replaceable cartridge 270. The housing 200 may be opened to replace a used or fouled cartridge 270 with a fresh cartridge. In the illustrated embodiment, the membranes 250 comprise hollow fiber membranes. However, according to alternative embodiments, the membranes 250 may take any other suitable shape (e.g., sheet membranes that separate the strip gas regions from the water regions within the module 70). According to various embodiments, the membranes 250 may comprise any suitable gas transfer membrane that permits transfer of fission gases therethrough without permitting water transfer. For example, Liqui-Cel® Membrana Gas Transfer Modules may be used as the membranes 250. See “Membrane Contactors: An Introduction to the Technology,” Ultrapure Water® May/June 1996—UP130427. “BATMAN—Best Available Technique Minimising All Nuclides,” IVL Swedish Environmental Research Institute, Apr. 18, 2006, discloses other membranes that may be used as the membrane 250. In the illustrated embodiment in FIGS. 1-2, a single module 70 is shown. However, multiple modules 70 may be used without deviating from the scope of the present invention. Multiple modules 70 may be arranged in parallel and/or series between the passageways 110, 120. Parallel modules 70 may be used to increase the flow rate of water through the water treatment device 20. Serial modules 70 may be used to increase the amount of fission gas removed from the water by the water treatment device 20. The module(s) 70 may be sized to collectively provide once-through processing at a liquid flow rate of approximately 25-500, 25-300, and/or 50-100 gallons/minute, which may be sufficient to turn over the full volume of the sipping canister 500 in approximately 1-2 minutes and/or less than 5, 4, 3, 2, and/or 1 minutes. An additional 1-2 minutes of flushing can be performed at the start of each cycle without a noticeable impact to the overall throughput of the sipping system 10. According to various embodiments, the strip gas being used in the module 70 is kept at a slight to moderate vacuum pressure. This vacuum causes dissolved air and other gases in the water (in addition to fission gas) to diffuse into the strip gas. Removal of dissolved air from the water that is used in the canister 500 during the sipping procedure improves the sensitivity of the sipping procedure to detecting small leaks in the fuel assembly 600. In particular, by reducing the fraction of dissolved air in the canister water, less air is released from the canister water into the air bubble at the top of the canister 500 during the sipping procedure. As a result, the air bubble being sampled will comprise a relatively larger portion of leaked fission gas because the air bubble is not being diluted as much by air being released from the water. The larger concentration of any leaked fission gas in the air bubble being sampled by the detector 580 facilitates more accurate/sensitive detection of leaks. The control system 90 includes manual and/or automatic controls to control the pumps 50, 80, and may include additional sensors (e.g., flow/pressure sensors) that are used to determine and adjust the devices (e.g., pumps 50, 80) being controlled by the control system 90. As shown in FIG. 1, the sipping device 30 includes a canister 500. The canister 500 has a fuel assembly opening 500a, a water inlet 500b connected to the passageway 120 by a valve 510, and a water outlet 500c connected to the fuel pool 160 via a valve 520. The canister 500 also includes a lid 540 that may be selectively opened and closed (e.g., manually or automatically) to uncover and cover the opening 500a. The device 30 also includes a passageway 560 that leads from an upper, interior end of the canister 500 to a vacuum pump 570 and detector 580 disposed outside of the pool 160. According to various embodiments, the pump 570 and radiation detector 580 may be mounted to the frame 85. According to various embodiments, the sipping device 30 works in substantially the same manner as conventional sipping devices, except that water from the water treatment device 20 is used to flush and fill the canister 500. A sipping device control system 590 connects to various components of the sipping device 30 such as the valves 510, 520, the pump 570, and the detector 580, to turn these components on or off at the appropriate times in the sipping cycle. The control system 590 may be connected to or integrally formed with the control system 90 of the water treatment device 20 to ensure that the water treatment device 20 is operated at the desired points in the sipping cycle. According to various embodiments, the sipping cycle proceeds through the following sequential steps: 1) With the canister 500 disposed in the pool 165 and the lid 540 open, a fuel assembly 600 is loaded into the canister 500 through the opening 500a. 2) The lid 540 of the canister 500 is then closed, isolating the water in the canister 500 from the pool 160. 3) Pool water is degassed/filtered by running it through the water treatment device 20 and then providing the clean/degassed water to the canister 500. 4) With the valves 510, 520 open, the canister 500 is flushed with clean/degassed water provided by the water treatment device 20. Water is pumped into the canister 500 through the inlet 500b, downward, and out through the outlet 520, thereby flushing ambient (and potentially fission-gas-laden) pool water out of the canister 500. The volume of water being degassed and pumped into the canister 500 through the inlet 500b may be at least one or at least two times the total volume of the inside of the canister 500 to provide improved flushing of ambient, potentially fission-gas-laden pool water out of the canister 500 before the sipping procedure begins. The canister 500 pressure may be maintained at or near normal pool ambient pressure during the flush to avoid extracting fission gas from a leaking fuel assembly 600 during the flush. 5) Air is pumped into the top of the canister to form a bubble (e.g., a several inch tall air gap at the top of the interior of the canister 500). 6) The water valves 510, 520 are closed to seal the canister 500. 7) Some of the air bubble is pumped out of the air space at the top of the canister 500 via the pump 570 and passageway 560, thereby reducing the canister 500 internal pressure. 8) The reduction of pressure inside the canister 500 draws fission gas out of any cladding leaks in the fuel assembly 600 and into the canister 500 water. 9) Fission gases migrate to the air bubble at the top of the canister 500, are drawn out of the canister 500 by the vacuum pump 570, and are measured by the radiation detector 580. 10) At the end of the cycle, the canister lid 540 is opened and fuel assembly 600 is removed from the canister 500.According to various embodiments, various of these steps may be omitted or reordered without deviating from the scope of the present invention. For example, steps 5 and 6 may be switched in order. Steps 3 and 4 may occur simultaneously (or an intermediate buffer tank in the passageway 120 may be used). Steps 7-9 may occur simultaneously. According to various embodiments, the degassing process performed by the water treatment device 20 and flushing the canister 500 with clean/degassed water reduces the background level of fission gases and provides a lower, more stable fission gas background level than if the flush water from the fuel pool 160 were not degassed. This results in a higher signal-to-noise ratio in the detector 580, which may result in an improved failure detection accuracy and/or the ability to detect smaller failures in the fuel assembly 600 being tested. According to various embodiments, the system 10 and its use may be modified in one or more of the following ways without deviating from the scope of the present invention. For example, the module 70 and gas transfer membranes 250 may be substituted by an alternate gas transfer component such as a spray degasser or column degasser. For example, “Membrane Contactors: An Introduction to the Technology,” Ultrapure Water® May/June 1996—UP130427, discloses a method for extracting fission gasses from waste water, which may be used in place of the module 70 without deviating from the scope of the present invention. All or some of the components of the water treatment device 20 may be submersible in the pool 160. For example, according to various embodiments, the module 70 and pump 50 may be mounted to the canister 500 and submerged in the pool 160. This may facilitate the shortening or elimination of the passageways 100, 120. The strip gas passageways 70c, 70d may be lengthened. As shown in FIG. 1, a reservoir or buffer tank 620 may be disposed in the passageway 120 for storage of cleaned/degassed water to support a higher flushing flow rate or multi-pass treatment of pool water. For example, a lower throughput water treatment device 20 may continuously fill the reservoir 620, and cleaned/degassed water in the reservoir 620 may be intermittently used to flush the canister 500 by opening the valve 510. The resulting flushing may be at a higher volumetric flow rate than would be possible if the flushing flow rate was limited to the flow rate possible from the pump 50 and module 70. According to various embodiments, the various water passageways 100, 110, 120 and gas passageways 130, 140 may comprise any type of suitable passageways (e.g., rigid or flexible hoses, pipes, etc.). According to an alternative embodiment, the water treatment device 20 may be used without the sipping device 30 such that the device 20 continuously degasses the water in the pool 160 by sucking gas-laden water out of the pool through the passageway 100, degassing the water, and returning the water to the pool 160 through the open outlet end of the passageway 120. Such continuous degassing may improve the sensitivity of a standalone canister sipping device 30 (or other fuel assembly leak detection device (e.g., a telescope leak detection device) by lowering the ambient fission gas concentration in the pool 160 water. As used herein, the terms “clean water,” “degassed water,” “filtered water” and the like refer to water that has been filtered to reduce the concentration of gasses (e.g., fission gases) therein. Such water is still considered clean/degassed/filtered even if it is not completely devoid of dissolved gases such as fission gases. According to various embodiments, the water treatment device 20 removes at least 5, 10, 20, 30, 40, 50, 60, 70, 80, 90, 95, and/or 99% of the fission gases originally present in the water such that the degassed water has at least 5, 10, 20, 30, 40, 50, 60, 70, 80, 90, 95, and/or 99% less fission gases than before the filtering/degassing. As discussed above, various techniques may be used to increase the degree of filtering/degassing (e.g., the use of larger or sequential modules 70 and/or additional degassing equipment). FIG. 3 illustrates a vacuum canister sipping system 1010 according to an alternate embodiment of the present invention. The system 1010 is substantially similar to the system 10, except that the system 1010 uses an alternative water treatment device 1020 that degasses the water already in the canister, rather than degassing water outside of the canister. The system 1010 includes a sipping device 1030 that is substantially similar to or identical to the above-discussed sipping device 30. As with the device 30, the sipping device 1030 includes a canister 1500 and a passageway 1560 that leads from an upper, interior end of the canister 1500 to a vacuum pump 1570 and detector 1580 disposed outside of the pool. The sipping device 1030 also includes a control system 1590. These features of the sipping device 1030 are substantially similar or identical to the above-discussed features of the sipping device 30. Accordingly, a redundant description of similar or identical features is omitted. The water treatment device 1020 comprises an air sparging system 1020 connected to the canister 1500. The air sparging system 1020 comprises a sparge air supply 1700 (e.g., a tank of compressed sparge air; a compressor configured to compress ambient air to provide pressurized sparge air, etc.). In the illustrated embodiment, the sparge gas comprises air. However, according to various alternative embodiments, the sparge gas may comprise any other type of suitable gas into which fission gasses can diffuse. The system 1020 includes a sparge gas supply line 1710 that extends from the sparge gas supply 1700 to a sparge gas inlet 1720 at or near the bottom of the canister 1500. The sparge gas inlet 1720 may lead into a gas diffuser that separates the sparge gas into small bubbles 1770. As shown in FIG. 3, the system 1020 also includes a sparge gas outlet passageway 1740 connected the canister 1500 at or near the top of the canister 1500. The outlet passageway 1740 may open directly into the fuel pool so that fission-gas-laden sparge gas can vent into the pool or may lead to a filter/scrubber or other waste treatment device. A valve 1750 is connected to the outlet 1740. The valve 1750 may comprise a check-valve that permits sparge gas to exit the canister through the outlet 1740, while preventing liquid or gas from entering the canister 1500 via the outlet 1740 when the system 1010 applies a vacuum within the canister 1500. Additionally and/or alternatively, the valve 1750 may comprise a valve that is controlled by the control system 1590 so that the control system 1590 closes the valve 1750 before a vacuum is applied to the canister 1500 during the sipping procedure. According to various embodiments, the outlet passageway 1740 comprises an instrument cluster side port in the canister 1500. According to various alternative embodiments, the outlet 1740 and valve 1750 are omitted altogether, and sparge gas that has bubbled through the canister 1500 and absorbed fission gasses is removed from the canister 1500 via the passageway 1560. As shown in FIG. 3, sparge gas is introduced into the canister 1500 by the water treatment device 1020 after the fuel assembly 600 is loaded and the canister lid 1540 is closed. As bubbles 1770 of sparge gas rise within the sipping canister 1500, fission gasses diffuse into the bubbles 1770 and are carried to the top of the canister 1500. The sparge gas exits to the pool via the outlet 1740, and/or to the sipping system detection unit 1580 via the gas sample line 1560. Pressure inside the canister 1500 is preferably maintained at or near local ambient pressure within the fuel pool during the sparging process to prevent or limit the escape of fission gasses from a leaking fuel assembly 600. Venting the sparge gas out of the canister 1500 via the outlet 1740 and/or passageway 1560 may allow some or all of the sparge gas to escape and maintain the pressure inside the canister 1500 at close to ambient. Alternatively, the rates of air sparging and gas sampling may be actively controlled by the control system 1590 to maintain canister 1500 pressure at the desired conditions. According to various embodiments, the control system 1590 is operatively connected to the water treatment device 1020 so as to control the water treatment device 1020 (e.g., by controlling the timing and rate at which sparge gas is introduced into the canister 1500 and the timing and rate at which sparge gas is allowed to exit the canister via the outlet 1740 and/or passageway 1560). Sparging may continue for a predetermined duration sufficient to extract the desired fraction of fission gasses from the canister water. Alternatively, sparging may continue until the radioactivity of sparge air exiting via the gas sample line 1560 reaches a desired level, as detected by the detector 1580. After background fission gasses have been reduced enough to achieve the desired reduction in dissolved fission gas concentration, the sipping cycle continues similarly to existing sipping systems except that the background radioactivity measurements are lower than if sparging had not been performed, and leaking assemblies 600 are easier to identify. According to various embodiments, the system 1010 operates as follows (one or more steps of which may be automated via the control system 1590): 1. A fuel assembly 600 is placed in the canister 1500 and the lid 1540 is closed such that ambient fuel pool water surrounds the fuel assembly 600 within the canister 1500. 2. Sparge gas is fed from the sparge gas supply 1700 through the passageway 1710 and into the canister via the inlet 1720. 3. The sparge gas bubbles 1770 float up through the canister water, absorbing fission gases along the way. 4. The fission-gas-laden gas bubbles 1770 are vented into the pool (or waste treatment device) via the vent line 1740 or vented into the ambient environment (or a waste treatment device) via the detector 1580. 5. According to various embodiments, the fission-gas-laden gas bubbles 1770 are vented through the detector 1580 while the detector 1580 is used to detect the radioactivity of the fission-gas-laden gas bubbles 1770. Gas sparging may continue for a predetermined amount or time (e.g., at least 30, 60, 90, 120, 240, 300, 360, 600, 1200, 2400, and/or 4800 seconds or more) or may continue until the detector 1580 detects that the radioactivity of the fission-gas-laden gas bubbles 1770 has fallen below a predetermined threshold (e.g., less than 10, 25, 50, 100, 500, 1,000, 2,000, and/or 5,000 counts per second). Alternatively, gas sparging may continue until a radioactivity/fission gas concentration in the canister water is reduced by a predetermined threshold (e.g., at least 20, 30, 40, 50, 60, 70, 80, 90, 95, and/or 99% relative to an initial concentration of the fission gas in the canister water). The chosen threshold or time may be chosen by balancing the desired amount of time to be provided for fission gas removal against the corresponding improvement in leak detection. 6. Gas bubbles 1770 are allowed to float up to gas bubble/pocket at the top of the canister 1500. This gas bubble/pocket is sampled by the detector 1580, and the detected fission gasses/radiation is used as a baseline ambient radiation level for the subsequent sipping procedure. 7. The sipping procedure then begins by subjecting the canister 1500 to a vacuum pressure that causes fission gas to leak from leaks in the fuel assembly 600 (if present) into the canister water, float into the gas bubble, and be detected by the detector 1580. Increases in detected radiation/fission gasses relative to the baseline radiation detected in the above step are attributable to fuel assembly 600 leak(s), rather than to ambient fission gas in the canister 1500.According to various embodiments, various of these steps may be omitted or reordered without deviating from the scope of the present invention. The above-discussed embodiments (e.g., systems 10, 1010) improve leak detection by removing fission gases from the water in the canister during the sipping procedure, thereby reducing the extent to which fission gases in the water contaminate the air bubble being sampled. FIG. 4 illustrates a vacuum canister sipping system 2010 according to an alternate embodiment. In this embodiment, leak detection sensitivity is improved by preventing fission gas that is present in the canister water from altering the concentration-based radioactivity of the gas bubble during the vacuum sipping procedure. This is because fission gasses that are released from the canister water into the gas bubble during the vacuum sipping procedure are released in proportion to dissolved gas (e.g., air) that is also in the canister water because the concentrations are equilibrated. The matching equilibrated release of such fission gasses and air reduces or eliminates the extent to which such release impacts the radioactivity detected by the detector. Conventional sipping procedures involve forming a clean gas bubble in the canister over the fuel assembly. The air bubble is formed from plant air, which is relatively reliably low in radioactivity/fission gasses. The air bubble is formed with this clean air at the start of the sipping procedure. The canister is then subjected to a vacuum pressure, which causes fission gases to leak into the canister water through any leaks in the fuel assembly. This leaked fission gas floats up and into the air bubble at the top of the canister. Samples of the air bubble are detected over time by a fission gas detector. At the start of the procedure, the air going through the detector is just clean air from the air bubble. As fission gasses leak out of any holes in the fuel assembly, the fission gas concentration in the air bubble gradually rises, and that rise is detected by the detector as a leak. However, as the procedure progresses, dissolved fission gasses that were already in the canister water come out of solution and mix with the air in the bubble space to increase the radioactivity/fission gas concentration of that air bubble. Thus, even when there is no leaking assembly, there is a gradual increase in the activity observed by the detector because background fission gasses are gradually extracted from the water volume. At the same time, if there is a leaking assembly it will increase the inventory of fission gas in the canister. The determination of a leak requires that this increase in inventory is sufficiently large as to be noticeable relative to the expected gradual increase in radioactivity caused by the ambient fission gasses already in the canister water. The results of the sipping procedure can be compared to the results of a control sipping procedure in which a fuel assembly is not disposed in the canister according to various embodiments. However, such use of an initial control can lead to errors based on, for example, (1) variations in fission gas concentrations in different batches of fuel pool water that are disposed in the canister during the sipping procedure, and/or (2) an artificial local increase in the fission gases in the canister water as a result of residual fission gas remaining in the canister from a prior test of a leaking fuel assembly. The system 2010 is configured to reduce the extent to which the concentration of fission gasses in the air bubble and gas sample loop are influenced by the action of dissolved gasses coming out of solution within the canister during the vacuum sipping procedure. As shown in FIG. 4, the system 2010 includes a canister 2500 and lid 2540 that are substantially similar or identical to the canister 1500 and lid 1540 of the system 1010. As shown in FIG. 4, the system 2010 incorporates a controlled carrier gas recirculation loop 2020. An outlet port 2550 connects at or near the top of the canister 2540 and opens into a passageway 2560. A carrier gas supply 2570 connects to the passageway 2560 via a valve 2580. The loop 2020 progresses sequentially from the passageway 2560 through a valve 2590, a pump 2600, a detector 2610 (similar or identical to the detector 580), a pump 2620, a valve 2630, and a passageway 2640, which connects to an inlet 2650 into the canister 2500 at or near the bottom of the canister 2500. A vent/exhaust passageway 2660 connects, by way of a valve 2670 to the loop 2020 between the detector 2610 and pump 2620. Pressure sensors 2690, 2700 are operatively connected to the passageways 2560, 2640, respectively. The carrier gas supply 2570 may comprise any suitable supply of air or other suitable gas (e.g., a compressor that compresses ambient plant air for use in the system 2010; a compressed air tank). According to various embodiments, the system 2010 includes a control system 2800 that operatively connects to the various valves 1750, 2580, 2590, 2670, 2630, sensors 2690, 2700, pumps 2600, 2620, and detector 2610 to control the operation of the system 2010. According to various embodiments, the control system 2800 is configured to automatically undertake some or all of the sipping procedure once the fuel assembly 600 is placed inside the canister 2500. According to various embodiments, the system 2010 is operated (e.g., by the control system 2800) as follows: 1. A fuel assembly 600 is placed in the canister 2500 and the lid 2540 is closed such that ambient fuel pool water surrounds the fuel assembly 600 within the canister 2500. 2. Carrier gas (e.g., air) is delivered from the carrier gas supply 2570 to the top of the canister 2500 (e.g., via the open valve 2580, passageway 2560, and outlet 2550, or via the loop 2020 and passageway 2640). 3. The valve 2580 is closed (if not already closed) and the valves 2590, 2630 are opened (if not already open) so that the loop 2020 forms a continuous recirculation loop. The valve 2670 may be opened or closed (e.g., the valve 2670 may be opened to (1) maintain an ambient pressure in the detector 2610 to avoid damage to the detector 2610, and/or (2) avoid pressure fluctuations that could affect detector measurement accuracy/consistency). One or more of the pumps 2600, 2620 are operated to pump the carrier gas around the loop 2020 (in a clockwise direction as shown in FIG. 4), which causes the carrier gas to continuously recirculate through the canister water. Pressure within the canister 2500 is preferably maintained at or around the ambient pressure of the fuel pool so that fission gasses do not leak from a leaky fuel assembly 600 into the canister water during this recirculation step. 4. Radioactivity/fission gases in the carrier gas are continuously monitored by the detector 2610 during the recirculation. As ambient fission gasses in the canister water diffuse into the recirculating carrier gas, fission gas concentration in the recirculating carrier gas rises and is detected by the detector 2610. 5. Eventually, the fission gas concentration in the recirculating carrier gas reaches or approaches a steady state equilibrium, which is detected by the detector 2610 because the detected concentration stops increasing or increases at a slow rate. 6. Once the equilibrium is reached or approached, the actual sipping procedure test is conducted by applying a vacuum pressure to the canister 2500 while continuously sampling the carrier gas disposed in the gas bubble at the top of the canister 2500 via the detector 2610 and opening the valve 2670 to facilitate flow of the carrier gas from the gas bubble to the detector 2610 via the passageway 2560. The valves 1750, 2580, 2630 may be closed during this step. With the vacuum applied, if the fuel assembly 600 has a leak, fission gas will leak out of the fuel assembly 600, float into the bubble of carrier gas, and be detected by the detector 2610 as an increase in radioactivity relative to the baseline radioactivity of the carrier gas in the bubble once equilibrium was reached/approached. Additionally and/or alternatively, the valve 2630 may be opened and recirculation of the carrier gas may continue during the vacuum sipping procedure, which may improve leak detection by encouraging fission gas that is leaked into the canister water to be absorbed into rising carrier gas bubbles and then be detected in the detector 2610. Because the radioactivity of the carrier gas and dissolved gas mixture in the canister water is already near equilibrium, there is little or no change in carrier gas composition when dissolved gasses come out of solution upon application of vacuum during the time period during which the actual sipping procedure is performed. Thus, an observed change in carrier gas composition (e.g. in increase in fission gas concentration) during the application of vacuum is more reliably attributable to the presence of a leak in the fuel assembly 600, and not to the presence of ambient dissolved fission gas in the canister water.According to various embodiments, various of these steps may be omitted, modified, or reordered without deviating from the scope of the present invention. As discussed above, the recirculation step continues until the detected radiation in the recirculating carrier gas reaches or approaches equilibrium. The terms “reaches” and “approaches” do not require an absolute equilibrium. Rather, they are used to identify a proximity to equilibrium. For example, as used herein, equilibrium is reached or approached when the rate of increase in detected radioactivity in the recirculating carrier gas falls below a predetermined threshold (e.g., less than a 1, 2, 3, 4, 5, 10, 15, 20, 30, and/or 40% increase in radioactivity/fission gas concentration over a predetermined time period (e.g., 1, 2, 5, 10, 20, 30, 60, 90, 120, 600, 1200 seconds)). According to various embodiments, the fission gas removal features of the systems 10, 1010 and the equilibrium-based leak-detection sensitivity of the system 2010 are combined. For example, such a hybrid approach can be accomplished using the system 2010. To reduce the absolute amount of ambient fission gas in the canister water after the fuel assembly 600 is placed in the canister 2500 and the lid 2540 closed, sparge gas is fed from the supply air 2570 through the valve 2590, pump 2600, detector 2610, pump 2620, valve 2630, passageway 2640, and inlet 2650 to introduce the sparge gas/air into the canister 2500. As the sparge gas bubbles float up through the canister water, they absorb fission gasses from the canister water. The fission-gas-laden air is then vented from the system 2010 via the vent 1740 and valve 1750. Thus, the fission gas removal step functions in a manner similar to the fission gas removal step of the system 1010. According to alternative embodiments, a membrane filter such as the filter 70 from the system 10 may be incorporated into the system 2010 to reduce the absolute fission gas concentration in the canister 2500 water. After the absolute fission gas concentration in the canister water is reduced via gas sparging and venting and/or filtering, the system 2010 progresses through its equilibrium and sipping steps as discussed above. However, the reduced absolute fission gas concentration in the canister water reduces the fission gas concentration in the air bubble at equilibrium, which may improve the system's leak detection sensitivity because a leak-based increase in radioactivity/fission gas concentration is larger relative to the background radioactivity/fission gas concentration in the gas bubble being sampled. The foregoing illustrated embodiments are provided to illustrate the structural and functional principles of embodiments of the present invention and are not intended to be limiting. To the contrary, the principles of the present invention are intended to encompass any and all changes, alterations and/or substitutions within the spirit and scope of the following claims.
	summary
055815898	abstract	A method and an apparatus for producing microspherical ceramic particles such as particulate ceramic fuels for use in nuclear reactors. To produce gel particles by performing microwave heating on the small droplets of a stock solution or sol that contains uranium and other elements for nuclear reactor fuels, the small droplets are allowed to fall down through a cavity resonator using microwaves having an engineering frequency of 2.45 GHz and a quartz rod is inserted into or withdrawn out of the resonator to adjust the resonant frequency so that it will lie within the spectrum width of the microwave frequency.
	description	This application claims priority to and the benefit of Korean Patent Application No. 10-2008-0125407 filed in the Korean Intellectual Property Office on Dec. 10, 2008, the entire disclosure of which is incorporated herein by reference. The present invention relates to an electromagnetic wave interference (EMI)/radio frequency interference (RFI) shielding resin composite material, and a molded product made using the same. Electromagnetic wave generation is increased through highly efficient, high power consuming, and highly integrated electro-electronic devices. Electromagnetic waves cause malfunctions to other devices and systems or damage to the human body, so effective electromagnetic wave shield techniques are required in order to shield the electromagnetic waves. EMI shielding effectiveness is represented by the following Equation 1.S. B. (shielding effectiveness)=R+A+B [Equation 1] In the above formula, R represents surface reflection of an electromagnetic wave, A represents internal absorption of an electromagnetic wave, and B represents loss by multi-reflection. The conventional method of shielding electromagnetic waves includes an applied printing and plating method using a metallic material. Since the metallic material has high conductivity (R value, impedance is low) and a high electromagnetic wave shield rate through the surface reflection of electromagnetic waves, it is possible for even a thin metal to effectively shield electromagnetic waves. However, the printing and plating techniques, particularly the plating process, include complicated steps such as removing oils, etching, neutralizing, activating, accelerating, metal depositing, activating, first plating, second plating, third plating, and so on. Accordingly, these techniques have drawbacks such as high production costs and low productivity, particularly in view of recent demands for increased productivity. In contrast, an electromagnetic wave shielding material using a polymer composite resin can be obtained by simply injecting a composite resin, so it is a very economical process with regard to production cost and productivity. However, in the case of a composite material using the polymer composite resin, since the electrical conductivity is lower than that of a metallic material, it is important to improve the surface reflection and internal absorption among the factors shown in Equation 1. Accordingly, the resin composite material has the drawback of deteriorated or reduced electromagnetic wave shielding efficiency when it is too thin. In order to increase the electromagnetic wave shielding efficiency of a resin composite material, the surface impedance thereof is decreased (electrical conductivity is increased), the R value is increased, and internal electromagnetic wave scattering/absorption is further induced, so that the A value is increased to provide a highly effective electromagnetic wave shielding composite resin. The following publications relate to shielding of electromagnetic waves coming from all electronic devices such as radio frequency interference (RFI): an electromagnetic wave shielding device including a polymer substrate coated with metal on its surface (U.S. Patent Application Publication No. 2007-0199738); an electromagnetic wave shielding material including a non-conductive polymer, a conductive polymer, and an electrical conductive metal powder (U.S. Patent Application Publication No. 2007-0056769); a method of manufacturing an electrical conductive immersed fiber by coating a conductive fiber with a compatibilizer such as an organic wetting agent, and compositing the same in a resin (U.S. Patent Application Publication No. 2002-0108699); an electrically conductive thermoplastic elastomer including a conductive filler of nickel plated with silver in a styrene-ethylene-butadiene-styrene copolymer (SEBS) based matrix material which is a non-conductive resin (U.S. Pat. No. 6,638,448); an electrically conductive composition in which a carbonaceous conductive filler is immersed in a blend of two polymer resins having different polarities and the carbonaceous conductive filler is disposed on one having the higher polarity (U.S. Pat. No. 6,409,942); and a thermoplastic electromagnetic wave shielding sheet including a sheet material or polymer carrier that is capable of becoming porous during a thermoforming process and including a low-melting point metal conductive filler (U.S. Pat. No. 5,869,412). In addition, U.S. Pat. No. 5,183,594 discloses a conductive resin composite material including a tetrapod whisker, or including a powder, a flake, or a fiber together with the tetrapod whisker. However, these materials do not satisfy the required electromagnetic wave shield effects. An exemplary aspect of the present invention provides an electromagnetic wave interference (EMI)/radio frequency interference (RFI) shielding resin composite material having high electrical conductivity and excellent electromagnetic wave shield effects. Another aspect of the present invention provides a molded product made using the EMI/RFI shielding resin composite material. According to one aspect of the present invention, an EMI/RFI shielding resin composite material is provided that includes (A) a thermoplastic polymer resin, (B) a tetrapod whisker, and (C) a low-melting point metal. The EMI/RFI shielding resin composite material includes about 40 to about 84 volume % of the thermoplastic polymer resin (A), about 5 to about 59 volume % of the tetrapod whisker (B), and about 1 to about 10 volume % of the low-melting point metal (C). The tetrapod whisker (B) may be made of ZnO. The tetrapod whisker (B) may have four feet and a body, the foot may have a length of about 2 to about 100 μm from end to end, and the body may have a size of about 0.15 to about 10 μm. The tetrapod whisker (B) may be coated with a conductive material on its surface, and the conductive material may include silver, copper, aluminum, iron, palladium, tin oxide, indium oxide, silicon carbide, zirconium carbide, titanium carbide, graphite, nickel, or a combination thereof. The low-melting point metal (C) may be a solid solution including at least two kinds of metal elements. The low-melting point metal (C) may have a main component including tin, bismuth, lead, or a combination thereof, and a minor component including copper, aluminum, nickel, silver, germanium, indium, zinc, or a combination thereof, and it may have a lower solidus temperature than temperatures used in the process of making the EMI/RFI shielding resin composite material. The EMI/RFI shielding resin composite material may further include (D) a glass fiber filler in an amount of about 50 parts by weight or less based on about 100 parts by weight of the EMI/RFI shielding resin composite material. According to another aspect of the present invention, a molded product made using the EMI/RFI shielding resin composite material is provided. Hereinafter, further embodiments of the present invention will be described in detail. The present invention now will be described more fully hereinafter in the following detailed description of the invention, in which some, but not all embodiments of the invention are described. Indeed, this invention 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 satisfy applicable legal requirements. As used herein, when specific definition is not provided, the term “electromagnetic wave interference (EMI)/radio frequency interference (RFI)” refers to “electromagnetic wave interference (EMI) or radio frequency interference (RFI)”. According to one embodiment, electromagnetic wave shielding effectiveness (S.B) represented by the following Equation 1 may be improved by decreasing impedance, that is, by improving conductivity.S.B.=R+A+B [Equation 1] In the above Equation 1, R represents surface reflection of electromagnetic waves (electrical conductivity), A represents inner absorption of electromagnetic waves, and B represents loss through multi-reflection. The EMI/RFI shielding resin composite material according to one embodiment of the present invention includes (A) a thermoplastic polymer resin, (B) a tetrapod whisker, and (C) a low-melting point metal. The EMI/RFI shielding resin composite material includes about 40 to about 84 volume % of the thermoplastic polymer resin (A), about 5 to about 59 volume % of the tetrapod whisker (B), and about 1 to about 10 volume % of the low-melting point metal (C). The EMI/RFI shielding resin composite material according to one embodiment is prepared by mixing the components. The composite material has a structure including a matrix of the thermoplastic polymer resin, and the tetrapod whisker and low-melting point metal dispersed in the matrix to provide a network. Exemplary components included in the EMI/RFI shielding resin composite material according to embodiments of the present invention will hereinafter be described in detail. (A) Thermoplastic Polymer Resin Exemplary thermoplastic polymer resins include without limitation polyamides; polyalkylene terephthalates such as polyethylene terephthalate, polybutylene terephthalate, and the like; polyacetals; polycarbonates; polyimides; polyphenylene oxides; polysulfones; polyphenylene sulfides; polyamide imides; polyether sulfones; liquid crystal polymers; polyetherketones; polyetherimides; polyolefins such as polyethylene, polypropylene, and the like; acrylonitrile-butadiene-styrene; polystyrenes; syndiotactic syndiotactic polystyrenes; and the like; and combinations and blends thereof. In one embodiment, the thermoplastic polymer resin is a crystalline polymer resin such as polyethylene, polypropylene, polyethylene terephthalate, polybutylene terephthalate, syndiotactic polystyrene, polyetherketone, polyphenylene sulfide, and the like, and combinations thereof. When the EMI/RFI shielding resin composite material includes polyphenylene sulfide, for example, it is possible to simultaneously increase crystallization speed and crystallinity, decrease viscosity and absorption rate, and improve heat resistance. The EMI/RFI shielding resin composite material may include the thermoplastic polymer resin in an amount of about 40 to about 84 volume %, and in another embodiment, about 60 to about 80 volume %, based on the total amount of EMI/RFI shielding resin composite material. When the EMI/RFI shielding resin composite material includes the thermoplastic polymer resin in an amount within this range, the process and the EMI shield efficiency can be improved. (B) Tetrapod Whisker The tetrapod whisker according to one embodiment is used for a filler, and can provide the benefit of easily providing an inter-filler network to effectively decrease the impedance. It is currently believed that the tetrapod whiskers may be uniformly distributed in the polymer matrix, regardless of the flow of the molded composition, and thus can be helpful for the formation of the network. In addition, the tetrapod whisker can easily induce the internal electromagnetic wave scattering of the EMI/RFI shielding resin composite material due to the four-directional complex shape, and simultaneously, can further effectively induce the internal electromagnetic wave scattering of EMI/RFI shielding resin composite material due to a partial wear and broken-cut of the feet. As a result, the internal absorption value (A value) of electromagnetic waves can be effectively increased. The tetrapod whisker may be made of ZnO. The ZnO may be obtained by thermal evaporation of Zn powder. In addition, the tetrapod whisker may be a whisker comprising four feet, each having a length ranging from about 2 to about 100 μm from end to end, and a body having a size of about 0.15 to about 10 μm. When the tetrapod whisker has feet and a body with sizes within these ranges, it can be more effective in providing an inter-whisker network. Specifically, when the foot length is about 10 to about 40 μm from end to end and simultaneously the foot aspect ratio is less than about 30, and in one embodiment, about 3 to about 30, and the body size is about 1 to about 10 μm, the tetrapod whisker may be more suitable for providing an inter-whisker network and a polymer fusion process. The term aspect ratio refers to the ratio of the length/width of the tetrapod whisker foot. When the aspect ratio is about 30 or more, the network effect may be deteriorated since the foot of the tetrapod whisker can be easily broken. The tetrapod whisker may be coated with a conductive material on its surface, and in that case, it can provide electrical conductivity. Exemplary conductive materials include without limitation silver, copper, aluminum, iron, palladium, tin oxide, indium oxide, silicon carbide, zirconium carbide, titanium carbide, graphite, nickel, and the like, and combinations thereof. The EMI/RFI shielding resin composite material may include the tetrapod whisker in an amount of about 5 to about 59 volume %, for example about 20 to about 40 volume %, based on the total amount of EMI/RFI shielding resin composite material. When the EMI/RFI shielding resin composite material includes the tetrapod whisker in an amount in this range, the inter-whisker network can be easily formed. (C) Low-Melting Point Metal The low-melting point metal according to one embodiment can maximize the inter-filler network. It can also easily form the inter-filler network together with the tetrapod whisker to effectively decrease the impedance, so as to further improve the electromagnetic wave shield efficiency. The low-melting point metal is a solid solution including at least two kinds of metal elements, and includes a main component (i.e., a majority component comprising greater than 50%, for example at least about 75%, or at least about 85%, or at least about 90%, or higher, of the total weight percent of the low-melting point metal) and a minor component (i.e., a minority component comprising less than 50% of the total weight of the low-melting point metal). Exemplary main components include without limitation tin, bismuth, lead, and combinations thereof, and exemplary minor components include without limitation copper, aluminum, nickel, silver, germanium, indium, zinc, and combinations thereof. According to one embodiment, the main component includes tin for an environmentally-friendly material. The low-melting point metal may have a solidus temperature (temperature at which solidification is completed) that is lower than temperatures used in the process of making the EMI/RFI shielding resin composite material. When the low-melting point metal has a solidus temperature that is lower than temperatures used to make the EMI/RFI shielding resin composite material by more than about 20° C., it can be beneficial for the process of manufacturing a composite material and forming an inter-filler network. In another embodiment, the low-melting point metal has a solidus temperature that is higher by more than about 100° C. than downstream processing temperatures, such as used in molding the composite material to form a final end product, which can provide the benefit of stability. A more detailed description follows. In order for the low-melting point metal to form a network in the thermoplastic polymer resin while manufacturing the EMI/RFI shielding resin composite material, the solidus temperature and the liquidus temperature (temperature at which solidification begins) of the low-melting point metal that affect the dispersion have the following order: liquidus temperature>melting point of thermoplastic polymer resin>solidus temperature. Such solidus temperature may be controlled by the amount ratio of the main component and the minor component of the low melting point metal. Thereby, it is possible to control the physical properties such as liquidus temperature and mechanical strength. Specifically, when the conductive coating material of the tetrapod whisker is aluminum, the solid solution beneficially includes aluminum; similarly, when the conductive coating material is copper, the solid solution beneficially includes copper. One method of controlling a solidus temperature of the low-melting point metal by adjusting the element amount of the main component and the minor component includes providing a solid solution with another metal. Non-limiting examples of the method include controlling the solidus temperature of tin/copper (97/3 weight ratio) to 227° C., or controlling the solidus temperature of tin/copper/silver (92/6/2 weight ratio) to 217° C. The EMI/RFI shielding resin composite material may include the low-melting point metal in an amount of about 1 to about 10 volume %, and in another embodiment, about 2 to about 5 volume % based on the total amount of the EMI/RFI shielding resin composite material. When the EMI/RFI shielding resin composite material includes the low-melting point metal in an amount within this range, it can be possible to further enforce the network of the tetrapod whisker. (D) Glass Fiber Filler The glass fiber filler may be further included in the EMI/RFI shielding resin composite material to improve the strength according to one embodiment. The glass fiber filler may have a diameter of about 8 to about 13 μm and a length of about 2 to about 5 mm, but is not limited thereto. When the glass fiber filler has a diameter and length within these ranges, it can benefit the enforcement effect and the process of making the composite material. The EMI/RFI shielding resin composite material may include the glass fiber filler in an amount of about 50 parts by weight or lower, and in another embodiment, about 2 to about 50 parts by weight, based on about 100 parts by weight of the EMI/RFI shielding resin composite material. When the EMI/RFI shielding resin composite material includes the glass fiber filler in an amount within this range, it can be possible to improve the strength of EMI/RFI shielding resin composite material. (E) Other Additives The EMI/RFI shielding resin composite material according to one embodiment may further include a variety of known additives, as required, such as an antioxidant, an ultraviolet (UV) absorber, a flame retardant, a lubricant, a dye and/or pigment, and so on, as long as they do not damage the effects of the invention. The skilled artisan will understand the types and amounts of additives and how to use additives in the present invention without undue experimentation. The EMI/RFI shielding resin composite material may include the additives in an amount of about 0 to about 60 parts by weight, and in another embodiment, about 1 to about 30 parts by weight, based on about 100 parts by weight of the EMI/RFI shielding resin composite material. Another embodiment of the present invention provides a molded product made using the EMI/RFI shielding resin composite material. The molded product is applicable in fields requiring an EMI/RFI shield, and it is particularly applicable for a display device such as a TV and a PDP requiring an excellent EMI/RFI shield, and an electro-electronic device such as a computer, a mobile phone, and an office automation device. The following examples illustrate the present invention in more detail. However, they are exemplary embodiments of the present invention and are not limiting. A person having ordinary skill in this art can sufficiently understand parts of the present invention that are not specifically described. (A) Thermoplastic Polymer Resin The thermoplastic polymer resin is polyphenylene sulfide (PPS). Ryton PR-35 manufactured by Chevron Phillips Chemical is used for the PPS resin, and it has a zero shear viscosity of 1000[P] measured at 315.5° C. under a nitrogen atmosphere. (B) Tetrapod Whisker The tetrapod whisker is made of ZnO obtained by thermal evaporation using Zn powder, and it has a foot length of 10 to 50 μm and a body diameter of 1 to 10 μm. In addition, the tetrapod whisker is coated with silver on its surface. (C) Low-Melting Point Metal A tin/copper/silver low-melting point metal having a main component of tin is used as the low-melting point metal. The mixing ratio of each element of the low-melting point metal is tin/copper/silver=92/6/2 wt %, the solidus temperature is 217° C., and the liquidus temperature is 375° C. (D) Glass Fiber Filler The glass fiber filler is ECS 03 T-717PL (manufactured by Nippon Electric Glass) having a diameter of 10 μm and a length of 3 mm, and coated with silanes on its surface to improve interface adherence with the thermoplastic polymer resin of PPS. Using the components, an EMI/RFI shielding resin composite material is prepared in each composition of Examples 1 to 4 shown in the following Table 1 and Comparative Examples 1 to 4 shown in the following Table 2, in accordance with the process (temperature: 300° C.), and is pressed with a common twin screw extruder and an injector to provide a pellet. As shown in the following Table 1, if the amount of glass fiber filler is recalculated into parts by weight, it would be 6.4 parts by weight based on the total of 100 parts by weight of the EMI/RFI shielding resin composite material. Using the obtained pellets, the specific volume resistance is measured in accordance with the ASTM D257 method, and the electromagnetic wave shield efficiency is measured in accordance with the ASTM D4935 method using a specimen having a thickness of 2.1T. The results are shown in the following Tables 1 and 2, respectively. TABLE 1ExamplesVolume %1234PPS88786863silver-coated tetrapod10203030whiskerSn/Cu/Ag low-melting 2 2 2 2point metalglass fiber filler——— 5specific volume3.0 × 10−21.8 × 10−26.6 × 10−38.3 × 10−3resistance [Ω/cm]average shielding26354540effect [dB] at 2.1 T TABLE 2Comparative ExamplesVolume %1234PPS78787878silver-coated tetrapod22———whiskersilver-coated whisker1)—22——silver flake2)——22—nickel powder3)———22specific volume3.4 × 10−29.8 × 10−12.2 × 10−25.8 × 102resistance [Ω/cm]average shielding281522 5.5effect [dB] at 2.1 T1)silver-coated potassium titanate whisker, which is a fiber-structured whisker having a diameter 0.3 to 0.6 μm and a length 10 to 20 μm2)silver flake having a diameter of 10 to 40 μm3)nickel powder of 5 to 30 μm From the results shown in Tables 1 and 2, Examples 2 to 4 including both the tetrapod whisker and the low-melting point metal have lower or comparative specific volume resistances than those of Comparative Examples 1 to 4, and an excellent shield effect. In addition, although Example 1 includes the tetrapod whisker and the low-melting point metal in smaller amounts than Comparative Example 1, it exhibits similar specific volume resistance and shield effects. Furthermore, comparing Example 2 including the tetrapod whisker and the low-melting point metal to Comparative Example 1 including only the tetrapod whisker in the same amount as the total amount of tetrapod whisker and low-melting point metal used in Example 2, Example 2 exhibits lower specific volume resistance than that of Comparative Example 1, and excellent shield efficiency. Particularly, it is understood that an inter-filler network formed by using the tetrapod whisker and the low-melting point metal according to the examples is superior to that of only using the tetrapod whisker or that of only using the silver flake or the nickel powder according to the comparative examples. In addition, comparing Comparative Example 1 to Comparative Example 2 confirms that the tetrapod whisker is superior to the fiber whisker. Comparing Comparative Example 1 to Comparative Example 3 confirms that the inter-filler network provided by using the silver-coated tetrapod whisker is superior to that of using the silver flake. Particularly, the examples confirm that the low-melting point metal provides the whisker network with enforcement effects. Many modifications and other embodiments of the invention will come to mind to one skilled in the art to which this invention pertains having the benefit of the teachings presented in the foregoing description. Therefore, it is to be understood that the invention is not to be limited to the specific embodiments disclosed and that modifications and other embodiments are intended to be included within the scope of the appended claims. Although specific terms are employed herein, they are used in a generic and descriptive sense only and not for purposes of limitation, the scope of the invention being defined in the claims.
	description	The present invention relates to radiation case. The radiation case can be used to irradiate a quartz crystal with gamma rays. Radiation rods are located in vertical cavities of the radiation case. Each radiation rod holds a radioactive pellet. The quartz crystal is irradiated in a radiation chamber of the radiation case, by gamma rays emitted from a radioactive pellet in each of the radiation rods. The vertical cavities are precisely positioned in the radiation case. The vertical cavities are formed so as to be in close proximity to a back wall of a back portion of the radiation case. A thin partition is formed between the vertical cavities and the back wall of the back portion of the radiation case. Again, each of the radiation rods contains a radioactive pellet. The radioactive pellets emit gamma rays. The gamma rays pass through the thin partition of the radiation case, and into a radiation chamber of the radiation case. The gamma rays can then pass into the quartz crystal, to dislodge positive ions interstitially located in the quartz crystal. The radiation case has a door and extended main section. The extended main section has a back portion, a floor portion, a ceiling portion, and side portions. The door, back portion, a floor portion, a ceiling portion, and side portions form a radiation chamber. Vertical cavities are formed near to the back wall of the radiation chamber. A thin partition is formed between the vertical cavities and the back wall of the radiation chamber. The vertical cavities are designed to hold radiation rods. The thin partition is thin enough, so that gamma rays, that are emitted from the radiation rods, can pass from the vertical rods and into the radiation chamber. One of the side portions of the main section has an channel through which an electrical cable and a vacuum hose pass, from beneath the radiation case, into the radiation chamber. The electrical cable and a vacuum hose are connected to an apparatus that is placed in the radiation chamber. The apparatus hold a quartz crystal that is irradiated within the chamber. A dolly supports the apparatus. The dolly allows the apparatus to be quickly moved into and out of the radiation chamber. The apparatus has an ion pump coupling. The ion pump coupling is connected to the vacuum hose. The apparatus has two electrodes for holding a quartz crystal. One of two electrodes is supported by a frame of the apparatus. Electrical conductor lines, that are in the electrical cable, are connected to the two electrodes. The door of the radiation case is quickly opened to place the dolly into the radiation chamber. Then the door is quickly closed, to prevent any undue amount of external radiation exposure. A radiation case, comprising a radiation-proof door; and a radiation-proof main section, the radiation-proof main chamber section comprising a back portion, a floor portion, a ceiling portion and side portions, vertical cavities formed in the back portion, the vertical cavities being a distance from a surface of a back wall of the back portion, radiation rods located in the vertical cavities, each radiation rod containing cobalt-60 pellet, the distance between the vertical cavities and the surface of the back wall of the back portion being less than a penetration distance for gamma rays coming out of each cobalt-60 pellet. FIG. 1 shows a sectional side view of a radiation case 10. The radiation case 10 has a radiation proof door 12 and a radiation proof main section 13. The radiation case 10 can contain gamma rays, such as gamma rays 11, that are within radiation case 10. The radiation-proof door 12 and radiation-proof main section are preferably made from a lead metal. The radiation-proof door 12 can be quickly opened and quickly closed. The quick opening and closing of door 12 prevents an escape of an undue number of gamma rays from radiation case 10. The radiation-proof main section 13 has a floor portion 14, a ceiling portion 15, a back portion 16, and side portion 17 and 18 shown in FIG. 2. The door 12, floor portion 14, ceiling portion 15, back portion 16, and side portions 17 and 18, form a radiation chamber 20. The radiation chamber 20 is positioned within the radiation case 10. Gamma rays are held within the radiation chamber 20 of radiation case 10, when door 12 is closed. The back portion 16 is extended away from the door 12 of the radiation case 10, to prevent an escape of an undue number of gamma rays from the chamber 20 of radiation case 10, when door 12 is open. Since the main section 13 is extended, a lesser number of gamma rays will come out of the radiation case 10 from the back portion 16 of the radiation case 10, when door 12 is open. The back portion 16 has a back wall 19. Aligned, cylindrical, vertical cavities, such as vertical cavity 24, are formed in the back portion 16. Such vertical cavities 23, 24, 25 and 26 are shown in FIG. 2. The vertical cavities are positioned close to the back wall 19, within the back portion 16 of radiation case 10. Between a surface 29 of the back wall 19 and the vertical cavities is a thin partition 30. The partition 30 is thin enough to allow gamma rays, such as gamma rays 11, to pass through partition 30 and into chamber 20. A separate cylindrical radiation rod is placed into each of the cylindrical, vertical cavities. Cylindrical radiation rod 31 is tightly placed into cylindrical, vertical cavity 24. As shown in FIG. 1, the radiation rod 31 holds a radioactive pellet 36 near its longitudinal center. The radioactive pellet 36 is made from cobalt-60. The cobalt-60 pellet 36 emits gamma rays 11. Each of the radiation rods 31, 32, 33 and 34 holds a cobalt-60 pellet. Each pellet emits gamma rays. The cylindrical, vertical cavities are located 0.2 centimeters from the surface 29 of the back wall 19 of back portion 16. The cavities are formed in the back portion 16 so that a 0.2 centimeter thick partition 30 is formed in back portion 16. The 0.2 thick partition 30 is thin enough to allow 1.173 Mev gamma rays from a cobalt-60 pellet to pass from a cavity into radiation chamber 20. The cavities are aligned to be parallel to the surface 29 of the back wall 19. A selected distance between the vertical cavities and the surface 29 of the lead back wall 19 is made to be less than a maximum penetration distance through partition 30, for gamma rays coming out of the cobalt-60 pellets in the radiation rods. Again, FIG. 2 shows a sectional top view of radiation case 10. FIG. 2 shows side portions 17 and 18 of radiation case 10. FIG. 2 shows cavities 23, 24, 25 and 26 of radiation case 10. Radiation rods 34, 31, 32 and 33 are positioned, respectively, in the vertical cavities 23, 24, 25 and 26. Each of the radiation rods holds a cobalt-60 pellet. FIG. 2 shows a narrow channel 41 that is located in side portion 17. An electrical cable 43 and vacuum hose 45, from vacuum equipment and electrical power equipment located below the case 10, pass through the channel 41. The channel 41 are filled with a lead based sealer 42, to keep gamma rays from passing through channel 41. While the present invention has been disclosed in connection with the preferred embodiment thereof, it should be understood that there may be other embodiments which fall within the spirit and scope of the invention as defined by the following claims.
051184621	claims	1. Manipulator for handling operations for non-destructive testing in the vicinity of the nozzle of a vessel in the primary loop of a nuclear power plant, comprising a carriage, means for moving said carriage in circumferential direction around a nozzle of a vessel, a sled disposed on said carriage, means for displacing said sled in the axial direction of the nozzle, a shoulder joint disposed on said sled, a scissors half having an upper arm with one end pivotably supported in said shoulder joint and another end, a lower arm with a free end, another joint pivotably connecting the other end of said upper arm to said lower arm, a holder, and a further joint pivotably connecting said holder to the free end of said lower arm. 2. Manipulator for handling operations in the vicinity of the nozzle of a vessel, comprising a carriage, means for moving said carriage in circumferential direction around a nozzle of a vessel, a sled disposed on said carriage, means for displacing said sled in the axial direction of the nozzle, a shoulder joint disposed on said sled, a scissors half having an upper arm with one end pivotably supported in said shoulder joint and another end, a lower arm with a free end, another joint pivotably connecting the other end of said upper arm to said lower arm, a holder, and a further joint pivotably connecting said holder to the free end of said lower arm. 3. Manipulator according to claim 2, including a tool disposed on said holder. 4. Manipulator according to claim 2, including a probe disposed on said holder. 5. Manipulator according to claim 2, including a rack drive mechanism for displaceably supporting said shoulder joint on said sled. 6. Manipulator according to claim 5, wherein said rack drive mechanism has a rack being disposed on said carriage and having a toothless guide element for rerailing and derailing said sled. 7. Manipulator according to claim 2, including a drive motor with a position transducer and a gear operatively connecting said drive motor and said sled for adjusting the position of said sled, and a control device connected to said position transducer. 8. Manipulator according to claim 7, including another drive motor with another position transducer and another gear operatively connecting said other drive motor and said shoulder joint for adjusting a pivoting angle of said upper arm, said control device being connected to said other position transducer. 9. Manipulator according to claim 7, including a further drive motor with a further position transducer and a further gear operatively connecting said further drive motor and said other joint for adjusting a pivoting angle between said upper and lower arms, said control device being connected to said further position transducer. 10. Manipulator according to claim 8, including a further drive motor with a further position transducer and a further gear operatively connecting said further drive motor and said other joint for adjusting a pivoting angle between said upper and lower arms, said control device being connected to said further position transducer. 11. Manipulator according to claim 10, wherein said gears are bevel gears. 12. Manipulator according to claim 1, wherein said other joint includes means for permitting said lower arm to be folded back onto said upper arm, and said shoulder joint includes means for permitting said upper arm to be folded onto said carriage. 13. Manipulator according to claim 1, wherein said lower arm has a shorter length than said upper arm. 14. Manipulator according to claim 1, wherein said lower arm is approximately two-thirds the length of said upper arm. 15. Manipulator according to claim 1, including an annular rail to be disposed concentrically about the nozzle of the vessel, and means for derailing said carriage from and rerailing said carriage onto said rail.
039502716	description	The invention will be further described in the following examples: EXAMPLE 1 An E-CTFE resin containing 98.05 weight % of approximately equimolar ethylene and chlorotrifluoroethylene, 1.0% trialkylisocyanurate, 0.3% of a phosphite of a stearically hindered phenol, 0.15% distearyl-thiodiproprionate and 0.5% calcium oxide was ball blended with 20 weight % gadolinium oxide of average particle size 10 .mu. and compression molded into plaques 2 .times. 2 in. .times. 1/4 inch. These plaques were divided into three groups and each group exposed to concentrated nitric acid containing 0.5N HF. In one case, the nitric acid was boiled, in another the mixture run hot and the third group was kept at room temperature. After five weeks all samples were weighed and examined and found to have suffered no weight change and no degradation. Similarly, exposure to high degrees of radiation for an extended period of time resulted in no degradation. These results indicate that this gadolinium filled composition would be superior for use as Raschig rings, containers, pipes or in other applications requiring exposure to acidic and/or radioactive environments. EXAMPLE 2 The procedure of Example 1 was repeated using a PCTFE resin and resulted in a composition possessing similar properties on exposure to acidic and radioactive environments. EXAMPLES 3 - 5 The procedure of Example 1 was repeated using equivalent amounts of gadolinium boride, gadolinium aluminate and gadolinium aluminum borate. Again superior results were obtained. EXAMPLES 6 - 19 In order to show that these chloro-fluoro substituted ethylene compositions can be highly filled with gadolinium compounds and still retain superior mechanical properties, the following experiments were performed: The filling procedure of Example 1 was repeated to fill both E-CTFE and PCTFE resins, using filler in amounts of 1 1/2, 10, 25, 50, 75 and 90 weight percent. Composite test plaques were molded under 300 psi pressure and at temperature of 260.degree.C. for 20 minutes. Molding pressures were increased to 10,000 psi for PCTFE samples containing 75 weight percent filler. The mechanical properties of these plaques were tested with both polymers using flexural strength as the criterion. The results of these tests are tabulated in Table I. Table I ______________________________________ Mechanical Properties of Polymers Filled with Gadolinium Oxide Vol. % Bulk Flexural Stress Flexural Wt.% Gd.sub.2 O.sub.3 Measured at 1% Strain Modulus In E-CTFE Density (psi) (psi) ______________________________________ 0 0 1.69 2010 2 .times. 10.sup.5 1.5 .4 1.705 2000 2.1 .times. 10.sup.5 10 2.47 1.832 2210 2.2 .times. 10.sup.5 25 7.05 2.12 2570 2.7 .times. 10.sup.5 50 18.5 2.74 4160 4.3 .times. 10.sup.5 75 40.7 3.92 6500 (max) 8.2 .times. 10.sup.5 In PCTFE 0 0 2.15 1540 1.8 .times. 10.sup.5 1.5 .5 2.17 2040 1.6 .times. 10.sup.5 10 3.1 2.31 2220 2.1 .times. 10.sup.5 25 8.8 2.60 2580 2.7 .times. 10.sup.5 50 22.5 3.29 4390 4.4 .times. 10.sup.5 75 46.5 3.74 7830 10.8 .times. 10.sup.5 ______________________________________
	summary
	claims	1. A charged particle beam irradiation apparatus that irradiates a particle beam to a subject, comprising:a particle beam generation unit from which the particle beam is radiated;a ridge filter, including a plurality of ridges, exhibiting a cyclic thickness distribution for causing the particle beam to exhibit a desired energy distribution, wherein a bottom of each ridge from the plurality of ridges is arranged at an angle different from respective angles of bottoms of the remaining ridges from the plurality of ridges. 2. The charged particle beam irradiation apparatus according to claim 1, wherein a bottom of each ridge from the plurality of ridges of the ridge filter is arranged to be perpendicular to an entering direction of the particle beam. 3. The charged particle beam irradiation apparatus according to claim 1, wherein a bottom of each ridge from the plurality of ridges of the ridge filter is further arranged at a position different from respective positions of bottoms of the remaining ridges from the plurality of ridges. 4. The charged particle beam irradiation apparatus according to claim 1, wherein the ridge filter has the arrangement spacing for each ridge modulated in a direction, in which a cyclic depth distribution is created, according to a change in the angle of each ridge. 5. The charged particle beam irradiation apparatus according to claim 1, wherein each ridge from the plurality of ridges is a bar ridge, and a shape of a bar ridge is varied depending on a distance from the center of a radiation field in the long-side direction of the ridge filter. 6. The charged particle beam irradiation apparatus according to claim 1, wherein a plurality of ridge filters are prepared in association with different conditions for radiation-field formation, and selectively disposed on a beam path, along which the particle beam passes, according to a desired condition for radiation-field formation.
	abstract	A system and method for monitoring a reliability status of an actuator include determining a virtual actuator load value (VALV) based on various load factor parameters of a structural system component controlled by the actuator, absent sensed load values from the actuator. A virtual output force value (VOFV) is determined based on various actuator operational control values that occur in response to a position command from a main control and monitoring system of the structural system. A virtual torque efficiency (VTE) of the actuator is calculated based on the VALV and the VOFV. The VTE of the actuator is periodically calculated as the measured load factor parameters and the measured actuator operational control values change during operation of the structural system. The VTE data is collected, stored and analyzed to monitor the reliability status of the actuator during the life of the actuator.
	claims	1. A core spray sparger assembly for a nuclear reactor comprising fuel assemblies, a top guide, coolant supply pipes and a shroud head, said core spray sparger assembly comprising: at least one coolant manifold, each said coolant manifold comprising a plurality of independent coolant manifolds joined by a plurality of resilient couplings; at least one coolant coupling in fluid communication with said coolant manifold; at least one mounting device configured to couple said coolant manifold to the nuclear reactor; a plurality of fluid conductors in a parallel array, positioned above the top guide, said fluid conductors in fluid communication with said coolant manifold; and a plurality of nozzles in fluid communication with said fluid conductors. 2. A core spray sparger assembly in accordance with claim 1 wherein said at least one coolant manifold comprises two fluidically independent, coaxial, substantially circular, coolant manifolds, joined by a plurality of resilient couplings. claim 1 3. A core spray sparger assembly in accordance with claim 1 , further comprising claim 1 at least one alignment guide configured to align said at least one coolant manifold with the top guide. 4. A core spray sparger assembly in accordance with claim 3 , wherein at least one of said coolant manifolds comprises at least one alignment guide configured to align said at least one coolant manifold with the coolant supply pipes. claim 3 5. A core spray sparger assembly in accordance with claim 1 , wherein said plurality of fluid conductors comprise at least two parallel arrays of fluidically independent, parallel fluid conductors configured to be positioned above the reactor top guide, each parallel array in fluid communication with one of said coolant manifolds. claim 1 6. A core spray sparger assembly in accordance with claim 5 , wherein said plurality of nozzles are formed in each said parallel array such that each said parallel array is configured to supply coolant to each fuel assembly. claim 5 7. A core spray sparger assembly in accordance with claim 1 , wherein said at least one mounting device comprises a plurality of spring retainers configured to secure at least one coolant manifold between the top guide and the shroud head. claim 1 8. A core spray sparger assembly in accordance with claim 1 , wherein said at least one mounting device comprises a plurality of hanger bolts configured to secure at least one said coolant manifold to the shroud head. claim 1 9. A core spray sparger assembly in accordance with claim 8 wherein said at least one mounting device further comprises a trunnion. claim 8 10. A core spray sparger assembly in accordance with claim 1 further comprising at least one alignment guide. claim 1 11. A core spray sparger assembly in accordance with claim 10 , wherein said at least one alignment guide comprises at least one alignment channel on at least one fluid conductors, said alignment channel configured to engage the reactor top guide. claim 10 12. A core spray sparger assembly in accordance with claim 1 further comprising at least one stabilizing member coupled to at least one of said coolant manifolds. claim 1 13. A core spray sparger assembly in accordance with claim 12 wherein said at least one stabilizing member is coupled to at least one of said fluid conductors by a welded support clip. claim 12 14. A core spray sparger assembly in accordance with claim 1 wherein each said fluid conductor comprises at least one connection section to facilitate differential thermal expansion between said at least one coolant manifold and each said fluid conductor. claim 1 15. A core spray sparger assembly in accordance with claim 1 wherein said plurality of nozzles comprise a plurality of orifices formed in said fluid conductors. claim 1 16. A core spray sparger assembly in accordance with claim 1 wherein said plurality of nozzles are formed in said fluid conductors such that each fuel assembly is supplied coolant from at least one nozzle. claim 1 17. A nuclear reactor comprising: a plurality of fuel assemblies; at least one coolant supply pipe; a top core guide above side fuel assemblies; a shroud head above said top core guide; and a core spray sparger assembly comprising: at least one coolant manifold, each said coolant manifold comprising a plurality of independent coolant manifolds joined by a plurality of resilient couplings; at least one coolant coupling in fluid communication with said coolant manifold; at least one mounting device coupling said coolant manifold to the nuclear reactor; a plurality of fluid conductors in a parallel array positioned above said top guide, in fluid communication with said coolant manifold; and a plurality of nozzles in fluid communication with said fluid conductors. 18. A nuclear reactor in accordance with claim 17 wherein said at least one coolant manifold comprises two fluidically independent, coaxial, substantially circular, coolant manifolds, joined by a plurality of resilient couplings. claim 17 19. A nuclear reactor in accordance with claim 17 wherein at least one of said coolant manifolds comprises at least one alignment guide configured to align said at least one coolant manifold with said at least one coolant supply pipe. claim 17 20. A nuclear reactor in accordance with claim 17 wherein said plurality of fluid conductors comprise at least two parallel arrays of fluidically independent, parallel fluid conductors, each parallel array in fluid communication with one of said coolant manifolds. claim 17 21. A nuclear reactor in accordance with claim 20 wherein said plurality of nozzles are formed in each said parallel array such that each parallel array supplies coolant to each fuel assembly. claim 20 22. A nuclear reactor in accordance with claim 20 wherein each of said fuel assemblies receives coolant from at least one nozzle in each parallel array. claim 20 23. A nuclear reactor in accordance with claim 20 wherein each of said plurality of nozzles are formed in each said parallel array to supply coolant to at least one fuel assembly. claim 20 24. A nuclear reactor in accordance with claim 17 wherein each of said fluid conductors are a spacing distance apart, said spacing distance greater than a width of one of said plurality of fuel assembly. claim 17 25. A nuclear reactor in accordance with claim 17 wherein said at least one mounting device comprises a plurality of adjustable spring retainers securing at least one coolant manifold between said top guide and said shroud head. claim 17 26. A nuclear reactor in accordance with claim 17 wherein said at least one mounting device comprises a plurality of hanger bolts securing at least one said coolant manifold to said shroud head. claim 17 27. A nuclear reactor in accordance with claim 26 wherein said at least one mounting device further comprises a trunnion. claim 26 28. A nuclear reactor in accordance with claim 17 further comprising at least one alignment device aligning said fluid conductors with said fuel assemblies. claim 17 29. A nuclear reactor in accordance with claim 28 wherein said at least one alignment device comprises at least one alignment channel, each said alignment channel extending from said fluid conductors, said alignment channels engaging said reactor top guide. claim 28 30. A nuclear reactor in accordance with claim 17 further comprising at least one stabilizing member coupled to at least one of said coolant manifolds and at least one of said fluid conductors. claim 17 31. A nuclear reactor in accordance with claim 30 wherein said at least one stabilizing member is coupled to at least one of said fluid conductors by a welded support clip. claim 30 32. A nuclear reactor in accordance with claim 17 each said fluid conductor comprises at least one connection section configured to facilitate differential thermal expansion between said at least one coolant manifold and each said fluid conductor. claim 17 33. A nuclear reactor in accordance with claim 17 wherein said plurality of nozzles comprise a plurality of orifices formed in said fluid conductors. claim 17 34. A nuclear reactor in accordance with claim 17 wherein said plurality of nozzles are formed in said fluid conductors such that at least one nozzle supplies coolant to each said fuel assembly. claim 17
	description	The present application is based on and claims priority to Japanese Patent Application No. JP 2016-151201 filed Aug. 1, 2016, the entire contents of which are incorporated herein by reference. Embodiments relate to a small nuclear reactor used, for example, in space, on the moon, and polar regions of the earth. Small nuclear reactors produce more energy per unit weight than other types of nuclear reactors. The small reactors have been used as power supplies in space for example. The small nuclear reactors include heat pipes which transmit heat generated in the core of the reactor to other parts in the reactor. The heat pipes transmit a large amount of heat per unit volume and do not need movable parts. The heat pipes with such a simple structure have realized simplification of the small nuclear reactors. The heat pipes without movable parts do not induce problems due to movable parts. Such heat pipes have improved reliability of the small nuclear reactors. In small nuclear reactors, a plurality of heat pipes with small diameter can be allocated in the cores. Present embodiments disclose a nuclear reactor including a heat pipe, a first fuel allocated around a side surface of the heat pipe parallel to a central axis of the heat pipe, the first fuel containing a fissile material as a first concentration, a second fuel allocated on an outer side of the first fuel and containing the fissile material at a second concentration smaller than the first concentration, and a core including a plurality of heat pipes arranged in parallel to each of the central axis in the first fuel or the first fuel and the second fuel. Hereafter, embodiments are described with reference to the drawings. An overview of a small nuclear reactor is described with reference to FIG. 1 and FIG. 2. The small nuclear reactor 1 shown in FIG. 1 is a small nuclear reactor used as a power generator, for example, in space, on the moon, on Mars, at polar regions on the earth, etc. In a small nuclear reactor, the acceptable highest temperature in a core is predetermined based on heat resistance of the reactor's structure. For example, in a small nuclear reactor including metal hydride, the acceptable highest temperature in a core is less than the hydrogen dissociation temperature to prevent hydrogen dissociation. Heat generated near the heat pipes transfers to the heat pipes easily and contributes to energy output of the reactor. On the other hand, heat generated relatively farther from the heat pipes is more difficult to transfer to the heat pipes and contributes less to energy output of the reactor. As a result, the temperature farther from the heat pipes in the core is higher than at the other parts in the core nearer the heat pipes. To maintain the temperature at parts farther from the heat pipes to be cooler than the predetermined temperature, the core temperature as a whole needs to be lowered. This means the energy outputted by the reactor has been restricted to maintain the temperature farther from the heat pipe cooler than the predetermined temperature. In the small nuclear reactor device 1 described in this embodiment, a rise in local temperature is prevented and the output energy from the reactor is improved. As shown in FIG. 1, the small nuclear reactor device 1 includes a nuclear reactor 2 containing nuclear fuel, a plurality of heat pipes 3 transferring heat generated by fission of the nuclear fuel in the nuclear reactor 2 to the outside of the nuclear reactor 2, a shielding 4 blocking radiation from inside the nuclear reactor 2, a power generating section 5 converting heat transmitted through the heat pipes 3 to electricity, and a plurality of radiators 6 dissipating the remaining heat of the power generating section 5. In a small nuclear reactor device 1 expected to be used in space, the shielding 4 can be between the nuclear reactor 2 and the power generating section 5, and thereby radiation from the nuclear reactor 2 does not reach any component part behind the shielding 4 from the nuclear reactor 2. The arrangement of the shielding 4 is not limited just to between the nuclear reactor 2 and the power generating section 5. In a small nuclear reactor device 1 expected to be used on earth, the shielding 4 can alternatively cover the entire circumference of the nuclear reactor 2. The nuclear reactor 2 further includes, for example, a container 7 covering the nuclear fuel and the heat pipes 3, and (not shown) a moderator decelerating neutrons and a control rod controlling fission reaction. The container 7 may include a neutron reflector reflecting neutrons emitted from nuclear fuel. A structure including nuclear fuel, the heat pipes 3, the moderator, the control rod, and the container 7 may be referred to as a core 32. The heat pipes 3 as a heat removal mechanism are included in the core 32. The function of the core 32 is the same as the reactor 2. A shape of the nuclear reactor 2 and the core 32 is not limited to a cylinder. In some embodiments, the shape of the nuclear reactor 2 and the core 32 are, for example, a cylinder, a rectangular, or a cone. The core 32 includes a plurality of the heat pipes 3 arranged in parallel to each central axis of the fuel containing fissile material. The power generating section 5 includes thermoelectric conversion elements to convert heat transmitted through the heat pipes 3 to electricity. The thermoelectric conversion elements generate electricity by temperature differences occurring in them. The power generating section 5 is not limited to generating electricity with thermoelectric conversion elements. The power generating section 5 could convert heat to electricity with, for example, turbines or a stirling engine. In that case, turbines rotate with steam generated by heat from the heat pipes 3 and generate electricity. The stirling engine is driven by a change of volume of gas sealed in the stirling engine and generates electricity. As shown in FIG. 2, showing one of the heat pipes 3, working fluid transfers heat in the heat pipe 3. The heat pipe 3 includes a pipe case 8 made of materials with high thermal conductivity, a volatile liquid as the working fluid being sealed in the pipe case 8, a vacant space 9, and wick 10 forming a capillary structure on an inside of the pipe case 8. The vaporized working fluid moves in the vacant space 9. The pipe case 8 and the wick 10 can be made of aluminum and copper, for example. The working fluid is, for example, a fluorocarbon substitute. The end of the heat pipe 3 is a high temperature section 11 which is expected to be heated from the outside, and the other end of the heat pipe 3 is a low temperature section 12 which is expected to be cooled from the outside. The cycle of the evaporation at the high temperature section 11 (absorption of latent heat) and condensation at the low temperature section 12 (release of latent heat) of the working fluid effects heat transfer in the heat pipe 3. The following is an example of that cycle in the heat pipe 3. The working fluid is heated at the high temperature section 11. The working fluid absorbs heat and evaporates to gas 13. The gas 13 moves to the low temperature section 12 through the vacant space 9. The gas 13 is cooled at the low temperature section 12. The gas 13 releases heat and condenses to liquid 14. The liquid 14 at the low temperature section 12 moves to the high temperature section 11 through the wick 10 by capillary action. That cycle is executed even if there is no difference in height between the high temperature section 11 and the low temperature section 12 or even if the heat pipe is in a zero gravity or law gravity condition. For example, in space heat transfers from the high temperature section 11 to the low temperature section 12 in the heat pipe 3 due to such a cycle of evaporation and condensation of the working fluid and moving the gas 13 and the liquid 14. High temperature sections 11 of the plurality heat pipes 3 are inserted in the core 32. Low temperature sections 12 of the heat pipes 3 extend linearly from the core 32 in the reactor 2 to end in the power generating section 5. The power generating section 5 converts heat to electricity, and heat generated in the core 32 is transferred to the power generating section 5 through the heat pipes 3. The reactor 2 is further described with reference to FIG. 3. The same configurations as those already described are given the same reference symbols, thereby omitting overlapping description. FIG. 3 is a cross-sectional enlarged view of a cell 20 of the core 32 perpendicular to the central axis of the heat pipe 3. A direction parallel to the central axis of the heat pipe 3 is referred to as the third direction. In FIG. 3 the internal structures of the heat pipe 3 are omitted from illustration. To aid understanding, some hatching may be omitted in each sectional view. Fuel 15 and fuel 16 are around the heat pipe 3. The fuel 15 and the fuel 16 contain fissile material causing a fission reaction and non-fission material not causing a fission reaction. The fissile material is, for example, Uranium 235 (U 235). The non-fission material is, for example, Uranium 238 (U 238). The fissile material is not limited to U 235. In some embodiments, the fissile material can be U233, Pu239, PU241, Am 242, Cm 243 and Cm 245. In some embodiments, the non-fission material can be U233, Pu239, PU241, Am 242, Cm 243 and Cm 245. The fuel 15 and 16 may contain various kinds of materials as a base material, for example, metal, oxide, nitride, carbide, chloride, and fluoride. Fuel with a high fissile material concentration generates heat highly per unit volume by a fission reaction. The fuel 15 contains fissile material at a first concentration. The fuel 16 contains fissile material at a second concentration. The first concentration is higher than the second concentration. Below, the fuel 15 is referred to as the first fuel 15, and the fuel 16 is referred to as the second fuel 16. According FIG. 3, the first fuel 15 is allocated around a side surface of the heat pipe 3 parallel to the central axis of the heat pipe 3. The second fuel 16 is allocated further from the heat pipe 3 than the first fuel 15. For example, the second fuel 16 is allocated on an outer side of the first fuel 15. In FIG. 3, the second fuel 16 is around the first fuel 15. In the core 32, a first area and a second area include the first fuel 15 and the second fuel 16 respectively in FIG. 3. Concentration of the fissile material in the first area is more than that in the second area. Heat generated in the first area is more easily transferred by the heat pipes 3 than heat generated in the second area. In the core 32 the heat pipes 3 are arranged at equal intervals. The shape of the first fuel 15 can be a cylinder with a heat pipe 3 as its central axis. The shape of the second fuel 16 can be a rectangular with the first fuel 15 inserted therein. A pair of the first fuel 15 and the second fuel 16 form the cell 20. A plurality of the cells 20 arranged parallel to the heat pipe 3 form the core 32. In a cross-sectional view of the core 32 perpendicular to the central axis of the core 32, the boundaries of the cells 20 form a grid. The shape of the first fuel 15 and the second fuel 16 is not limited to a cylinder or a rectangular. In embodiments, the first fuel 15 and the second fuel 16 could be shaped as a circle, oblong, triangle, rectangle, or hexagon in a cross-sectional view of the core 32 perpendicular to the central axis of the core 32. The shape of the heat pipe 3 is not limited to a circular tube. The cross-sectional shape of the heat pipe 3 perpendicular to its central axis is not limited to a circle. In embodiments, the cross-sectional shape of the heat pipe 3 perpendicular to its central axis could be, for example, an oval, triangle, quadrangle, or hexagon. Each of the plurality of the heat pipes 3 is not limited to the same shape. In embodiments, the diameter of each heat pipe 3 could be different. In one heat pipe 3, there may be parts with different diameters. During criticality, the first fuel 15 generates heat with higher power density than the second fuel 16. This is because the concentration of fissile material of the first fuel 15 is higher than that of the second fuel 16. Formula (1) is the relationship between the thickness of a plate-like heat conductor and heat conducted by a heat conductor. λ is thermal conductivity of the heat conductor. A is area conducting heat. ΔT is temperature difference in the conductor. l is thickness of the heat conductor. q is heat conducted by the heat conductor. According to the formula (1), ΔT and l are inversely proportional. By decreasing l, more heat can be transferred even if ΔT is limited. q = λ ⁢ ⁢ A ⁢ ⁢ Δ ⁢ ⁢ T l ( 1 ) In FIG. 3, thickness of fuel around the heat pipe 3 in the direction perpendicular to the central axis is l in equation (1). Heat generated in the fuel will be more greatly transferred to the heat pipe 3, in the thickness direction, with a temperature farther from the heat pipe 3 being higher than the temperature nearer the heat pipe 3. Such temperature difference in the thickness direction is ΔT. According to FIG. 3, the first fuel 15 is arranged nearer around the heat pipe 3. This means that the l of the first fuel 15 is limited to a smaller amount in the core 32. _For example, consider the two cases that the thickness of the first fuel 15 is l or l′, in which l′ is longer than l. When the first fuel 15 within l and l′ transfer the same amount of the heat to the heat pipes 3, ΔT of the first fuel 15 within l is smaller than ΔT of the first fuel 15 within l′. Thereby temperature difference within the core 32 is suppressed. As described above, the second fuel 16 generates less heat than the first fuel 15. Thereby, using the second fuel 16 with a lower concentration of fissile material arranged around the first fuel 15 prevents a rising temperature locally even if heat generated in the second fuel 16 is not as easily transferred to the heat pipes 3. Also, the second fuel 16 around the first fuel 15 maintains the ΔT in the first fuel 15. In other words, the first fuel 15 around the heat pipes 3 and the second fuel 16 around the first fuel 15 prevent the temperature farther from the heat pipes 3 in the core 32 from becoming much higher than in other parts in the core 32. Also that results in the core outputting more energy with a smaller temperature difference in the core 32. The thickness of the first fuel 15 is preferably within an appropriate range obtained by preliminary experiments. For example, when l is shorter than the appropriate range length, the distance from the middle of the second fuel 16 to the heat pipes 3 increases. Then heat at the middle of the second fuel 16 is less likely to be transferred to the heat pipe 3 and the temperature in the middle of the second fuel 16 increases. As a result, the temperature difference within the core 32 may increase. The first fuel 15 is arranged as a cylinder around the heat pipe 3 in the shown embodiment. Between the first fuel 15 and the heat pipe 3, there is a clearance or gap 23. The clearance or gap 23 may be in a cylinder shape around the heat pipe 3. That is, the inner diameter of the cylindrical first fuel 15 can be greater than the outer diameter of the heat pipe 3. During operation of the nuclear reactor 2, the volume of the first fuel 15 and the second fuel 16 expand compared with their volume before the nuclear reactor 2 operation. The clearance 23 prevents the core 32 from bursting due to expanding of the first fuel 15 and the second fuel 16. Before the nuclear reactor 2 starts operating, the heat pipe 3 is covered with a metal foil 24. The metal foil 24 is made of metal that melts at the nuclear reactor's 2 operating temperature. For example the metal foil 24 is made of Gallium, sodium, lithium, lead, bismuth, and alloy. The metal foil 24 may include a single metal layer or multiple metal layers. When the metal layer 24 melts at a temperature during operation of the nuclear reactor 2, the melted metal layer 24 fills the clearance or gap 23. The melted metal layer 24 improves heat conduction efficiency from the first fuel 15 to the heat pipe 3. During operation of the nuclear reactor 2, the metal layer 24 is liquid and flexible in the clearance or gap 23. The clearance or gap 23 and the metal layer 24 prevent the core 32 from bursting due to expanding of the fuels 15, 16. The metal layer 24 is not limited to being between the first duel 15 and the heat pipe 3. In some embodiments, the metal layer 24 can be arranged between the first fuel 15 and the second fuel 16. The metal layer 24 may be arranged in other parts in the core 32. The metal layer 24 is solid during constructing and before the nuclear reactor 2 starts operation. Thereby, the metal layer 24 can easily cover the heat pipe 3. The loading of the metal layer 24 can be adjusted by changing the number of wrappings of the metal foil around the heat pipe 3. After the nuclear reactor 2 starting operating, the metal layer 24 becomes a liquid in the clearance or gap 23. The metal layer 24 is not limited to a metal foil before the nuclear reactor 2 starts operating. Before the nuclear reactor's 2 operating, the metal layer 24 may be metal particles or metal powder filled in the clearance or gap 23. An arranged nuclear reactor 2 is described with reference to FIG. 4. It is noted that the same configurations as those described above are given the same reference symbols, thereby omitting overlapping description. FIG. 4 is an enlarged cross section of the core 32 perpendicular to the heat pipes 3. In the enlarged cross section of the core 32, there is a first area 25 and a second area 26. There are more heat pipes 3 per unit area in the first area 25 than those in the second area 26. The second area 26 is arranged around the first area in cross section of the core 32 perpendicular to the heat pipes 3. Heat transmitted to the heat pipes 3 per unit area in the first area 25 is greater than that per unit area in the second area 26, due to the difference of the number of the heat pipes 3 between the first area 25 and the second area 26. In the second area 26, there may be no heat pipes 3. All the heat pipes 3 have the same shape and the same diameter. A distance between two adjacent heat pipes 3 in the first area 25 is smaller than that in the second area 26. Thereby, heat to be transferred by the heat pipes 3 per unit area in the second area 26 is less than that per unit area in the first area 25. Heat generated in the second area 26 is thus less transferred by the heat pipes 3 than in the first area 25. As in FIG. 3, the first area 25 includes the first fuel 15 containing fissile material at a first concentration and the second fuel 16 containing fissile material at a second concentration. The second concentration is less than the first concentration. The distance between the first fuel 15 and a heat pipe 3 is smaller than that between the second fuel 16 and the heat pipe 3. For example, the first fuel 15 is allocated around a side surface of the heat pipe 3 parallel to a central axis of the heat pipe 3, and the second fuel 16 is allocated on an outer side of the first fuel 15. In FIG. 4, the second fuel 16 is around the first fuel 15. The second area 26 includes the third fuel 17 containing fissile material at a third concentration and a fourth fuel 18 containing fissile material at a fourth concentration. The distance between the third fuel 17 and a heat pipe 3 is smaller than that between the fourth fuel 18 and the heat pipe 3. For example, the third fuel 17 is allocated around a side surface of the heat pipe 3 parallel to a central axis of the heat pipe, and the fourth fuel 18 is allocated on an outer side of the third fuel 17. In FIG. 4, the fourth fuel 18 is around the third fuel 17. Concentration of fissile material in the first area 25 is different from that in the second area 26. The second concentration is less than the first concentration. The third concentration is less than the second concentration. The fourth concentration is less than the third concentration. Thereby, concentration of fissile material per unit area in the second area 26 is less than that in the first area 27. Thereby, a temperature rise in the second area 26 is suppressed, even though the heat transferred to the heat pipes 3 per unit area in the second area 26 is less than that per unit area in the first area 25. Fissile material concentration of either the first fuel 15 or the second fuel 16 may be the same fissile material concentration of either the third fuel 17 or the second fuel 16. For example, fissile material concentration of the second fuel 16 may be the same as that of the third fuel 17. Fissile material concentration of the first fuel 15 may be the same as that of the third fuel 17 when the fissile material concentration of the fourth fuel 18 is less than that of the second fuel 16. Fissile material concentration of the second fuel 16 may be the same as that of the fourth fuel 18 when the fissile material concentration of the third fuel 17 is less than that of the first fuel 18. The first fuel 15 is in a cylindrical shape surrounding the heat pipe 3. The first fuel 15 is inserted by the heat pipe 3. The central axis of the first fuel 15 is parallel to that of the heat pipe 3. The second fuel 16 is a quadrangular prism surrounding the first fuel 15. The second fuel 16 is inserted around the first fuel 15. The central axis of the second fuel 16 is parallel to that of the first fuel 15. The third fuel 17 is in a cylindrical shape surrounding the heat pipe 3. The third fuel 17 is inserted by the heat pipe 3. The central axis of the third fuel 17 is parallel to that of the heat pipe 3. The fourth fuel 18 is a quadrangular prism surrounding the third fuel 17. The fourth fuel 18 is inserted around the third fuel 17. The central axis of the fourth fuel 18 is parallel to that of the third fuel 17. The first cells 21 include a pair of the first fuel 15 and the second fuel 16. The second cells 22 include a pair of the third fuel 17 and the fourth fuel 18. In a cross section of the core 32 perpendicular to its central axis, a sectional area of the second cells 22 is larger than that of the first cells 21. For example, in FIG. 4, vertical and horizontal dimensions of the second cells 22 are twice as that of the first cells 21 respectively. The four first cells 21 are allocated around a control rod 19 with each cell facing the control rod 19. The control rod 19 is at the same position as the central axis of the core 32. To control the fission reaction, extracting and inserting of the control rod 19 is controlled. The first area 25 made of the four first cells 21 is a higher importance area. The control rod 19 arranged in the first area 15 as the higher importance area improves absorption efficiency of neutrons by the control rod 19. This means that fewer control rods can control the output from the nuclear reactor 2. Extracting and inserting of the control rod 19 may be controlled by control rod drive mechanisms. The control rod 19 is not limited to a bar. For example, the control rod 19 may be a material that expands with a rise in temperature. The control rod 19 beats into the core 32 by expansion and absorbs neutrons. An embodiment of an arranged nuclear reactor 2 is described with reference to FIG. 5 below. It is noted that the same configurations as those described above are given the same reference symbols, thereby omitting overlapping description. FIG. 5 is an enlarged sectional view of the nuclear reactor 2. This sectional view is perpendicular to the direction parallel to the central axis of the heat pipe 3. The first fuel 15 contains fissile material at a first concentration. The second fuel 16 contains fissile material at a second concentration. The first fuel 15 is allocated around a side surface of the heat pipe 3 parallel to a central axis of the heat pipe 3. The second fuel 16 is allocated farther from the heat pipe 3 than the first fuel 15. A layer 31 includes a plurality of heat pipes 3 parallel to their central axes. The direction in which the heat pipes 3 are aligned in parallel is defined as the first direction. The layer 31 includes the first fuel 15 around the heat pipe 3. The layer 31 also includes the second fuel 16 around the first fuel 15 around each of the heat pipes 3 next to each other in the layer 31. A thickness of the second fuel 16, perpendicular to the first direction and the third direction, is the same as the outer diameter of the first fuel 15 or a little larger than the outer diameter of the first fuel 15. The first heat conductor 27 is along a side surface of the layer 31 and parallel to the third direction. The thermal conductivity of the first heat conductor 27 is larger than that of the second fuel 16. The first layer 31 is between the two first heat conductors 27. For example, the first layer 31 is made of beryllium. In the second fuel 16, most of heat generated near the first fuel 15 is likely to be transferred to the heat pipe 3. In the second fuel 16, most of heat generated farther from the first fuel 15 is harder to be transferred to the heat pipe 3. This means that heat in the middle of the second fuel 16 is hard to have transferred. The middle part of the second fuel 16 is thereby referred to as a low conductive area 28. The first heat conductors 27 are provided to transfer heat in the low conductive area 28 to nearer the first fuel 15. In other word, heat generated in the second fuel 16 is transferred to the first fuel 15 by the first heat conductors 27. And heat in the first fuel 15 is transferred to the heat pipe 3. Thereby a temperature rise in the second fuel 16 is suppressed. The first heat conductor 27 operates as a bypass by transferring heat from the second fuel 16 to the first fuel 15. The first heat conductor 27 made of beryllium has high thermal conductivity and increases neutrons radiated from the fuel 15, 16 to promote the fission reaction. Beryllium included in the first heat conductor 27 increase neutrons by (n, 2n) reaction and improves criticality of the fuel 15, 16. Material included in the first heat conductor 27 is not limited to beryllium. For example, the material may be copper, liquid, or another solid. An embodiment of an arranged nuclear reactor 2 is described with reference to FIG. 6 below. It is noted that the same configurations as those described above are given the same reference symbols, thereby omitting overlapping description. FIG. 6 is an enlarged sectional view of the nuclear reactor 2. This sectional view is perpendicular to the third direction. The first fuel 15 contains fissile material at a first concentration. The second fuel 16 contains fissile material at a second concentration. The first fuel 15 is allocated around a side surface of the heat pipes 3 parallel to a central axis of the heat pipes 3. The second fuel 16 is allocated farther from the heat pipes 3 than the first fuel 15. The first layer 41 and the second layer 42 include a plurality of heat pipes 3 parallel to their central axes, respectively. Each layer 41, 42 includes the first fuel 15 around a heat pipe 3. Each layer 41, 42 also includes the second fuel 16 around the first fuel 15 around each of the heat pipes 3 next to each other in the layer 41, 42. The thickness of the second fuel 16, perpendicular to the first direction and the third direction, is the same as the outer diameter of the first fuel 15 or a little larger than the outer diameter of the first fuel 15. The second layer 42 is stacked on the first layer 41 in the direction perpendicular to the first direction and the third direction. The heat pipe 3 of the second layer 42 is allocated between two heat pipes 3 next to each other in the first layer 41. With this order, as seen from the second direction, the heat pipes 3 in the layers next to each other are prevented from overlapping. Thereby, a local temperature rise in the core 32 is suppressed. The layers 41, 42 are sandwiched between the two first heat conductors 27, respectively. The first heat conductor 27 is in a plate shape. Thermal conductivity of the first heat conductor 27 is higher than that of the second fuel 16. A moderator 43 is provided between the first layer 41 and the second layer 42, the layers 41 and 42 are sandwiched between the first heat conductors 27. The moderator 43 is made of, for example, a solid metal hydride. For example, the moderator 43 includes calcium hydride, zirconium hydride, lanthanum hydride, praseodymium hydride, or graphite. In FIG. 6, the first heat conductor 27 made of beryllium is arranged closer to the fuels 15, 16 than moderator 43. Thereby, neutrons reach the first heat conductor 27 without going through the moderator 43. The first heat conductor 27 increases neutrons by (n, 2n) reaction without an influence of deceleration caused by the moderator 43. A second heat conductor 44 is in the second fuel 16 in the layers 41, 42 and parallel to the first direction. The second heat conductor 44 is arranged between two heat pipes 3 in a layer. A third heat conductor 45 connects two overlapping layers which are in the first layer 41 and the second layer 42. One end of the third heat conductor 45 is closer to the heat pipe 3 than the other end. In other words, the third heat conductor 45 connects the second fuel 16 closer to the heat pipe 3 in the first layer 41 and the low conductive area 28 in the second layer 42. The other third heat conductor 45 connects the second fuel 16 closer to the heat pipe 3 in the second layer 42 and the low conductive area 28 in the first layer 41. The third heat conductor 45 is parallel to the second direction. The heat conductors 44, 45 are a plate or a bar. Heat conductivity of the heat conductors 44, 45 is higher than that of the second fuel 16. Heat conductivity of the heat conductors 44, 45 may be the same as or higher than that of the first heat conductor 27. The heat conductors 44, 45 are made of materials with high thermal conductivity such as beryllium or copper. The heat conductors 44, 45 may be liquid or solid. The heat conductors 44, 45 transfer heat generated in the low conductive area 28 to nearer the first fuel 15. The heat conductors 44, 45 transfer heat generated farther from the heat pipe 3 to nearer the first fuel 15. Thereby, a temperature rise caused by heat generated in the low conductive area 28 is suppressed. Heat 29 generated in the second fuel 16 is transferred to the first fuel 15 through the first heat conductor 27. Heat 29 is transferred to the heat pipe 3 through the first fuel 15. Thereby, a temperature rise in the second fuel 16 is suppressed. An embodiment of an arranged nuclear reactor 2 is described with reference to FIG. 7 to FIG. 16 below. It is noted that the same configurations as those described above are given the same reference symbols, thereby omitting overlapping description. The nuclear reactor 2 and the core 32 are each a cylinder. The core 32 includes a plurality of the heat pipes 3 parallel to the central axis of the core 32. Each of the heat pipe 3 has the same structure. FIG. 7 is a cross-sectional view of a nuclear reactor, perpendicular to the third direction. As shown in FIG. 8 layers 51, 52, 53, 54 are concentrically arranged cylinders that include the heat pipes 3 and the first fuels 15. The third layer 53 is inside of the fourth layer 54. The second layer 52 is inside of the third layer 53. The first layer 51 is inside of the second layer 52. The layers 51, 52, 53, 54 are each in a cylindrical shape. This arrangement of the layers 51-54 improves efficiency to transfer heat. This arrangement of the layers 51-54 also contributes to miniaturization of the nuclear reactor 2. FIG. 8 is a cross-sectional view of a quarter of nuclear reactor 2, perpendicular to the third direction. The control rod 19 is arranged in the central axis of the core 32. The layers 51, 52, 53, 54 are arranged as multiple cylindrical shaped layers concentrically around the control rod 19. The control rod 19 is arranged adjacent the first layer 51. In each of the layers 51, 52, 53, 54, the same number of heat pipes 3 are arranged at equal intervals. Thereby, the number of the heat pipes 3 per unit area decreases with distance from the center of the core 32, so that the number of the heat pipes 3 per unit area of the first layer 51 is the highest in the core 32. The number of the heat pipes 3 per unit area of the fourth layer 54 is the lowest in the core 32. The first layer 51 is defined as a first area. The second layer 52 is defined as a second area. The third layer 53 is defined as a third area. The fourth layer 54 is defined as a fourth area. Heat transferred to the heat pipes 3 per unit area in the first area is more than that of the second area. Heat transferred to the heat pipes 3 per unit area in the second area is more than that of the third area. Heat transferred to the heat pipes 3 per unit area in the third area is more than that of the fourth area. The third area tends to have the highest temperature. In the third area, the heat transferred to the heat pipes 3 per unit area is smaller than other areas Except sometimes, the heat transferred to the heat pipes 3 per unit area in the fourth area is less than that of the third area. However, the temperature in the fourth area is lower than the third area because there is no more fuel outside the fourth layer 54. The moderators 43 are between the first layer 51 and the second layer 52, between the second layer 52 and the third layer 53, and between the third layer 53 and the fourth layer 54. A neutron reflector 55 is on the fourth layer 54. The neutron reflector 55 contains beryllium. The neutron reflector 55 reflects neutrons from the fuel towards the center of the core 32. As shown in FIG. 9 showing a portion around a heat pipe 3 in FIGS. 7-8, the first fuel 15 is allocated around a side surface of the heat pipe 3 parallel to a central axis of the heat pipe 3. The second fuel 16 is allocated outside of the first fuel 15, farther from the heat pipe 3 than the first fuel 15. The first fuel 15 contains fissile material at a first concentration. The second fuel 16 contains fissile material at second concentration. The first concentration is higher than the second concentration. A first direction is a circumferential direction of each layer 51, 52, 53, 54. The second fuel 16 is arranged along the first direction between respective first fuels 15 next to each other in the layers 51, 52, 53, 54. A respective first heat conductor 27 contacts respective side surfaces of the layers 51, 52, 53, 54 parallel to a central axis of a respective heat pipe 3. Each first heat conductor 27 contacts with a first fuel 15 and a second fuel 16. The contact area of the first fuel 15 and the first heat conductor 27 is curved. The contact area is thereby larger than when the contact area is a plane. The larger contact area improves the heat conduction between the first heat conductor 27 and the first fuel 15. The moderator 43 is allocated along the first heat conductor 27. The thickness of the first heat conductor 27 is thinner than that of the moderator 43. The neutrons from the fuels 15, 16 can go through the first heat conductor 27. The first heat conductor 27 nearer the fuels 15, 16 tends to increase the number of neutrons due to reflection. The neutrons generated in a layer are decelerated by the moderator 43 and fission reaction is likely to occur until a neutron reaches another layer. Neutrons hit the first fuel 15 efficiently in the core 32, in which the layers 51, 52, 53, 54 are arranged as multiple cylindrical shaped layers concentrically. As shown in FIG. 8, neutrons 56 from the first layer 51 can hit the first fuel 15 in the second layer 52. And the neutrons 56 from the second layer 52 can hit the first fuel 15 in the third layer 53, or may hit the first fuel 15 in the fourth layer 54. The fissile material concentration of the fuel in a layer is not limited to be the same. The fissile material concentration may be different in the third direction. For example, the first and second fuels 15, 16 in every layers 51, 52, 53, 54 can have a divided plurality of sections 57 in the third direction. In each of the sections 57, concentration in the fuels 15, 16 can be adjusted appropriately. FIGS. 10-11 show how the concentrations of fissile material in the first and second fuels 15, 16 can vary different sections 57 in the height or third direction in each of the different layers 51, 52, 53, 54. The fissile material concentration in the first fuel 15 of each of the sections 57 is described in FIG. 10. The Z axis in FIG. 10 is the third or height direction. The R axis in FIG. 10 is the second direction. The first fuel 15 is allocated in each section 57. According FIG. 10, each fissile material concentration of the first fuel 15 in the sections 57 is 20% or 8%. In the layers 51, 52 and 54, the fissile material concentration of the first fuel 15 in each section 57 is 20%. In the third layer 53, the fissile material concentration of the first fuel 15 is 20% or 8%. In the third layer 53, the heat transferred to the heat pipe 3 per unit area is different in the third direction. The heat transferred to the heat pipe 3 per unit area in the middle of the core 32 may be lower than in the other parts. The heat transferred to the heat pipe 3 per unit area in the third layer 53 may be lower than in the other parts. The fissile material concentration in the first fuel 15 of the center of the core 32 in the third direction, the shaded parts in FIG. 10, can be lower than the other parts. Due to this allocation of the material, the difference of the temperature in the core 32 can be further suppressed. The concentration of the first fuel 15 in the first layer 51, the second layer 52, the third layer 53 and fourth layer 54 is L11, L12, L13 and L14, respectively. The average concentration of the first fuel 15 can then be L11=L12=L14>L13, or L11≥L12≥L14>L13, or L11>L12>L14>L13, or L11>L12>L13>L14. The fissile material concentration in the second fuel 16 of each of the sections 57 is described in FIG. 11. The Z axis in FIG. 10 is the third or height direction. The R axis in FIG. 10 is the second direction. The second fuel 16 is allocated in each section 57. As shown in FIG. 11, each fissile material concentration in the second fuel 16 in the sections 57 is 20%, 15%, 8%, 5% or 3%. The heat transferred to the heat pipe 3 in the core 32 differs in the second direction and the third direction. For example, the heat transferred to the heat pipe 3 in the middle of the third direction in the core 32 is lower than in the other parts. Especially the heat transferred to the heat pipe 3 in the third layer 53 is lower than in the other parts. The fissile material concentration in the second fuel 16 of the center of the core 32 in the third direction, the shaded parts in FIG. 11, can be lower than the other parts. Due to this allocation of the material, the difference of the temperature in the core 32 can be further suppressed. The fissile material concentration of the second fuel 16 in the first layer 51, the second layer 52, the third layer 53 and fourth layer 54 is L21, L22, L23 and L24, respectively. The average concentration of the first fuel 15 can then be L21>L22=L24>L13, or L21>L22≥L24>L23, or L21>L22>L24>L23, or L21>L22>L23>L24. The average fissile material concentration of the first fuel 15 and the second fuel 16 in the layers 51, 52, 53 and 54 are described as the first average concentration, the second average concentration, the third average concentration, and the fourth average concentration, respectively. The average concentration in each layer is not limited to be the same with all the other layers. For example, the second average concentration can be lower than the first average concentration, the fourth average concentration can be lower than the second average concentration, and the third average concentration can be lower than the fourth average concentration. The fissile material concentration in the fuels 15 and 16 is arranged in the third direction and the second direction. Due to this allocation of the fuel, the difference of the temperature in the core 32 can be further suppressed. The heat generated per unit time in each section 57 is described with reference to FIGS. 12 and 13. The unit of each numerical value is W/cm3. FIG. 12 indicates the heat generated in the first fuel 15 in each section 57. FIG. 12 corresponds to FIG. 10. FIG. 13 indicates the heat generated in the second fuel 16 in each section 57. FIG. 13 corresponds to FIG. 11. FIG. 14 indicates heat generated in the first fuel 15 in each section 57 when the fissile material concentration in the first fuel 15 in each section 57 is the same with the other sections 57. As shown in FIG. 14, the fissile material concentration in the first fuel 15 in every section 57 is 20%. FIG. 15 indicates heat generated in the first fuel 16 in each section 57 when the fissile material concentration in the first fuel 16 in each section 57 is the same with the other sections 57. As shown in FIG. 15, the fissile material concentration in the first fuel 16 in every section 57 is 15%. As shown in FIG. 14 and FIG. 15, heat in the middle in the third direction in each layer 51-54, the shaded parts in FIG. 14 and FIG. 15, is higher than in the other parts. Especially, heat generated in the center of the core 32 in the third direction and the second direction is the highest in the core 32. Heat distribution of the core 32 is a cosine distribution centered on the center of the core 32. Thereby, the temperature on the center of the core 32 is much higher than in the other parts. As shown in FIG. 12, the section 57 generating the most heat in the second layer 52 does not overlap the section 57 generating the most heat in the third layer 53 in the second direction. The section 57 generating the most heat in the third layer 53 does not overlap the section 57 generating the most heat in the fourth layer 54 in the second direction. These arrangements prevent the core 32 from increasing local temperature differences. As shown in FIG. 13, the section 57 generating the most heat in the first layer 51 does not overlap the section 57 generating the most heat in the second layer 52 in the second direction. The section 57 generating the most heat in the second layer 52 does not overlap the section 57 generating the most heat in the third layer 53 in the second direction. The section 57 generating the most heat in the third layer 53 does not overlap the section 57 generating the most heat in the fourth layer 54 in the second direction. These arrangements prevent the core 32 from increasing local temperature differences. A method of heat transferring from inside of the core 32 to outside of the core 32 is described below. FIG. 16 is a flowchart of the method. In the flow chart, for example, “step 11” is written as “S11”. At first, the first fuel 15 generates heat in S11. Heat generated in the first fuel 15 is transferred to the heat pipes 3 directly at S12 as a first transferring operation. The second fuel 16 generates heat in S13. Heat generated in the second fuel 16 is transferred to the first fuel 15 at S14. Heat generated in the second fuel 16 is transferred to the heat pipe 3 through the first fuel 15 at S15 as the second transferring operation. Heat transferred to the heat pipe 3 is transferred to the power generating section 5 at S16 as a transferring operation. The power generating section 5 generates electric power from heat from the heat pipe 3 at S17. Excess heat after S17 is transferred to the radiator 6 to be released to the atmosphere at S18. Although some embodiments have been described, these embodiments are presented by way of examples and are not intended to limit the scope of the embodiments. These novel embodiments can be carried out in other various forms, and various omissions, replacements, and modifications can be made thereto without departing from the spirit thereof. These embodiments and their variations shall be included in the scope and spirit of the inventions, and also in the range of inventions and its equivalents recited in the claims of the patent. For example, the metal foil 24 may cover the heat pipes 3. The second heat conductor 44 and the third heat conductor 45 may be arranged into any of the cores 32 described in this specification. The heat pipe 3 is not limited to having liquid inside. A heat pipe 3 that does not have an inside cavity may be used. Heat pumps may be used instead of the heat pipes 3. The first direction, the second direction, and the third direction may be the X axis, Y axis and Z axis, respectively. The first direction, the second direction, and the third direction may be the circumferential direction, the radial direction, and the axial direction of the cylinder, respectively. The section perpendicular to the central axis of the core 32 is not limited to a circle. The section perpendicular to the central axis of the core 32 may be an ellipse or oval. Obviously, numerous modifications and variations of the embodiments are possible in light of the above teachings. It is therefore to be understood that within the scope of the appended claims, the embodiments may be practiced otherwise than as specifically described herein.
	summary
	summary
050135210	summary	FIELD OF THE INVENTION The invention relates to an internal shell of a fast-neutron nuclear reactor, with an improved structure. BACKGROUND OF THE INVENTION Fast-neutron nuclear reactors cooled by liquid metal generally comprise a main vessel of large dimensions enclosing the liquid cooling metal, most often sodium, in which the reactor core is immersed. Inside the main vessel there are arranged internal structures supporting the core and enabling the sodium circulating inside the vessel to be channelled and the internal volume of the vessel to be separated into different parts where the liquid sodium is at different temperatures. One part of these structures, which comprises an ogival toric shoulder surmounted by a substantially cylindrical sleeve, forms an internal shell separating the internal volume of the main vessel above the core support into a hot collector situated inside the internal shell and a cold collector situated outside the internal shell. The components of the reactor, such as the pumps for circulating the liquid sodium and the intermediate heat exchangers in the case of an integrated type nuclear reactor, are immersed in the liquid sodium filling the main vessel and pass with their bottom part through the internal shell in the region of the shoulder. The top part of main vessel is closed by a plate which also supports the vessel and the components. The liquid metal in the main vessel has an upper free level above which there is an inert gas such as argon, inside a space situated underneath the closure plate. This upper free level of the liquid sodium is essentially variable during operation of the reactor; it is in fact capable of moving between two well-defined positions inside the internal shell which will be referred to below as the high level and low level. The level is, on the other hand, constant inside the main vessel of the reactor owing to the presence of a run-off assembly. The top end of the internal shell is located above the high level of the sodium, such that this shell permanently ensures separation of the internal volume of the main vessel into two zones. The top part of this shell is therefore subjected to a high axial temperature gradient due to the presence of the liquid sodium/inert gas interface, both during permanent-state operation of the reactor and during transient operational states which are accompanied by variations in the free level of the liquid sodium. This axial temperature gradient is accompanied by thermomechanical stresses in the top part of the internal shell, such that it is necessary to monitor the operating performance of this shell and avoid operational states accompanied by excessively rapid displacements of the free level. In particular, when the reactor is started up, it is necessary to limit the rise in temperature of the sodium, and this increases the duration of these start-up operations. The inert gas located above the liquid sodium is in fact at a temperature which constantly remains well below the temperature of the liquid sodium. The emersed part of the internal shell is therefore much colder than the immersed part. Furthermore, in order to limit the effect of any stresses of seismic origin on the internal shell, it is necessary to fix on the top part of this shell a reinforcing structure which generally consists of an annular part welded to the top end of the shell. The presence of this reinforcement accentuates the thermal inertia phenomena and increases the axial temperature gradient during transient states. FR-A-2,532,629 and GB-A-1,431,371 describe devices for reducing the thermal stresses in the wall of a vessel such as the external vessel of a nuclear reactor cooled by liquid sodium, consisting of an annular enclosure which is supplied with hot liquid metal either via its bottom part or via its top part, by annexed means. In this way, the part of the wall of the vessel in contact with the hot liquid metal introduced into the enclosure is kept at a fixed temperature whatever the variations in the level of the liquid metal inside the vessel. Such a device, which requires annexed means for supplying the annular enclosure with hot liquid metal, cannot be applied to the case of the internal shell of a fast-neutron nuclear reactor and is not capable of limiting the thermal stresses in the wall, during all stages of use of the reactor vessel. In fact, protection is not ensured when the reactor is shut down. SUMMARY OF THE INVENTION The object of the invention therefore, is to propose an internal shell contained in the main vessel of a fast-neutron nuclear reactor enclosing liquid metal for cooling the reactor, the upper free level of which is capable of moving inside this internal shell, during normal operation of the reactor, between two defined positions respectively referred to as "high level " on the one hand and low level. The internal shell inside the main vessel comprises at least one cylindrical sleeve with a vertical axis, the top part of which is located above the high level of the liquid metal inside the main vessel, and an annular enclosure on the internal periphery of the cylindrical sleeve, open upwards at its top part and delimited by a secondary sleeve arranged substantially coaxially and inside the cylindrical sleeve, and by an annular base fixed to the bottom end of the secondary sleeve and onto the internal surface of the cylindrical sleeve, below the low level of the liquid metal, this internal shell having a structure enabling the thermomechanical stresses to be limited considerably as a result of the annular enclosure, without using annexed means for supplying liquid metal, and in all situations in which the reactor is used. For this purpose, the top end of the secondary sleeve is located below the high level of the liquid metal. According to a preferred embodiment, the annular base of the enclosure is made solid so as to form a reinforcement for the cylindrical sleeve of the internal shell, enabling it to withstand the stresses accompanying a tremor.
	abstract	Embodiments of the invention provide a method of reducing thermal energy accumulation during a plasma ion implantation process for forming patterns including magnetic and non-magnetic domains on a magnetically susceptible surface on a substrate. In one embodiment, a method of controlling a substrate temperature during a plasma ion implantation process includes (a) performing a first portion of a plasma ion implantation process on a substrate having a magnetically susceptible layer formed thereon in a processing chamber for a first time period, wherein a temperature of the substrate is maintained below about 150 degrees Celsius, (b) cooling the temperature of the substrate after the first portion of the plasma ion implantation process has been completed, and (c) performing a second portion of the plasma ion implantation process on the substrate, wherein the temperature of the substrate is maintained below 150 degrees Celsius.
046719261	claims	1. A nuclear fuel assembly comprising a plurality of vertical fuel rods, a top tie plate arranged at the upper ends of the fuel rods, a fuel channel formed with approximately square cross-section and surrounding said fuel rods and said top tie plate, and a fixing member of stainless steel fixed to the fuel channel, said top tie plate having a vertical, upwardly-directed projection disposed in a corner portion of said fuel channel and arranged in a tensile force-transmitting mechanical connection with said fixing member via a vertical bolt which is passed through a through-hole provided in said fixing member, said bolt being made of a material whose coefficient of thermal expansion is smaller than the coefficient of thermal expansion of said stainless steel, and wherein said bolt is surrounded by a sleeve means comprising a metallic sleeve member, a first engaging means for compressive engagement between said sleeve member and said fixing member, a second engaging means for compressive engagement between said sleeve member and said projection, and a third engaging means for compressive engagement between said sleeve member and said bolt, said second and third engaging means providing substantially no play between said bolt, said sleeve member and said projection, and said first engaging means providing such play between said fixing member and said projection that said metallic sleeve which is loaded with a compressive force corresponding to a major part of the tensile force acting on said bolt, said sleeve being made of a metallic material whose coefficient of thermal expansion is smaller than the coefficient of thermal expansion of said stainless steel. 2. Fuel assembly according to claim 1, in which said sleeve and said bolt are made of the same material. 3. Fuel assembly according to one of the preceding claims, in which said sleeve is arranged in said through-hole provided in said fixing portion. 4. Fuel assembly according to claim 1, in which said sleeve has a vertical extension which constitutes at least 40% of the vertical extension of a part of said bolt which is loaded with a constant tensile force. 5. A fuel assembly according to claim 1, in which said sleeve has a vertical extension which constitutes at least 60% of the vertical extension of a part of said bolt which is loaded with a constant tensile force. 6. A fuel assembly according to claim 1, in which the coefficient of thermal expansion of said sleeve is equal to or smaller than the coefficient of thermal expansion of said bolt. 7. A fuel assembly according to claim 1, in which the play between said fixing member and said projection comprises a gap that is not reduced to 0 upon heating the fuel assembly to operating temperature. 8. A fuel assembly according to claim 1, in which the thermal expansion coefficient of said bolt is less than a thermal expansion coefficient of said fixing member and the thermal expansion coefficient of said sleeve is less than the thermal expansion coefficient of said bolt, and in which the difference between the coefficients of said bolt and said fixing member is substantially smaller than the difference between the coefficients of said sleeve and said bolt. 9. A fuel assembly according to claim 8, in which there is substantially no gap between said fixing member and said projection.
	claims	1. Beam blanker for interrupting a beam of charged particles, the beam blanker having an axis along which charged particles propagate before entering the beam blanker, the beam blanker comprising means for generating an electric field perpendicular to said axis, the electric field for deflecting the charged particles, and an aperture in a diaphragm, the aperture transmitting the beam when the beam is not interrupted and the diaphragm stopping the beam when the beam is interrupted, characterized in that the electric field is generated by a resonant structure with a resonant frequency f, the resonant structure equipped to generate an electric field that sweeps the beam over the aperture, the resonant structure comprising a resonant transmission line and a grounded conductor, the electric field generated between the two resonant transmission line and the grounded conductor, as a result of which the beam is transmitted through the aperture twice per period of the frequency f. 2. Beam blanker for interrupting a beam of charged particles, the beam blanker having an axis along which charged particles propagate before entering the beam blanker, the beam blanker comprising means for generating an electric field perpendicular to said axis, the electric field for deflecting the charged particles, and an aperture in a diaphragm, the aperture transmitting the beam when the beam is not interrupted and the diaphragm stopping the beam when the beam is interrupted, characterized in that the electric field is generated by a resonant structure with a resonant frequency f, the resonant structure equipped to generate an electric field that sweeps the beam over the aperture, the resonant structure excited by a signal supplied by an electrical circuit comprising a photoreceptor, as a result of which the beam is transmitted through the aperture twice per period of the frequency f. 3. Beam blanker for interrupting a beam of charged particles, the beam blanker having an axis along which charged particles propagate before entering the beam blanker, the beam blanker comprising means for generating an electric field perpendicular to said axis, the electric field for deflecting the charged particles, and an aperture in a diaphragm, the aperture transmitting the beam when the beam is not interrupted and the diaphragm stopping the beam when the beam is interrupted, characterized in that the electric field is generated by a resonant structure with a resonant frequency f, the resonant structure equipped to generate an electric field that sweeps the beam over the aperture, the aperture comprising a slit or a hole with a dimension in the direction in which the beam is deflected of less than 10 μm, as a result of which the beam is transmitted through the aperture twice per period of the frequency f. 4. The beam blanker according to claim 3 in which the resonant frequency f is injection locked, phase locked, or frequency locked to a driving signal, or derived from the driving signal by frequency multiplication. 5. The beam blanker according to claim 3 in which the resonant structure compring an element for tuning the resonant frequency f or shifting the phase of the resonant frequency f. 6. The beam blanker according to claim 3 in which the electric field is generated in a cavity resonator. 7. The beam blanker according to claim 3 equipped to be used in an apparatus with a beam of charged particles in which, in working, the electric field in the direction in which the particles of beam propagate has a length such that the particles traverse the electric field in a time less than 1/10 of a period of the resonant frequency. 8. Particle-optical apparatus comprising a particle source for producing charged particles, one or more charged particle lenses, and a sample position for mounting a sample, the particle-optical apparatus equipped with the beam blanker of claim 3. 9. The particle-optical apparatus of claim 8 further comprising a pulsed laser for producing a train of optical pulses for optically exciting a sample at the sample position, the train of optical pulses synchronized to the resonant frequency f. 10. Beam blanker for interrupting a beam of charged particles, the beam blanker having an axis along which charged particles propagate before entering the beam blanker, the beam blanker comprising means for generating an electric field perpendicular to said axis, the electric field for deflecting the charged particles, and an aperture in a diaphragm, the aperture transmitting the beam when the beam is not interrupted and the diaphragm stopping the beam when the beam is interrupted, characterized in that the electric field is generated by a resonant structure with a resonant frequency f, the resonant structure equipped to generate an electric field that sweeps the beam over the aperture, the resonant structure comprising a negative resistance element, as a result of which the beam is transmitted through the aperture twice per period of the frequency f. 11. Particle-optical apparatus comprising a particle source for producing charged particles, one or more charged particle lenses, and a sample position of mounting a sample, the particle-optical apparatus equipped with:a beam blanker for interrupting a beam of charged particles, the beam blanker having an axis along which charged particles propagate before entering the beam blanker, the beam blanker comprising means for generating an electric field perpendicular to said axis, the electric field for deflecting the charged particles, and an aperture in a diaphragm, the aperture transmitting the beam when the beam is not interrupted and the diaphragm stopping the beam when the beam is interrupted, characterized in that the electric field is generated by a resonant structure with a resonant frequency f, the resonant structure equipped to generate an electric field that sweeps the beam over the aperture, the aperture comprising a slit or a hole with a dimension in the direction in which the beam is deflected of less than 10 μm, as a result of which the beam is transmitted through the aperture twice per period of the frequency f,the beam blanker positioned such that the beam enters the beam blanker with an energy lower than the energy with which the beam impinges on the sample. 12. The particle-optical apparatus of claim 8 in which the beam blanker producing bunches of charged particles, and the apparatus comprises a second beam blanker that in a period of time selectively transmits one or more bunches of charged particles and blocks one or more other bunches of charged particles. 13. Particle-optical apparatus comprising a particle source for producing charged particles, one or more charged particle lenses, and a sample position of mounting a sample, the particle-optical apparatus equipped with:a beam blanker for interrupting a beam of charged particles, the beam blanker having an axis along which charged particles propagate before entering the beam blanker, the beam blanker comprising means for generating an electric field perpendicular to said axis, the electric field for deflecting the charged particles, and an aperture in a diaphragm, the aperture transmitting the beam when the beam is not interrupted and the diaphragm stopping the beam when the beam is interrupted, characterized in that the electric field is generated by a resonant structure with a resonant frequency f, the resonant structure equipped to generate an electric field that sweeps the beam over the aperture, the aperture comprising a slit or a hole with a dimension in the direction in which the beam is deflected of less than 10 μm, as a result of which the beam is transmitted through the aperture twice per period of the frequency f,the beam blanker producing bunches of charged particles, and the apparatus comprising a second beam blanker that in a period of time selectively transmits one or more bunches of charged particles and blocks one or more other bunches of charged particles,the beam blanker and the second beam blanker showing positional overlap and/or sharing components. 14. A particle-optical apparatus having an optical axis along which a beam of the charged particles travel, comprising:a particle source for producing charged particles;one or more charged particle lenses;a beam blanker for interrupting the beam of charged particles, including:a resonator for generating an electric field perpendicular to said axis, the electric field for deflecting the charged particles, andan aperture in a diaphragm, the aperture transmitting the beam when the beam is not interrupted and the diaphragm stopping the beam when the beam is interrupted,the resonant structure comprising a resonant transmission line and a grounded conductor, the electric field generated between the two resonant transmission line and the grounded conductor,the resonator having a resonant frequency f and equipped to generate an electric field that sweeps the beam across a line that is in the place of the aperture, that intersects the aperture, and that is perpendicular to the direction of deflection, and as a result, the beam crosses the line twice per period of the frequency f; anda sample position for mounting a sample in the particle optical apparatus. 15. The particle optical apparatus of claim 14 further comprising a second beam blanker -along the optical axis, the second beam blanker generating a second electric field perpendicular to the first electric field. 16. A particle optical apparatus having an optical axis along which a beam of the charged particles travel, comprising:a particle source for producing charged particles;one or more charged particle lenses;a beam blanker for interrupting the beam of charged particles, including:a resonator for generating an electric field perpendicular to said axis, the electric field for deflecting the charged particles, andan aperture in a diaphragm, the aperture transmitting the beam when the beam is not interrupted and the diaphragm stopping the beam when the beam is interrupted,the resonator having a resonant frequency f and equipped to generate an electric field that sweeps the beam across a line that is in the place of the aperture, that intersects the aperture, and that is perpendicular to the direction of deflection, and as a result, the beam crosses the line twice per period of the frequency f;a sample position for mounting a sample in the particle optical apparatus; anda second beam blanker along the optical axis, the second beam blanker generating a second electric field perpendicular to the first electric field, the second beam blanker sharing an aperture with the first beam blanker. 17. The particle optical apparatus of claim 14 in which the electric field sweeps the beam over the aperture twice per period, as a result of which the beam is transmitted through the aperture twice per period of the frequency f. 18. A particle-optical apparatus having an optical axis along which a beam of the charged particles travel, comprising:a particle source for producing charged particles;one or more charged particle lenses;a beam blanker for interrupting the beam of charged particles, including:a resonator for generating an electric field perpendicular to said axis, the electric field for deflecting the charged particles, andan aperture in a diaphragm, the aperture transmitting the beam when the beam is not interrupted and the diaphragm stopping the beam when the beam is interrupted,the resonator having a resonant frequency f and equipped to generate an electric field that sweeps the beam across a line that is in the place of the aperture, that intersects the aperture, and that is perpendicular to the direction of deflection, and as a result, the beam crosses the line twice per period of the frequency f, the resonant structure is excited by a signal supplied by an electrical circuit comprising a photoreceptor; anda sample position for mounting a sample in the particle optical apparatus. 19. The beam blanker according to claim 14 in which the resonant frequency f is injection locked, phase locked, or frequency locked to a driving signal, or derived from the driving signal by frequency multiplication. 20. A method of rapidly blanking a beam in a particle optical apparatus having a beam blanker including a deflector and an aperture, the beam blanker positioned along a system optical axis, comprising:providing a charged particle beam along the optical axis;providing a deflection signal from a resonator at its resonant frequency to generate an electric field perpendicular to the optical axis, the resonant structure comprising a resonant transmission line and a grounded conductor, the electric field generated between the two resonant transmission line and the grounded conductor, the resonator having a resonant frequency f and generating an electric field that sweeps the beam across a line that is in the plane of an aperture, that intersects the aperture, and that is perpendicular to the direction of deflection, and as a result, the beam crosses the line twice per period of the frequency f, the beam periodically crossing the aperture and being transmitted to a sample;periodically impinging the beam onto a work piece. 21. The method of claim 20 in which providing a deflection signal includes providing a deflection signal that sweeps the beam over the aperture twice per period, as a result of which the beam is transmitted through the aperture twice per period of the frequency f. 22. The beam blanker according to claim 3 in which the aperture is a slit or a hole with a dimension in the direction in which the beam is deflected of less than 1 μm. 23. The beam blanker according to claim 3 equipped to be used in an apparatus with a beam of charged particles in which, in working, the electric field in the direction in which the particles of beam propagate has a length such that the particles traverse the electric field in a time less than 1/100 of a period of the resonant frequency.
050842374	abstract	A grid spacer for a nuclear reactor fuel assembly is formed of two superposed combs. Each comb has a base strip which extends the width of the assembly. To it are attached parallel grid strips which are free at their other ends. The grid strips are perpendicular to the base strip and are formed for spring engagement with the fuel rods of the assembly. The combs can be inserted from the sides of the assembly, so that an additional or substitute spacer can be added to the assembly without disassembling the latter.
	summary
052241440	summary	BACKGROUND OF THE INVENTION The present invention is directed to an improved method and apparatus for imaging objects with penetrating radiation. The use of penetrating radiant energy is well known in the imaging of both inanimate and animate objects. As an example, systems of this type find important use in the inspection of objects to discover secreted contraband which may present a security threat. One type of inspection system uses a flying spot of penetrating radiation to scan the object being inspected. After interaction with the object, the radiant energy may be incident on detector(s) which detect transmitted and/or scattered energy. For example, such a system is disclosed in U.S. Pat. No. Re 28,544, assigned to American Science and Engineering, Inc. This type of system is an improvement over the prior art, in that it allows operation without exposing personnel to excessive radiation. An additional advantage is that since the object may be divided into elemental frontal areas which ar scanned by the flying spot, and the detector signal may be time divided into intervals denoted as pixels which correspond to these elemental areas, the radiant energy which is incident on either a transmission or scatter detector at any given time may be referred to a particular elemental area of the object. In such a flying spot inspection system, the apparatus for producing the flying spot includes a source of penetrating radiation, an absorber plate of highly absorbing material, preferably a high Z material such as lead which has a fixed slit therein, and a chopper wheel, which also must be made of highly absorbing material. The chopper wheel has radially directed slits, and when the chopper wheel is rotated, these slits rotate past the fixed slit in the absorber plate. As such rotation occurs, a flying spot of radiation is created along the direction of the fixed slit in the absorber plate, which is used to scan the object being inspected. Additionally, the system is arranged so that there is relative translation movement between the object and flying spot in a direction perpendicular to the scanning direction of the flying spot, so that the entire object is scanned in successive lines. In the prior art flying spot scanning system, the flying spot is scanned in straight lines, which typically lie in the vertical or horizontal direction. This follows from the fact that the fixed slit in the absorber plate, which defines the scanning direction, is inevitably a straight line in the prior art. With such a system, in order to provide a scan line of adequate length, the radial slits in the chopper wheel, which must maintain coincidence with the projection of the radiation passing through the fixed slit as the wheel rotates, must be of substantial length. While the flying spot scanning system described above has many advantages over the prior art, a problem with it is that because the radial slits must be of substantial length, it is necessary for the chopper wheel to have so much mass, that at the rotation speeds which are required for some applications, it may break apart. Additionally, the high mass of the wheel requires that a powerful motor be used, which may increase the expense of the equipment. By way of example, in a chopper wheel of current design, an annular disk or doughnut of lead which contains the radial slots is embedded in an aluminum wheel. For a typical application, the length of the radial slots must be 14.6 cm, which also defines the radius of the doughnut. Since the thickness of lead required is about 2.15 cm, the resultant weight of the lead doughnut is about 510 lbs (230 Kg.). SUMMARY OF THE INVENTION It is therefore an object of the present invention to provide a method and apparatus of imaging radiation which uses a chopper wheel having reduced mass. It is a further object of the invention to provide a flying spot scanner which uses a chopper wheel which may be rotated at higher speeds. It is still a further object of the invention to provide a flying spot scanner which uses a chopper wheel which may be more easily rotated. In accordance with the invention, the above objects are accomplished by scanning an object with a flying spot of penetrating radiation along curved scan lines. The use of such curved scan lines permits a flying spot scanner of more advantageous structure to be used. After the radiant energy interacts with the object being inspected, it is detected, and a detection signal is provided which is divided into pixels, wherein each pixel corresponds to an elemental frontal area of the object being inspected. The pixels are addressed to a utilization means such as a memory or a display in such manner that the pixels which correspond to a scan line of the object define a curved line in the utilization means which has the same shape as the curved scanning line. In this way, a proper image of the object is presented. In the preferred embodiment of the invention, the object is scanned with a flying spot of penetrating radiation along arcuate scan lines, wherein the term "arcuate" refers herein to curved scanning lines having the shape of a circular arc The pixels representative of the scanning are addressed to a utilization means such that the pixels which correspond to a scan line define an arcuate line in the utilization means. In a flying spot scanner in accordance with a preferred embodiment of the invention, a source of penetrating radiation, an absorber plate having a fixed, arcuate slit, and a chopper wheel are provided. The chopper, wheel has an insert of high Z material in the shape of an annular ring or doughnut, which has relatively short radial slits therein. As the chopper wheel is rotated, the radially directed slits maintain coincidence with the projection of the penetrating radiation which passes through the fixed slit in the absorber plate, to form a flying spot along an arcuate path. Additionally, means for providing relative translation motion between the fixed slit and the object being illuminated is provided, so that the object is scanned in successive, arcuate scanning lines.
050892177	description	DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS Turning in detail to the drawings, where like numbers refer to like items, FIGS. 1 and 2, in combination, represent a schematic flow diagram of one preferred embodiment of the present invention. Other configurations are possible and do not affect the method and apparatus of the present invention. Referring now to FIG. 1, primary system process fluids containing suspended and dissolved solids from the chemical decontamination process are removed from the primary system of a nuclear reactor in fluid flow 10, which includes a means for providing a pressure head, passing out of the containment structure 12 of the nuclear reactor and into the chemical clean-up sub-system. The fluids flow through piping 14 into a back-flushable filter 16. A pressure head needed for operation of the chemical clean-up sub-system is preferably provided in fluid flow 10 by one of the pumps already being used in a reactor auxiliary system. In one preferred embodiment the pressure head is provided by one or more of the residual heat removal system pumps. Further discussion of this aspect is included in co-pending Application Ser. No. 07/62/120. Particulates generated by a standard contamination process will consist of metals (chromium, iron, and nickel) and manganese dioxide. Although the exact quantity of metals will depend upon the crud film thickness, the total quantity will typically be between 400 and 1,000 pounds (180 and 450 Kg). In normal operation of the decontamination system, the majority of this mass will be dissolved by the decontamination chemicals. As for the undissolved particulates, tests have shown that about 70% of the particles will be in the range of 2-8 microns, and their concentration within the process fluids will be in the range of 10-15 parts per million. The manganese dioxide is generated during the alkaline/permanganate step that is common to both of the known CAN-DEREM and LOMI techniques. It is desireable to remove all of this manganese dioxide as particulates, rather than allowing it to become a dissolved solid as the result of its subsequent chemical steps since more solid wastes in the form of spent resin will be generated in removing it as a dissolved solid than would be generated in the form of a particulate slurry. The expected particle size of the manganese dioxide is in the range of 0.7-1.7 microns. Based on the relatively high solids concentration, the large mass of solids would have an adverse effect on downstream resin beds in terms of excessive pressure drop or coating of the resins. Therefore, it is preferable to remove at least a substantial portion of the suspended solids prior to utilization of any ion-exchange, demineralization beds. Thus, a back-flushable filter capable of removing particles larger than about 10 microns is used. The limitation on the particle removal size is based on current filter technology, which indicates that back-wash efficiency is poor with filters rated below 5 or 10 microns. Other ratings are possible without departing from the principle of the present invention. A back-flushable filter 16 can be back-flushed with process fluid, demineralized water, or nitrogen, depending upon the design chosen. In a preferred embodiment, as shown in FIG. 1, nitrogen 18 is provided to a accumulator 20 for use in back-flushing via piping 22 and valve 24. A demineralized water source 26 can also be provided as needed via valve 28. If nitrogen is used to back-flush, a further flush with demineralized water is recommended. If process fluid is used, demineralized water is not necessary. When back-flushing, the back-flushable filter 16, valves 30 and 31 will be closed and valve 32 will be opened to direct the back-flushed material to a filtrate collection tank 34. One or more of these valves may preferably be remotely operated as a motor valve or an air valve to minimize personnel radiation exposure. Demineralized water can also be directed to the filtrate collection tank 34 from the demineralized water source 26 by means of piping 36 and valve 38. At a convenient time, the collected contents of the filtrate collection tank 34 can be removed. When the present clean-up sub-system is utilized in conjunction with the resin processing system described in co-pending Application Ser. No. 07/62/130 entitled "Resin Processing System," and incorporated herein by reference, the contents of the filtrate collection tank 34 can be directed to the spent resin storage tank 40 by means of piping 42 and pump 44. Pump 44 is preferably an air-operated diaphragm pump, which can operate to pump both wet and dry materials at low cost. In operation, the back-flushable filter 16 will typically be back-washed when the pressure drop across it reaches 20-25 psi (1000-1300 mmHg). Because the procedures used in both the CAN-DEREM and LOMI processes extend over several days, it is expected that only a few back-washes will be necessary. Therefore, it is reasonable to size the filtrate collection tank 34 for a single back-wash. One or more replaceable cartridge filters 46 are preferably located downstream of the back-flushable filter 16 to which the process fluids are directed by means of piping 48. At least two cartridge filters 46 are recommended, so that one can be changed while the other, or others, is in service. In the embodiment shown in FIG. 1, four cartridge filters 46 are shown, each having front close-off valves 50 and back close-off valves 52 so that individual cartridge filters 46 can be operated, or maintenance performed thereon, independently of the operation of the other cartridge filters 46. One preferred filter media is polypropylene or glass fiber. Pleated paper is typically not acceptable because the decontamination chemicals of the standard processes will dissolve the paper. The cartridge filters will typically have a nominal one micron rating to allow for finer filtration of suspended solids. The combination of the back-flushable filter 16 and the cartridge filters 46 protect the downstream resin beds from fouling and high pressure drop. After passage through the back-flushable filter 16 and the cartridge filters 46, the processed fluids are directed via piping 54 to one or more banks of demineralizers 56. The demineralizer banks 56 can be selectively chosen by means of valves 58. Additionally, the demineralizer banks 56 can be totally bypassed using bypass piping 60 and valve 62. In a preferred embodiment when used with the CANDEREM chemical decontamination process, three banks of demineralizers 56 are aligned in parallel. Two of the banks would be aligned alternately for the alkaline/permanganate steps and a third bank would contain a smaller vessel or vessels called a Regen bed that would be dedicated to the regeneration step (when 70-80% of the curies will be removed from the primary system). When used in conjunction with the resin processing system described in co-pending Application Ser. No. 07/62/130, the first two banks of demineralizers 56 will require resin replacement while the third, the Regen bed, will not require resin replacement. When operating with the LOMI chemical decontamination process, the same two banks of demineralizers 56 wherein the resin is regenerated during operation can be used. The Regen beds are not required for the LOMI decontamination. Looking now at FIG. 2, which focuses on one of the particular banks of demineralizers 56 that are suitable for replacement of resin during operation, the processed fluids are directed to the bank of demineralizers 56 via piping 54 and valve 58. The bank of demineralizers 56 will contain one or more resin bed tanks 64. The resin bed tanks 64 are uniquely sized and arranged in order to optimize a variety of factors including: total resin volume requirements; cation, anion, or mixed bed resin, depending upon the particular process step; resin bed replacement between process steps; adequate flow rate to achieve proper sub-system clean-up within a viable time period; use of multiple units for operating flexibility and ease of transport; and proper resin loading. A chosen arrangement should preferably not require numerous bed replacements since this would significantly affect the critical path time. The amount of resin loading should allow for sufficient residence time to obtain efficient ion exchange. It is preferable to achieve roughly 99% removal of any chemicals injected within the primary system in less than about 8 hours. Thus, a flow rate in the range of 1,000-1,500 gallons (3800-5700 liters) per minute will be necessary for a system volume of approximately 100,000 gallons (380 cubic meters). Based on all of the above factors, the number of demineralizer banks 56 required in a preferred embodiment for each chemical process was determined as discussed above (three for CAN-DEREM and two for LOMI). Further, in one preferred embodiment as illustrated in FIG. 2 each of the demineralizer banks 56 contains three resin bed tanks 64 sized such that each resin bed tank 64 will only require resin replacement once during chemical decontamination. While alternative arrangements are possible, it is preferable to utilize the resin processing system described in co-pending Application Ser. No. 07/62/130. Such a system provides sluice water 66 when needed through valves 68 to flush out the spent resin from the resin bed tank 64 through valve 70 and to a spent resin collection tank 72. Alternate flow for venting and other purposes, such as initial fluffing of the resin prior to removal, is provided by piping 74 and valve 76. Fresh resin can thereafter be provided to the resin bed tank 64 through valve 78. In normal operation, the process fluids enter through piping 54 and valve 58 and are directed to one or more of the resin bed tanks 64 by use of valve 80. After undergoing ion exchange within the resin bed tanks 64 to remove dissolved solids, the processed fluid is removed via screened outlets 82 and piping 84 through valve 86. Valves 89 and 91 can be used to isolate fluid flow from individual demineralizer banks 56. An alternate line of piping 87 is arranged such that two demineralizer banks 56 can be operated in series with isolation valve 85. This configuration is useful when performing a LOMI-type decontamination process. While the process fluids, after passing through the demineralizer banks 56 can be recycled directly to the primary system, in one preferred embodiment they are first sent through one or more resin fines filters 88 by means of piping 90. The resin fines filters 88 will catch resin fines from the resin bed tanks 64. This is especially preferable if several resin bed changeouts are performed during the course of a full chemical decontamination cycle. In addition, the resin fines filters 88 provide assurance that a resin bed tank 64 will not be accidentally dumped into the primary system by operator error during a resin bed replacement operation. These resin fines filters 88 are typically cartridge filters that are replaceable and, thus, it is preferred that more than one such filter be provided. In FIG. 2, four resin fines filters 88 are depicted, each with front valves 92 and back valves 94 so that individual resin filters 88 can be closed off for replacement and maintenance purposes as well as for proper flow regulation. A filter rating of 25 microns or less is recommended. After passing through the resin fines filters 88, the process fluid flows through isolation valve 96 and returns to the primary system 11 via piping 98 and valve 100. Again, these valves may preferably be remotely operated. Chemicals 102 for the chemical decontamination process can be injected just prior to return of the processed fluid to the primary system as necessary. In a standard 5-step CAN-DEREM decontamination process, the resin replacement steps would be as follows: (1) Regeneration step: a first demineralizer bank 56 containing the Regen beds is aligned for service while the CAN-DEREM chemical is recirculated in the system. After the regeneration step, a second demineralizer bank 56 is aligned for removal of the CAN-DEREM chemical. After depletion, this second demineralizer bank 56 has its resin replaced. The time available to replace the resin within this second demineralizer bank 56 is about 15 hours. (2) After the alkaline/permanganate step, flow is once again aligned through the second demineralizer bank 56 for clean-up. When the resin in this second bank 56 is exhausted, the bank is isolated, and a third demineralizer bank 56 is aligned. During the time that the third demineralizer bank 56 is in service, the resin can be replaced within the second demineralizer bank 56. The time available for this resin replacement is approximately 30 hours. The time thereafter available for the third demineralizer bank 56 resin replacement is 26 hours. (3) Repeat steps (1)-(2). (4 ) Repeat step (2). (5) Repeat step 1 except that there is no need to replace the resin within the second demineralizer bank 56 after the CAN-DEREM chemical clean-up. Alternatively, when the standard LOMI chemical decontamination process is used, as mentioned, only two demineralizer banks 56 are required. The resin replacement steps for such a process would normally occur as follows: (1) After the alkaline/permanganate step, flow is aligned through the first demineralizer bank 56 for clean-up. When the resin is exhausted within this first demineralizer bank 56, this bank is isolated, and a second demineralizer bank 56 is aligned. The first demineralizer bank 56 can be replaced with resin for step (2) below during the time that the second demineralizer bank 56 is in service. The time available for resin replacement in the first demineralizer bank 56 is approximately 7 hours. (2) After the LOMI application, the first demineralizer bank 56, filled with cation resin, and the second demineralizer bank 56, filled with weak base anion resin, are aligned in series. For this reason resin replacement cannot begin until clean-up is completed. Each of the banks is replaced with resin for step (3) below. The time available for replacement of resin in the first demineralizer bank 56 is approximately 9 hours while the time available for replacement in the second demineralizer bank 56 is approximately 13 hours. (3) Repeat steps 1 and 2. The apparatus and methods of the present invention are seen to provide significant advantages. Chemical decontamination fluids of any particular decontamination step can be cleaned-up of substantially all suspended and dissolved solids within a reasonable period of approximately 8 hours. The apparatus can be located outside the containment, thereby providing easier access for removal of solid waste. Further, by utilizing a pressure head provided by the primary system itself, overall costs can be minimized. Thus, a clean-up sub-system of the present invention provides efficient, on-line removal of dissolved and suspended solids generated during decontamination of large volume pressurized water reactor fluid systems. It utilizes known technology in a unique arrangement to provide clean-up in a timely manner to minimize the overall scheduled requirements for large system decontamination. Having thus described the invention, it is to be understood that the invention is not limited to the embodiments set forth herein for purposes of exemplification. It is to be limited only by the scope of the attached claims, including a full range of equivalents to which each claim thereof is entitled.
	claims	1. An ion beam delivery method for delivering, through an beam delivery nozzle, an ion beam extracted from a synchrotron, the method comprising the steps of:rotating a wheel installed in said beam delivery nozzle and having a plurality of stepped portions arranged in the rotating direction and having different thicknesses in an axial direction of the wheel to change energy of said ion beam passing said wheel; andintroducing said ion beam to pass at least a part of said plurality of stepped portions of said wheel and controlling start and stop of extraction of said ion beam from said synchrotron in each of the stepped portions in said at least a part thereof which said ion beam passes, during rotation of said wheel. 2. The ion beam delivery method according to claim 1, wherein the control for the start of extraction of said ion beam from said synchrotron is performed in accordance with a rotational angle of said wheel. 3. The ion beam delivery method according to claim 1, wherein the start and stop of extraction of said ion beam from said synchrotron are performed by controlling start and stop of supply of an RF power to an RF-applying device associated with said synchrotron. 4. The ion beam delivery control method according to claim 1, wherein said start of extraction of said ion beam from said beam generator is controlled in accordance with a rotational angle of said wheel measured by a rotational angle sensor provided in said beam delivery nozzle. 5. The ion beam delivery method according to claim 1, wherein the extraction control is performed in accordance with the rotational angle of said wheel, and the extraction stop control is performed in accordance with a measured value of the dose of said ion beam passing said beam delivery nozzle. 6. The ion beam delivery method according to claim 1, wherein the extraction of said ion beam is stopped when an accumulated value of the measured dose of said ion beam passing said beam delivery nozzle has reached a target dose. 7. The ion beam delivery method according to claim 1, wherein control of performing the start and stop of extraction of said ion beam from said synchrotron includes control for start of the extraction of said ion beam in at least the part of said plurality of stepped portions and control for stop of the extraction of said ion beam, the latter control being performed in accordance with a measured dose value of said ion beam passing said beam delivery nozzle. 8. An ion beam delivery equipment for delivering a ion beam to an irradiation target, the equipment comprising:a synchrotron for accelerating said ion beam;an beam delivery nozzle including a wheel having a thickness varied in the rotating direction to change energy of said ion beam passing said wheel, and delivering said ion beam having passed said wheel to said irradiation target; anda controller for controlling start and stop of extraction of said ion beam from said synchrotron during rotation of said wheel,said wheel having a plurality of stepped portions arranged in the rotating direction and having different thicknesses in an axial direction of the wheel,said controller introduces said ion beam to pass at least a part of said plurality of stepped portions of said wheel and controls the start and stop of extraction of said ion beam from said synchrotron in each of the stepped portions in said at least a part thereof which said ion beam passes. 9. The ion beam delivery equipment according to claim 8, wherein said synchrotron includes an RF-applying device, and said controller controls the start and stop of extraction of said ion beam from said synchrotron and controls start and stop of supply of an RF power to said RF-applying device. 10. The ion beam delivery equipment according to claim 8, further comprising a dose monitor for detecting a dose of said ion beam passing said beam delivery nozzle,wherein said controller controls the extraction of said ion beam from said synchrotron in the part of said plurality of stepped portions in accordance with the rotational angle of said wheel, and stops the extraction of said ion beam from said synchrotron in one of the part of said plurality of stepped portions when a dose value detected by said dose monitor has reached a target dose for the one stepped portion. 11. The ion beam delivery equipment according to claim 8, further comprising a dose monitor for detecting a dose of said ion beam passing said beam delivery nozzle,wherein said controller stops the extraction of said ion beam from said synchrotron when an accumulated dose value detected by said dose monitor has reached a target dose. 12. The ion beam delivery equipment according to claim 8, wherein saidcontroller controls said start of extraction of said ion beam from said synchrotron in accordance with a rotational angle of said wheel. 13. The ion beam delivery equipment according to claim 12, wherein said synchrotron includes an RF-applying device, and said controller controls the start and stop of extraction of said ion beam from said synchrotron and controls start and stop of supply of an RF power to said RF-applying device. 14. The ion beam delivery equipment according to claim 12, wherein said beam delivery nozzle includes a rotational angle sensor for detecting the rotational angle of said wheel, and said controller controls the start of extraction of said ion beam in accordance with the rotational angle measured by said rotational angle sensor. 15. The ion beam delivery equipment according to claim 14, wherein said synchrotron includes an RF-applying device, and said controller controls the start and stop of extraction of said ion beam from said synchrotron and controls start and stop of supply of an RF power to said RF-applying device.
	description	This application is the U.S. National Phase under 35 U.S.C. §371 of International Application No. PCT/JP2010/058229, filed on May 14, 2010, which in turn claims the benefit of Japanese Application No. 2009-131218, filed on May 29, 2009, the disclosures of which Applications are incorporated by reference herein. The present invention relates to a liquid metal ion source. The invention relates, for example, to a technique for ion emission from a Ga liquid metal ion source. A liquid metal ion gun includes a liquid metal ion source (hereinafter referred to as LMIS) arranged in a vacuum container. The liquid metal ion gun emits an ion beam corresponding to an amount of ion emission (emission current) responsive to voltages by thermally energizing the emitter (electrode) of the LMIS through a high-voltage cable or applying a high voltage (extraction voltage) across parts of an extracting electrode. The ion beam that is emitted from the LMIS is received by a beam limiting aperture, which limits the emission current and spreading of the ion beam that will propagate toward the downstream side of the beam limiting aperture. For the aforementioned liquid metal ion gun, for example, a gallium liquid metal ion gun, emission stability is maintained when the balance between the amount of gallium consumed by an emitter for ion irradiation purpose and the amount of gallium supplied to the emitter from a reservoir is kept, i.e., when the amount of consumed gallium is in equilibrium with the amount of supplied gallium. However, if a foreign material mixes with gallium, then such a foreign material functions to inhibit gallium from being supplied, or the purity of the gallium is changed and the liquid metal suffers from changes in physical properties. As a result, the amount of consumed gallium is not in equilibrium with the amount of supplied gallium, which will lead to reduction in emission stability. One example of the foreign material that will be mixed in the liquid metal of the LMIS is particles (sputtered particles) generated while the beam limiting aperture is irradiated and sputtered with the ion beam. A conventional technique for suppressing the mixing of the sputtered particles in the liquid metal of the LMIS is described in Patent Document 1. The technique described in Patent Document 1 is as follows. A degradation in stability of an ion current is suppressed in such a manner that a part that is included in the beam limiting aperture and irradiated with the ion beam is constituted by a sintered body having low-melting-point metal mixed therein and the sputtered particles are melted in the liquid metal so as not to form a slag. In addition, the following technique is also known. In the technique, metal that is the same type as the liquid metal of the LMIS is used for a protective diaphragm (refer to Patent Document 2 and the like). Furthermore, the following technique is also known. In the technique, the surface or all parts of an emitter electrode are made of molten metal used for the tip of the emitter electrode or made of one or more types of metal constituting an alloy used for the tip of the emitter electrode (refer to Patent Document 3 and the like). In addition, the following technique is also known. In the technique, a beam limiting aperture that includes, as a base material, the same metal as the emitter (electrode) of the LMIS is formed, and accumulated liquid metal that is the same as the LMIS is provided in a region that is irradiated with the ion beam (refer to Patent Document 4 and the like). Furthermore, the following technique is also known. In the technique, the liquid metal of the LMIS is placed in a container having a recessed portion that includes an aperture hole in the lowest surface, while the ion beam passes through the aperture hole (refer to Patent Document 5 and the like). Patent Document 1: Japanese Patent No. 3190395 Patent Document 2: JP-2001-160369-A Patent Document 3: JP-4-14455-B Patent Document 4: JP-2005-174604-A Patent Document 5: JP-2006-79952-A In the conventional techniques described in Patent Documents 1 to 3, however, a constituent material (base material) of the beam limiting aperture is not described. Thus, it is considered that the beam limiting aperture is irradiated with the ion beam so that the constituent material (base material) of the beam limiting aperture is exposed, sputtered particles that are generated from the base material are mixed in the liquid metal of the LMIS, and the emission stability is thereby degraded. In addition, it is preferable that the LMIS be used for 1000 hours or more. As a cumulative time period for use is increased, the emission stability is gradually degraded. It is necessary to restore the state of the emission to a stable state. In the conventional techniques described in Patent Documents 1 to 3, however, the difficulty level of the restoration is not described. Meanwhile, in the conventional techniques described in Patent Documents 4 and 5, a constituent material that is the same as that of the LMIS is used as the base material of the beam limiting aperture, and liquid metal that is the same as the LMIS is used for a part that is irradiated with the ion beam so that an effect of sputtered particles on the emission is suppressed and the emission stability can be easily restored. The lifetime of the liquid metal ion source is increased. However, with the increase in the lifetime of the liquid metal ion source, an increase in the lifetime of the beam limiting aperture has room for improvement. In order to achieve an increase in the lifetime of the beam limiting aperture, it is considered that the amount of the liquid metal (Ga) that is used for the beam limiting aperture is increased, for example. However, when the amount of the liquid metal is simply increased, an opening through which the ion beam passes may be clogged with the liquid metal. In addition, it is considered that the opening has a larger (wider) size in order to prevent the opening from being clogged with the liquid metal (Ga). In this case, an effect of the aperture is reduced, and the increase in the size of the opening runs counter to the purpose of the beam limiting aperture. The present invention has been made in view of the above problems. An object of the present invention is to provide a liquid metal ion gun that is capable of increasing the lifetime of a beam limiting aperture, maintaining a stable emission for a long time period and restoring the state of the emission to a stable state while reproducibility is excellent. In order to accomplish the aforementioned object, according to the present invention, a liquid metal ion gun includes: a liquid metal ion member that is made of a first metal material; a liquid metal ion source that includes a reservoir and a emitter, the reservoir being made of a second metal material and holding the liquid metal ion member, the emitter being made of the second metal material; and a beam limiting aperture that is formed with a liquid metal member made of the first metal material placed on a base made of the second metal material, has an opening enabling an ion beam extracted from the liquid metal ion source to pass therethrough, and limits the diameter of the ion beam, wherein the beam limiting aperture has a structure that causes the liquid metal member to gather into a region located around the opening. According to the present invention, it is possible to increase the lifetime of the beam limiting aperture, keep stable an emission performed by the liquid metal ion gun for a long time period, and easily restore the emission stability. Embodiments of the present invention are described below with reference to the accompanying drawings. FIG. 1 is an outline diagram illustrating the entire configuration of an ion beam device according to a first embodiment of the present invention. Referring to FIG. 1, the ion beam device according to the present embodiment includes a vacuum container 1; a liquid metal ion gun 3 that is arranged in the vacuum container 1 and emits an ion beam 2 to a sample (not illustrated) arranged in the vacuum container 1; a vacuum discharge device 10 that is connected to the vacuum container 1 through a gate valve 11 and evacuates the vacuum container 1; an ion pump 12 that is connected to the vacuum container 1 and evacuates the vacuum container 1; and a high-voltage power supply unit 20 that supplies power to the liquid metal ion gun 3. The liquid metal ion gun 3 includes a liquid metal ion source (hereinafter referred to as LMIS) 31; an extracting electrode 32 that extracts the ion beam (ions) 2 from the liquid metal ion source 31; a beam limiting aperture 33 that is attached to the extracting electrode 32 and limits spreading of the ion beam 2; and an earth electrode 34 that accelerates the ion beam 2 that has passed through the beam limiting aperture 33. The high-voltage power supply unit 20 includes a grounded earth 21; an extraction power supply 28 that applies an extraction voltage between the LMIS 31 and the extracting electrode 32 to extract the ion beam 2 from the LMIS 31; an acceleration power supply 24 that applies an acceleration voltage between the earth 21 and the LMIS 31 to accelerate the ion beam 2; a heating power supply 29 that performs electric heating on the LMIS 31; high-voltage cables 22, 25 that connect the constituent elements of the high-voltage power supply unit 20 to each other; a high-voltage connecting unit 23 that connects the high-voltage cable 22 to the LMIS 31; and a high-voltage connecting unit 26 that connects the high-voltage cable 25 to the extracting electrode 32. The ion beam 2 that is generated by the liquid metal ion source 31 and extracted by the extracting electrode 32 passes through the beam limiting aperture 33 while the spreading of the ion beam 2 is limited by the beam limiting aperture 33. The ion beam 2 that has passed through the beam limiting aperture 33 is accelerated by the earth electrode 34 and propagates toward a downstream side. The ion beam device has the aforementioned configuration and may have a function (not illustrated) of narrowing the ion beam 2 with a lens, controlling a deflection of the ion beam 2 and irradiating the sample with the ion beam 2, and a signal detection system for detecting a signal from the sample irradiated with the ion beam 2, thereby forming a processing observation device. FIG. 2 is a diagram illustrating in detail the configuration of the LMIS 31 included in the liquid metal ion gun 3. Referring to FIG. 2, the LMIS 31 is a gallium liquid metal ion source (hereinafter referred to as Ga LMIS) that includes gallium (hereinafter referred to as Ga). The Ga LMIS 31 includes an emitter 35 that has a conical tip portion (in other words, a needle-shaped tip portion); a reservoir 36 that stores Ga; a filament 37 that performs electric heating (flashing) on the emitter 35 and the Ga of the reservoir 36; a conductive terminal 38 that is connected to the filament 37; and an insulator base 39 that fixes the conductive terminal 38. The emitter 35, the reservoir 36 and the filament 37 are made of tungsten (hereinafter referred to as W), for example. In addition, the reservoir 36 is filled with Ga. The principle of emission operation of the thus-configured LMIS 31 is outlined below. The gradient of an electric field that is inclined in an axial direction is stronger as the electric field is closer to the tip portion of the emitter 35 under the condition that the extraction voltage is applied to the LMIS 31. This is due to the fact that the tip portion of the emitter 35 is conical. Thus, Ga is supplied to the tip portion of the emitter 35 around which a strong electric field is generated due to electric field stress, and the liquid metal that is located in the vicinity of the tip portion of the emitter 35 has a pointed corner and a conical shape. A strong electric field of approximately several volts/Å exists around the corner of the liquid metal that has the conical shape due to the electric field stress. Thus, an ionization potential is reduced and electrons are lost. Therefore, ionization is easily performed. Specifically, when the strong electric field of approximately several volts/Å exists, ions fly in a vacuum and an ion current is generated. For example, when the Ga LMIS 31 is used, it is necessary to supply the liquid metal Ga with an amount corresponding to the amount of Ga ions to be emitted during the aforementioned ion emission. However, since the Ga acts as a fluid, a pressure gradient occurs. Thus, the surface current flow of a continuous fluid occurs. Since the surface current flow is a Poiseuille flow that is caused by surface tension, the amount of the surface current flow is changed due to a change in the surface tension. In addition, pure Ga has an excellent wettability with clean W. When the W has a thin groove, the Ga diffuses through the groove due to a capillary action caused by the surface tension. Thus, as the configuration of the emitter 35, the emitter 35 that is made of W and has a thin groove extending in the axial direction is known. Similarly, the Ga that is stored in the reservoir 36 made of W is diffused through the aforementioned groove due to the capillary action caused by the surface tension and is supplied to a region located near the tip portion of the emitter. When a balance between the amount of Ga consumed by the emitter 35 for the Ga ion emission and the amount of Ga to be supplied to the emitter 35 from the reservoir 36 is maintained, or when an equilibrium between the amount of the consumed Ga and the amount of the supplied Ga is maintained, the emission stability is maintained. However, when the surface tension of the Ga is changed, the amount of the Ga to be supplied is changed and whereby the state of the emission is changed. A basic configuration of the liquid metal ion gun 3 is described below. In the basic configuration of the liquid metal ion gun 3, a general beam limiting aperture 133 is used in the liquid metal ion gun 3 as an example. FIG. 3 is a diagram illustrating the case in which the general beam limiting aperture 133 is used in the extracting electrode 32 of the liquid metal ion gun 3. Referring to FIG. 3, the liquid metal ion gun 3 that has the exemplary basic configuration includes the Ga LMIS 31; the extracting electrode 32 that extracts the ion beam (ions) 2 from the Ga LMIS 31; and the beam limiting aperture 133 that is arranged on the downstream side (side toward which the ion beam 2 propagates) of the extracting electrode 32 and limits spreading of the ion beam 2. The extracting electrode 32 is made of stainless steel and has an opening 40 which faces the emitter 35 of the Ga LMIS 31 and through which the ion beam 2 passes, for example. For example, the opening 40 has a diameter φ of 3 mm and an opening-side wall having a thickness of 1 mm. In addition, the extracting electrode 32 is arranged so that a distance between a tip of the emitter 35 and an upper surface of the extracting electrode 32 is 0.8 mm and the extracting electrode 32 is not directly irradiated with the ion beam 2. The extracting electrode 32 is configured so that parts of the extracting electrode 32 other than the opening 40 of the extracting electrode 32 are not directly irradiated with the ion beam 2 emitted from the Ga LMIS 31. The beam limiting aperture 133 is made of tin (Sn) and has an opening 41 that is located at a location to be irradiated with the ion beam 2 extracted from the Ga LMIS 31, for example. The ion beam 2 passes through the opening 41. For example, the opening 41 has a diameter φ of 0.3 mm and an opening-side wall having a thickness of 3 mm. In addition, the beam limiting aperture 133 is arranged so that a distance between the upper surface of the extracting electrode 32 and an upper surface of the beam limiting aperture 133 is 5 mm. The beam limiting aperture 133 is included in the extracting electrode 32, and a potential of the beam limiting aperture 133 is equal to a potential of the extracting electrode 32. The beam limiting aperture 133 has a function that is different from a function of the extracting electrode 32. The beam limiting aperture 133 has, as a diaphragm, a function of limiting the ion beam 2 that will propagate toward the downstream side thereof. The extracting electrode 32 has a function of causing ions (ion beam) 2 to be emitted by applying the voltage between the emitter 35 of the Ga LMIS 31 and the upper surface of the extracting electrode 32. In the thus-configured liquid metal ion gun 3, the beam limiting aperture 133 is sputtered by the ion beam 2 emitted from the Ga LMIS 31 so that sputtered particles 42 are generated. A speed (processing speed) at which the beam limiting aperture 133 is sputtered depends on an incident angle. The dependence on the incident angle is described below with reference to FIG. 4. FIG. 4 is a diagram illustrating an example of the dependence of a sputtering yield of a monoatomic solid imaging device on the incident angle. The abscissa indicates the incident angle (°), while the ordinate indicates a sputtering yield for each of incident angles, using, as a reference, a sputtering yield that is obtained when the incident angle is 0 degrees. In other words, FIG. 4 is a diagram illustrating the dependence of the sputtering speed at which the beam limiting aperture 133 is sputtered on the incident angle. As illustrated in FIG. 4, regarding the sputtering yield of the monoatomic solid, as the incident angle θ (°) of the ion beam is larger, a collision cascade occurs closer to a surface. As indicated by the Zygmund's theory, for cos−fθ (f=1 to 2), the incident angle θ is increased until the incident angle θ reaches an angle θopt. For an incident angle θ that is larger than θopt, as the incident angle θ is larger, it is more difficult for the ion beam to pass through the surface due to the fact that a collision coefficient is limited by a shielding effect of adjacent atoms present on the surface. When the incident angle θ is further increased, the ion beam to be incident is reflected without providing energy to the solid, and the sputtering yield is rapidly reduced. In addition, when the monoatomic solid is irradiated with the ion beam, the sputtering speed varies due to a variation in a crystal orientation of a crystal grain. Thus, an irregularity is formed on the surface of the monoatomic solid, while the crystal grain is used as an original point. The irregularity is further promoted due to the dependence of the sputtering speed on the incident angle. Thus, the processing speed of the surface of the monoatomic solid is increased. Therefore, the lifetime of the beam limiting aperture is reduced. The beam limiting aperture 33 according to the present embodiment is replaced with the beam limiting aperture 133 of the liquid metal ion gun 3 having the aforementioned exemplary basic configuration. The details are described below with reference to FIGS. 5(a) to 5(c). FIGS. 5(a) to 5(c) are diagrams illustrating the beam limiting aperture 33 according to the present embodiment and are cross-sectional views taken along a plane including an optical axis of the ion beam 2. FIGS. 5(a) to 5(c) illustrate a change in the amount of the liquid metal on the beam limiting aperture 33. Referring to FIGS. 5(a) to 5(c), the beam limiting aperture 33 includes a base 46 having a recessed portion that faces the ion beam 2 extracted from the Ga LMIS 31, and liquid metal 44 arranged on the recessed portion of the base 46. An opening 41, through which the ion beam 2 passes, is provided at the lowest point located on a surface of the recessed portion of the base 46. In other words, the opening 41 is the farthest from the Ga LMIS 31. The base 46 is made of the same material (i.e., W) as the emitter 35 of the Ga LMIS 31. The liquid metal 44 is the liquid metal (i.e., Ga) that is used for the Ga LMIS 31. A groove portion 45 is provided on the recessed portion of the beam limiting aperture 33 so that the groove portion 45 surrounds the opening 41. In addition, a tapered structure is provided between an outer circumferential part of the recessed portion of the beam limiting aperture 33 and the groove portion 45. The position of the tapered structure is higher toward the outer circumferential part from the side of the groove portion 45. Thus, a change in the shape of the surface of the liquid metal Ga 44 due to a change in the amount of the liquid metal Ga 44 is suppressed. The radius (distance between the opening and the outer circumference of the recessed portion) of the recessed portion of the beam limiting aperture 33 is smaller than the length (k−1) of a capillary of the liquid metal Ga 44. The length (k−1) of the capillary is expressed by the following equation, where the surface tension of the liquid metal 44 is γ, the concentration of the liquid metal 44 is ρ, and a gravity acceleration is g. [ Equation ⁢ ⁢ 1 ] ⁢ k - 1 = γ ρ · g ( Equation ⁢ ⁢ 1 ) In a region that is separated from the opening by a distance range that is larger than the length (k−1) (expressed by the aforementioned Equation 1) of the capillary, the gravity is dominant for the shape of the surface (interface) of the liquid metal Ga 44. In a region that is not separated from the opening by the distance range that is larger than the length (k−1) of the capillary, the surface tension is dominant for the shape of the surface (interface) of the liquid metal Ga 44. The length (k−1) of the capillary of the liquid metal Ga is approximately 3.5 mm. When Ga is placed as the liquid metal 44 on the thus-configured beam limiting aperture 33 and the surface of the W is clean, the recessed portion of the beam limiting aperture 33 completely becomes wet with the Ga. This is due to the fact that Ga has excellent wettability with clean W. The liquid metal Ga 44 is accumulated and the surface of the liquid metal Ga 44 is spherical so that a side wall of a head portion of the liquid metal Ga 44 is an outer circumference. Thus, as the liquid metal Ga 44 is farther from the opening 41 of the beam limiting aperture 33, the position of the surface of the liquid metal Ga is higher. This is due to the fact that in the region that is separated from the opening by the distance range larger than the length (k−1) (the length of the capillary of the gallium is approximately 3.5 mm) of the capillary, the gravity is dominant so that the surface of the liquid metal is a flat surface, and that in the region that is not separated from the opening by the distance range larger than the length (k−1) of the capillary, the surface tension is dominant so that the surface of the liquid metal Ga 44 is a curved surface. In addition, the shape of the surface of the liquid metal Ga 44 is not affected by a structure that has the liquid metal 44 placed thereon and is located in the region that is separated from the opening by the distance range that is not larger than the length (k−1) of the capillary. In other words, the shape of the surface of the liquid metal Ga 44 is not affected by an effect of the groove portion 45. Thus, the shape of the surface of the liquid metal Ga 44 does not depend on the amount of the liquid metal Ga 44. For the case in which the amount of the liquid metal Ga 44 placed on the recessed portion of the beam limiting aperture 33 is small, as the liquid metal Ga 44 is farther from the opening 41, the position of the surface of the liquid metal Ga 44 is higher, as illustrated in FIGS. 5(b) and 5(c). Thus, a part at which the groove portion 45 that is located on an outer circumferential portion of the opening 41 of the beam limiting aperture 33 is formed, is covered with the liquid metal Ga 44 having a thickness that is equal to or larger than the depth of the groove portion 45. In addition, the beam limiting aperture 33 is configured so that a bottom surface of the recessed portion of the beam limiting aperture 33 is smaller than an upper surface of the recessed portion (or a cross-sectional shape is trapezoidal) and an outer circumference of the groove portion 45 is in a range corresponding to the length of a side part of the recessed portion. Thus, as illustrated in FIG. 5(b), even when the amount of the liquid metal Ga 44 is reduced, the shape of the surface of the liquid metal Ga 44 is not significantly changed. In addition, as illustrated in FIG. 5(c), even when the amount of the liquid metal Ga 44 is further reduced, the liquid metal Ga 44 flows from the side part of the recessed portion onto the groove portion 45 due to the surface tension of the liquid metal Ga 44 to fill in the region. Thus, the surface of the liquid metal Ga 44 is maintained flat. Next, how easy the opening 41 becomes clogged with the liquid metal Ga 44 is considered from the perspective of internal pressure of the liquid metal Ga 44. The surface tension and internal pressure (Laplace pressure) of a liquid such as liquid metal are balanced so that surface energy that is given by surface tension (and is not given by the gravity) is minimized. FIG. 6 is a diagram illustrating the state in which the surface tension and internal pressure (Laplace pressure) of the liquid metal Ga 44 located around the opening 41 are balanced. In FIG. 6, the internal pressure ΔP is expressed by the following Equation 2, where a curvature of a curved surface relating to a surface parallel to the opening 41 and a curvature of a curved surface relating to a surface perpendicular to the opening 41 are r1 and r2, respectively. [ Equation ⁢ ⁢ 2 ] ⁢ Δ ⁢ ⁢ P = γ · ( 1 r 1 + 1 r 2 ) ( Equation ⁢ ⁢ 2 ) In the aforementioned Equation 2, it is assumed that a direction in which the internal pressure ΔP (Laplace pressure) is applied from the side of the surface of the liquid metal 44 to the inner side of the liquid metal 44 in the opening 41 is a positive direction. In addition, it is assumed that when the original points of the curvature radii r1 and r2 exist on the side of an inner surface of the liquid metal, the curvature radii are positive, and that when the original points of the curvature radii r1 and r2 exist on the side of an outer surface of the liquid metal, the curvature radii are negative. In FIG. 6, r1<0, and 0<r2. For example, when the amount of the liquid metal Ga 44 is increased, the thickness of the opening 41 becomes nearly equal to the diameter of the opening and θE (contact angle) is large, the Laplace pressure immediately becomes negative (θP<0) and the opening 41 becomes clogged with the liquid metal Ga 44. As described above, in the region that is separated from the opening by the distance range that is larger than the length (k−1) of the capillary, the gravity is dominant for the shape of the surface (interface) of the liquid metal Ga 44. In the region that is separated from the opening by the distance range that is not larger than the length (k−1) of the capillary, the surface tension is dominant for the shape of the surface of the liquid metal Ga 44. The length (k−1) of the capillary of the liquid metal Ga is approximately 3.5 mm. For example, when an inner diameter φ of the recessed portion of the beam limiting aperture 33 is set to a value that is equal to or smaller than 7 mm (or a distance between the opening and the outer circumference is set to a value that is equal to or smaller than 3.5 mm), the liquid metal Ga 44 is placed, and a curvature of the surface of the liquid metal Ga 44 is equal to or smaller than 3.5 mm, the surface tension of the liquid metal Ga 44 is larger than the gravity and an effect of the gravity is small. However, when the liquid metal Ga 44 is placed on the beam limiting aperture 33 so that the curvature of the surface of the liquid metal Ga 44 is equal to or larger than 3.5 mm, the effect of the gravity is large. Especially, since a curvature of the liquid metal Ga 44 in the vicinity of the opening 41 is small, the liquid metal Ga 44 in the vicinity of the opening 41 has negative surface tension. Thus, the opening 41 easily becomes clogged with the liquid metal Ga 44. A shock that is applied from the outside to the liquid metal Ga 44 in the vicinity of the opening 41 is several G. In order to prevent the opening 41 from being clogged due to a movement (caused by the shock) of the liquid metal Ga 44, it is necessary that the curvature r1 (shown in the Equation 2) according to the amount of the liquid metal Ga 44 of the beam limiting aperture 33 be several times smaller than the length (k−1) of the capillary in consideration of the fact that the length (k−1) of the capillary is proportional to a value (1/√g) (g: impulsive force). An effect of the thus-configured present embodiment is described below with reference to FIGS. 7(a) and 7(b). FIGS. 7(a) and 7(b) are diagrams illustrating a beam limiting aperture 933 according to a comparative example with respect to the first embodiment. In FIGS. 7(a) and 7(b), parts that are the same as the parts illustrated in FIGS. 5(a) to 5(c) are indicated by the same reference numerals. In the comparative example illustrated in FIGS. 7(a) and 7(b), the beam limiting aperture 933 includes a base 946 having a recessed portion that faces the ion beam 2 extracted from the Ga LMIS 31, and liquid metal 44 that is arranged on the recessed portion of the base 946. In addition, an opening 41 through which the ion beam 2 passes is provided on a bottom surface of the recessed portion of the base 946. The base 946 is made of W and the liquid metal 44 is Ga. In the thus-configured beam limiting aperture 933 according to the comparative example, when the amount of the liquid metal Ga 44 placed on the recessed portion is sufficient, particles that are generated by sputtering by the ion beam 2 are Ga particles. Even when the sputtered particles are attached to the emitter 35 of the Ga LMIS 31, the sputtered particles do not cause a change in a property of the Ga of the emitter 35. Thus, it is possible to suppress a reduction in the stability of the emission from the Ga LMIS 31. However, when the amount of the liquid metal Ga 44 placed on the recessed portion of the beam limiting aperture 933 is reduced due to an emission performed for a long time period, the tungsten (W) that is a base material is exposed so that sputtered tungsten particles are generated. When the W is attached to the emitter 35, the supply of the Ga from the reservoir 36 to the emitter 35 is prevented so that the emission from the Ga LMIS 31 is reduced. Especially, when the sample is to be irradiated with the ion beam 2 of a large current and a high current density, the diameter of the beam is reduced by reducing the distance between the Ga LMIS 31 and the beam limiting aperture 933. Thus, the amount of sputtered particles (W) attached to the emitter 35 increases and the reduction in the emission is more noticeable. This is due to the following fact. When the distance between the Ga LMIS 31 and the beam limiting aperture 933 is reduced, the area of a region that is included in the beam limiting aperture 933 and irradiated with the ion beam 2 is reduced and an irradiation current density is increased. Thus, the number of sputtered particles to be generated per unit area is increased. Thus, the amount of sputtered particles attached to the emitter 35 is increased, and a viewing angle (solid angle to be formed by sputtered particles capable of being attached to the Ga LMIS 31, and the angle is expressed by dividing the area by the square of the distance) of the Ga LMIS 31 is increased. Therefore, the amount of the sputtered particles attached to the emitter 35 increases. When the rate of the reduction in the emission is increased, and the emission is controlled to be constant by changing the extraction voltage on the basis of the reduction in the emission, the optical axis and a focus are changed. When the extraction voltage is not controlled and the reduction in the emission is not controlled, the beam current of the ion beam 2 is reduced and a region to be irradiated cannot be processed. In addition, the reduction in the emission due to the attachment of the sputtered particles (W) to the emitter 35 can be eliminated by flashing when necessary so that the original emission is achieved. However, when a base material of the beam limiting aperture 933 is sputtered, sputtered W particles are generated. In this case, the opening through which the ion beam 2 passes becomes spread, and the lifetime of the beam limiting aperture 933 is reduced. In addition, it can be considered that the exposure of the W that is the base material is suppressed by increasing the amount of the liquid metal 44 (Ga) to be placed on the recessed portion of the beam limiting aperture 933. However, when the amount of Ga is simply increased, the opening through which the ion beam passes may become clogged with the Ga. In addition, it can be considered that the size of the opening is increased to prevent the opening from becoming clogged with the Ga. In this case, an effect of reducing the diameter of the ion beam is reduced, and the increase in the size of the opening runs counter to the purpose of the beam limiting aperture. In contrast, in the present embodiment, the beam limiting aperture has the following structure: the liquid metal gallium (Ga) is placed on the base 46 made of tungsten (W); the opening that enables the ion beam 2 extracted from the Ga LMIS 31 to pass therethrough is provided; the recessed portion is provided on the base 46 of the beam limiting aperture 33 that limits the diameter of the ion beam 2 so that the recessed portion includes the opening; the groove portion 45 is provided on the inner side of the recessed portion and surrounds the opening 41; and the liquid metal Ga is gathered into a region located around the opening 41. Thus, the liquid metal Ga gathers into the region located around the opening of the beam limiting aperture 33. The thickness of the liquid metal Ga located in the vicinity of the opening is thereby large. Thus, it is possible to increase a time period from the time when the amount of the liquid metal Ga 44 is reduced by sputtering caused by the irradiation with the ion beam 2 to the time when the tungsten W that is the base material is exposed and starts to be sputtered. Therefore, the lifetime of the beam limiting aperture 33 can be increased. Since the amount of the liquid metal Ga 44 that can be placed on the beam limiting aperture 33 is increased by provision of the groove portion 45, the lifetime of the beam limiting aperture 33 can be further increased. In addition, the distance between the opening 41 of the beam limiting aperture 33 and the outer circumference of the recessed portion is smaller than the length (k−1) of the capillary of the liquid metal Ga 44. Thus, even when the amount of the liquid metal Ga 44 is reduced, the liquid metal Ga 44 flows and gathers into the region located around the opening 41 due to the surface tension of the liquid metal Ga located around the opening 41 and the lifetime of the beam limiting aperture 33 can be increased. In addition, gallium (Ga) is placed as the liquid metal on the beam limiting aperture 33. When the amount of the liquid metal Ga 44 is sufficient, particles that are generated by sputtering by the ion beam are Ga particles. Even when the sputtered particles are attached to the emitter 35 of the Ga LMIS 31, the sputtered particles do not cause a change in the property of the Ga of the emitter 35. Thus, it is possible to suppress a reduction in the stability of the emission from the Ga LMIS 31. Therefore, the emission that is performed by the liquid metal ion gun can be maintained stable for a long time period. For example, when the liquid metal ion gun operates so that the ultimate degree of vacuum in the liquid metal ion gun is 10−7 Pa, the extraction voltage applied between the emitter 35 and the extracting electrode 32 is 7 kV, and an emission current is 2.4 μA, the emission can be continuously maintained stable for a time period of 120 hours without maintenance such as the flashing, emission control and an adjustment of the beam focus. In addition, the base material of the base 64 of the beam limiting aperture 33 is tungsten (W). Thus, when the amount of the liquid metal Ga 44 is reduced, the base 46 is exposed and sputtered by the ion beam 2, and sputtered particles (W) generated by the sputtering are attached to the emitter 35 so as to reduce the emission, the emission can be restored to the original emission by the flashing of the Ga LMIS 31 when necessary. Thus, the emission stability for the liquid metal ion gun can be easily restored. In addition, as a contamination of the Ga LMIS 31, a Ga oxide that is generated by the emission can be considered. As the amount of the Ga oxide is increased, the emission is reduced. However, the state (the emission current, a necessary extraction voltage and the stability) of the emission can be reproducibly restored by performing the flashing (heating) at a temperature of approximately 700° C. for a time period of approximately 30 seconds. In addition, since the base material of the base 64 of the beam limiting aperture 33 is tungsten (W), the speed at which the base is sputtered by the ion beam 2 is lower compared with the case in which the base material of the base 64 is tin (Sn). The diameter of a crystal grain of Sn is in a range of approximately 6 μm to 10 μm, while the diameter of a crystal grain of W is approximately 1 μm and small. Thus, an irregularity is hardly formed on the surface of the W at the time of the irradiation with the ion beam due to a variation in a crystal orientation of the crystal grain, and the sputtering speed is low. Thus, the lifetime of the beam limiting aperture 33 can be increased. If sputtered particles are tin (Sn) and attached to the Ga LMIS 31, a compound of the sputtered particles and liquid metal Ga is generated. When the mass percent concentration of the sputtered particles in the liquid metal Ga is equal to or lower than approximately 10 wt %, the melting point of the compound is equal to or lower than approximately 30° C. Thus, the sputtered particles are melted into the liquid metal Ga. As a result, physical characteristics such as the surface tension and melting point of the liquid metal Ga are changed, and the emission becomes unstable. In contrast, in the present embodiment, since the base material of the base 64 of the beam limiting aperture 33 is tungsten (W), the emission that is performed by the liquid metal ion gun can be maintained stable for a long time period. In the first embodiment, the tapered structure is provided between an outer circumferential portion of the beam limiting aperture 33 and the groove portion 45 so that as the tapered structure is closer to the outer circumferential portion from the groove portion 45, the position of the tapered structure is higher. As illustrated in a modified example (illustrated in FIG. 8) of the first embodiment, the outer circumferential portion and a part that corresponds to the bottom surface of the recessed portion may form a substantially right angle. A second embodiment of the present invention is described below with reference to FIG. 9. In the present embodiment, a beam limiting aperture 233 is provided instead of the beam limiting aperture 33 of the liquid metal ion gun according to the first embodiment. Referring to FIG. 9, the beam limiting aperture 233 includes a plate-like base 246 made of the same material (i.e., W) as the emitter 35 of the Ga LMIS 31, and liquid metal 244 placed on the base 246. The liquid metal 246 is the liquid metal (i.e., Ga) that is used for the Ga LMIS 31. In addition, an opening 241 through which the ion beam 2 passes is provided at a location that is included in the base 246 and irradiated with the ion beam 2. The base 246 has a groove portion 245 that surrounds the opening 241. A lump of liquid metal Ga is placed on the surface of the beam limiting aperture 233 and melted. Then, the beam limiting aperture 233 is brought in a supercooled atmosphere having a temperature of −20° C. to solidify the melted liquid metal Ga placed on the beam limiting aperture 233. The liquid metal 244 can thereby move with the beam limiting aperture 233 in an integrated manner and be easily placed in the liquid metal ion gun. Other configurations are the same as those of the first embodiment. In the thus-configured present embodiment, the liquid metal gallium (Ga) is placed on the base 246 made of tungsten (W) so that the beam limiting aperture 233 is formed. The groove portion 245 is provided and surrounds the opening 241 of the beam limiting aperture 233. The beam limiting aperture 233 has a structure that causes the liquid metal Ga 244 to gather into a region located around the opening 241. Thus, the liquid metal Ga gathers into the region located around the opening of the beam limiting aperture 233. The thickness of the liquid metal Ga in the vicinity of the opening is thereby large. Thus, it is possible to increase a time period from the time when the amount of the liquid metal Ga 244 is reduced by sputtering caused by the irradiation with the ion beam 2 to the time when the tungsten W that is the base material is exposed and starts to be sputtered. Therefore, the lifetime of the beam limiting aperture 233 can be increased. Since the amount of the liquid metal Ga 244 that can be placed on the beam limiting aperture 233 is increased by provision of the groove portion 245, the lifetime of the beam limiting aperture 233 can be further increased. In addition, gallium (Ga) is placed as the liquid metal on the beam limiting aperture 233. Thus, when the amount of the liquid metal 244 is sufficient, particles that are generated by sputtering by the ion beam 2 are Ga particles. Even when the sputtered particles are attached to the emitter 35 of the Ga LMIS 31, the sputtered particles do not cause a change in the property of the Ga of the emitter 35. Thus, it is possible to suppress a reduction in the stability of the emission from the Ga LMIS 31. Therefore, the emission that is performed by the liquid metal ion gun can be maintained stable for a long time period. In addition, the base material of the base 264 of the beam limiting aperture 233 is tungsten (W). Thus, even when the amount of the liquid metal Ga 244 is reduced, the base 246 is exposed and sputtered by the ion beam 2, and sputtered particles (W) generated by the sputtering are attached to the emitter 35 so that the emission is reduced, the emission can be restored to the original emission by performing the flashing on the Ga LMIS 31 when necessary. Thus, the emission stability for the liquid metal ion gun can be easily restored. A third embodiment of the present invention is described below with reference to FIG. 10. In the present embodiment, an annular member 347 is provided on the beam limiting aperture 233 of the liquid metal ion gun according to the second embodiment and surrounds the groove portion 245. Referring to FIG. 10, a beam limiting aperture 333 includes a plate-like base 346 made of the same material (i.e., W) as that of the emitter 35 of the Ga LMIS 31, and liquid metal 344 placed on the base 346. The liquid metal 344 is the liquid metal (i.e., Ga) that is used for the Ga LMIS 31. In addition, an opening 341 through which the ion beam 2 passes is provided at a location that is included in the base 346 and irradiated with the ion beam 2. The base 346 has a groove portion 345 that surrounds the opening 341. In addition, the annular member 347 is provided so as to surround the groove portion 345 and not to overlap a region that is irradiated with the ion beam 2. The annular member 347 forms a recessed portion that includes the opening 341. The annular member 347 is made of tungsten or a sintered tungsten body, for example. In the thus-configured present embodiment, an effect that is the same as or similar to the second embodiment can be obtained. A fourth embodiment of the present invention is descried below with reference to FIG. 11. In the present embodiment, a beam limiting aperture 433 is provided instead of the beam limiting aperture 33 of the liquid metal ion gun according to the first embodiment. Referring to FIG. 11, the beam limiting aperture 433 includes a plate-like base 446 made of the same material (i.e., W) as that of the emitter 35 of the Ga LMIS 31, and liquid metal 444 placed on the base 446. The liquid metal 444 is the liquid metal (i.e., Ga) that is used for the Ga LMIS 31. An opening 441 through which the ion beam 2 passes is provided at a location that is included in the base 446 and irradiated with the ion beam 2. The base 446 has a groove portion 445 that surrounds the opening 441. In addition, an annular member 447 surrounds the groove portion 445 and is provided in a region that is not irradiated with the ion beam 2. The annular member 447 forms a recessed portion that includes the opening 441. In addition, a groove portion 449 is formed in an outer circumferential part of the recessed portion in the base 446, or is formed at a boundary between the base 446 and the annular member 447 so as to surround the recessed portion. The annular member 447 is made of a substance (such as an aluminum oxide (Al2O3), molybdenum (Mo) or the like) that does not react with the liquid metal Ga 444 at a temperature of 100° C. or lower. Wettability of tungsten (W) and gallium (Ga) is described here. Tungsten (W) and gallium (Ga) have excellent wettability with the other substance. However, when tungsten (W) is oxidized, tungsten (W) and gallium (Ga) have poor wettability with the other substance. To obtain a clean tungsten (W) surface that is not oxidized, the base 446 made of W is dipped in a sodium hypochlorite solution for one hour. Alternately, after a contamination on the surface is removed by electric field-assisted polishing by use of an electrolyte such as NaOH, ultrasonic cleaning is performed using pure water so that the clean tungsten (W) surface that is not oxidized is obtained. Other configurations are the same as those of the first embodiment. In the present embodiment thus configured, an effect that is the same as or similar to that of the second embodiment can be obtained. A fifth embodiment of the present invention is described below with reference to FIG. 12. In the present embodiment, a beam limiting aperture 533 is provided instead of the beam limiting aperture 33 of the liquid metal ion gun according to the first embodiment. Referring to FIG. 12, the beam limiting aperture 533 includes a plate-like base 546 made of the same material (i.e., W) as that of the emitter 35 of the Ga LMIS 31, and liquid metal 544 placed on the base 546. The liquid metal 544 is the liquid metal (i.e., Ga) that is used for the Ga LMIS 31. In addition, an opening 541 through which the ion beam 2 passes is provided at a location that is included in the base 546 and irradiated with the ion beam 2. Furthermore, an annular member 547 surrounds the opening 541 and is provided in a region that is not irradiated with the ion beam 2. The annular member 547 forms a recessed portion that includes the opening 541. The annular member 547 is made of a material (such as stainless steel (SUS)) that has poor wettability with the liquid metal Ga. In addition, the diameter of the recessed portion, or a distance between the opening 541 and an outer circumference of the recessed portion, is smaller than the length of the capillary of the liquid metal Ga 544. In addition, in order to make the tungsten (W) used for the base 546 and the liquid metal Ga 544 have complete wettability with the other substance, the base 546 is heated in vacuum at a temperature of 1000° C. or higher to completely remove oxygen and a contamination. After that, the gallium (Ga) is placed on the base and then melted. Other configurations are the same as those of the first embodiment. An effect of the thus-configured present embodiment is described below with reference to FIG. 13. FIG. 13 is a diagram illustrating a beam limiting aperture 633 according to a comparative example with respect to the present embodiment. In the comparative example illustrated in FIG. 13, the beam limiting aperture 633 includes a base 646 made of W and liquid metal Ga 644 placed on the base 646. In addition, an opening 641 through which the ion beam 2 passes is provided on the base 646. In the thus-configured beam limiting aperture 633 according to the comparative example, the liquid metal Ga 644 placed on the base 646 spreads on the base 646. Thus, it is difficult to achieve a sufficient thickness of the liquid metal Ga 644 in a region that is irradiated with the ion beam 2. In contrast, in the present embodiment, the beam limiting aperture 533 includes the base 546 made of tungsten (W), and the annular member 547 that is provided on the base 546 so as to surround the opening 541 and form the recessed portion including the opening 541. The annular member 541 is made of the material that has poor wettability with the liquid metal Ga 544. Thus, the annular member 541 does not become wet with the liquid metal Ga 544 placed on the recessed portion of the beam limiting aperture 533, and the liquid metal Ga 544 is repelled from the annular member 541 and gathers into a region located around the opening 541. Since the liquid metal Ga 544 gathers into the region located around the opening 541 of the beam limiting aperture 533, the thickness of the liquid metal Ga 544 in the vicinity of the opening is thereby large. Thus, it is possible to increase a time period from the time when the amount of the liquid metal Ga 544 is reduced by sputtering caused by the irradiation with the ion beam 2 to the time when the tungsten W that is the base material is exposed and starts to be sputtered. Therefore, the lifetime of the beam limiting aperture 533 can therefore be increased. In addition, the distance between the opening 541 of the beam limiting aperture 533 and the outer circumference (i.e., annular member 541) of the recessed portion is smaller than the length (k−1) of the capillary of the liquid metal Ga 544. Thus, even when the amount of the liquid metal Ga 544 is reduced, the liquid metal Ga 544 flows and gathers into the region located around the opening 541 due to the surface tension of the liquid metal Ga 544 located around the opening 541, and the lifetime of the beam limiting aperture 533 can be increased. In addition, gallium (Ga) is placed as the liquid metal on the beam limiting aperture 533. Thus, when the amount of the liquid metal Ga 544 is sufficient, particles that are generated by sputtering by the ion beam 2 are Ga particles. Even when the sputtered particles are attached to the emitter 35 of the Ga LMIS 31, the sputtered particles do not cause a change in the property of the Ga of the emitter 35. Thus, it is possible to suppress a reduction in the stability of the emission from the Ga LMIS 31. Therefore, the emission that is performed by the liquid metal ion gun can be maintained stable for a long time period. In addition, the base material of the base 546 of the beam limiting aperture 533 is tungsten (W). Thus, when the amount of the liquid metal Ga 544 is reduced, the base 546 is exposed and sputtered by the ion beam 2, and sputtered particles (W) generated by the sputtering are attached to the emitter 35 so that the emission is reduced, the emission can be restored to the original emission by performing the flashing on the Ga LMIS 31 when necessary. Thus, the emission stability for the liquid metal ion gun can be easily restored. As a contamination of the Ga LMIS 31, a Ga oxide that is generated by the emission can be considered. As the amount of the Ga oxide is increased, the emission is reduced. However, the state (the emission current, a necessary extraction voltage and the stability) of the emission can be reproducibly restored by performing the flashing (heating) at a temperature of approximately 700° C. for a time period of approximately 30 seconds. The embodiments of the present invention are described above. However, each of the embodiments of the present invention is susceptible to various changes and can be combined with the other embodiments within the spirit of the invention. FIG. 14 is a cross-sectional view of a beam limiting aperture according to a first example and illustrates the example when the present invention is applied. A beam limiting aperture illustrated in FIG. 14 has an outer diameter φ of 10 mm and a thickness of 1 mm. The beam limiting aperture illustrated in FIG. 14 has an upper surface having a diameter φ of 6.85 mm and a bottom surface having a diameter φ of 5 mm, and is tapered 33 degrees from the upper surface to the bottom surface. A distance between the upper surface and the bottom surface of a recessed portion is 0.6 mm. In addition, an opening that has a diameter φ of 0.4 mm is provided at the center of the bottom surface of the recessed portion. Furthermore, the bottom surface has a groove portion having a depth of 0.2 mm. The groove portion is located in a range from a diameter φ of 0.6 mm to a diameter φ of 4 mm from the center of the bottom surface. A back surface of the opening is chamfered by 0.2 mm in a height direction. The amount of the liquid metal Ga that is accumulated on the groove portion of the thus-configured beam limiting aperture is approximately 15 mg, and corresponds to the amount (of the liquid metal Ga) that is reduced by sputtering caused by ion irradiation performed for approximately 1000 hours while an emission current from the Ga LMIS is set to 3.2 μA. In other words, the lifetime of the beam limiting aperture can be increased by approximately 1000 hours compared with a beam limiting aperture that does not have the groove portion. In addition, when the liquid metal Ga with an amount of 60 mg is placed on the beam limiting aperture 733, the shape of an interface of the liquid metal Ga corresponds to the shape of an interface of the liquid metal Ga whose amount is 45 mg and that is placed in the case in which the groove portion is not provided. Thus, the amount of the liquid metal Ga can be increased without changing the state of the liquid metal Ga sputtered by the ion beam 2. When the emission current from the Ga LMIS 31 is 2 μA, and the distance between the Ga LMIS 31 and the beam limiting aperture 733 is 5 mm, a region that is located on the beam limiting aperture 733 and irradiated with the ion beam 2 has a diameter φ of 2.6 mm. Thus, the region to be irradiated with the ion beam 2 is located in the range of up to the diameter φ of 4 mm while the groove portion having the depth of 0.2 mm is provided in the range. The base material is hardly exposed in such a range even when the amount of the liquid metal Ga is reduced. Thus, the lifetime of the beam limiting aperture can be increased compared with the case in which the groove portion is not provided. In addition, the distance between the opening of the beam limiting aperture and the outer circumference of the recessed portion is smaller than the length (k−1≅3.5 mm) of the capillary of the liquid metal Ga. Thus, even when the amount of the liquid metal Ga is reduced, the liquid metal Ga flows and gathers into the region located around the opening due to the surface tension of the liquid metal Ga located around the opening. The lifetime of the beam limiting aperture can therefore be increased. In addition, gallium (Ga) is placed as the liquid metal on the beam limiting aperture. Thus, when the amount of the liquid metal Ga is sufficient, particles that are generated by sputtering by the ion beam are Ga particles. Even when the sputtered particles are attached to the emitter of the Ga LMIS, the sputtered particles do not cause a change in the property of the Ga of the emitter. Thus, it is possible to suppress a reduction in the stability of the emission from the Ga LMIS 31. Therefore, the emission that is performed by the liquid metal ion gun can be maintained stable for a long time period. In addition, the base material of the base of the beam limiting aperture is tungsten (W). Even when the amount of the liquid metal Ga is reduced, the base is exposed and sputtered by the ion beam, and sputtered particles (W) generated by the sputtering are attached to the emitter so that the emission is reduced, the emission can be restored to the original emission by performing the flashing on the Ga LMIS when necessary. Thus, the emission stability for the liquid metal ion gun can be easily restored. 1 Vacuum container 2 Ion beam 3 Liquid metal ion gun 10 Vacuum discharge device 11 Gate valve 12 Ion pump 20 High-voltage power supply unit 21 Earth electrode 22, 25 High-voltage cable 23, 26 High-voltage connecting unit 24 Acceleration power supply 27 Extraction power supply 29 Heating power supply 31 Liquid metal ion source 32 Extracting electrode 33 Beam limiting aperture 34 Earth electrode 35 Emitter 36 Reservoir 37 Filament 38 Conductive terminal 39 Insulator base 40, 41 Opening 42 Sputtered particle
042591520	description	DESCRIPTION OF THE PREFERRED EMBODIMENT Referring now to FIG. 1, there is shown a nuclear reactor vessel 10 including a sealed roof structure 12. Within the vessel 10 is a nuclear core comprised of a plurality of fuel assemblies 14 each having a plurality of sealed fuel rods 16 containing nuclear fuel and a gas plenum. The fuel assemblies 14 are supported upon core support structures including a perforated lower core plate 18 joined to other components such as a transition 20, a cone 22 and a core barrel 24 by welds 26. A liquid coolant, such as sodium, enters the vessel 10 through a plurality of inlet nozzles 28, flows upwardly about the fuel rods 16, absorbing heat energy, and is discharged from the vessel through outlet nozzles 30 to heat exchange apparatus, typically for the ultimate purpose of electric power generation. The coolant then returns to the inlet nozzles 28, completing a circuit through the substantially sealed reactor system. The level 32 of coolant within the vessel 10 is maintained so as to provide a cover gas space 34 within the reactor vessel. Cover gas samples are continuously or intermittently withdrawn from the gas space 34 by a monitoring system 36 through conduit 38 and directed to an analyzer 40 for determining the gas content. The samples can be returned to the reactor system or further treated. As well known, the fuel rods 16 can be provided with tag gases, typically non-radioactive isotopes, which, in the event of fuel rod failure, are released into the circulating coolant along with fission product gases and flow upwardly, ultimately entering the cover gas 34. The analyzer 40, upon monitoring of the cover gas, indicates the presence of a tag gas and accordingly a fuel rod failure. The analyzer 40 in presently proposed reactor systems can be activated by a trigger signal from activator 41 which reacts to the presence of a specified level of radioactive fission product gases released with the tag gases upon a fuel rod failure. A typical tag gas system utilizing especially blended krypton and xenon isotopes is discussed in an article entitled "Design and Manufacture of Gas Tags for FFTF Fuel and Control Assemblies", Nuclear Technology, Vol. 26, August 1975, incorporated herein by reference. In accordance with the invention, the monitoring system 36 can advantageously be utilized to similarly indicate weld failure or crack initiation. FIG. 2 is representative of the welds, for example, the weld 26' between the transition 20 and cone 22. Subsequent to making the weld, a hole or chamber 42 is drilled or otherwise made through selected portions of the component base metal 44, 46 and the weld deposit 48. In the embodiment shown, chamber 42 extends from the surface 50 of the base metal 44, through a portion of the base metal 44, preferably the heat affected zone, through the weld deposit 48, and into the base metal 46. A preselected tag gas 52, different than the fuel rod tag gases, is then injected into the chamber 42 and sealed therein by sealing means such as a plug 54. In order to avoid spurious weld failure indications such as through leakage passed the plug 54, the plug is preferably both threaded into position and additionally welded or otherwise sealed about its periphery. Plural plugs in series can also be utilized for added integrity. Thus, if a failure in the welded area is initiated, such as a crack or a separation along a fusion line 56 which communicates with the chamber 42 and the outer surface of the welded area, the tag gas 52 is released to the surrounding environment 58. In the nuclear reactor system, the surrounding environment is the reactor coolant through which the gas is directed to the cover gas 34 and the monitoring system 36. While fuel rod failures emit not only the selected tag gas but also radioactive fission product gases which trigger the actuator 41, no similar releases of fission product gases arise from weld failure. Accordingly, a specific trigger gas, such as long-lived radioactive krypton-85 can be incorporated in the tag gas to similarly activate the detection system 36. It will be evident that the trigger gas should be fast-acting in actuation of the detection system, since cover gas cleanup systems would otherwise remove the tag gases prior to their detection. The tag gas weld failure detection arrangement can equally be applied to welded structures in other nuclear and also non-nuclear applications, as exemplified in FIG. 3. A weld deposit 48 is disposed between two base metal components 44, 46. A portion of the environment about the welded area, illustrated by the dashed line 60, is directed by drive means such as a pump or fan 62 to an analyzer 40. Tag gas 52 released from the chamber 42 in the event of weld area failure is passed to the analyzer 40 and detected. A trigger gas can also be incorporated with the tag gas to actuate the analyzer 40. FIG. 3 also shows an alternative chamber 42 configuration, the chamber extending from a surface 50 of the base metal 44, through weld deposit 48 and base metal 46 to a surface 64. Two plugs 54 are accordingly utilized to seal the tag gas 52 within chamber 42. The chamber 42 can be of additional geometric configurations and can extend over additional selected areas. A plurality of tag gas filled chambers 42 can also be utilized in conjunction with a single weld deposit. For any application, the diameter of a hole of circular cross section or the size of any other configuration can be selected compatible with the size of the weld, the weld stress conditions and material sensitivity to stress concentration. The pressure of the tag gas placed within the chamber can also be selected in conjunction with the chamber volume to provide a sufficient quantity of tag gas for detection. The spacing of plural chambers along a weld should also be based upon the critical crack size for the specific application. It will be recognized that the presence of a chamber will necessarily create a stress concentration which can reduce the fatigue life of a welded connection. This affect can be reduced by minimizing the size of a chamber and by adjusting the configuration of the chamber in any number of manners. Where, as is often the case, loading or bending stresses are lower in the central area of welded components and higher at the outer surfaces, a configuration such as shown in FIG. 4 can be utilized, the chamber volume being greater in a central region 66 and smaller at outer extensions 68. The enlarged central region 66 is machined into one or both of the base metal members and sealed by plug 70 prior to final machining of the base metal member preparatory to welding. The plug 70 preferably extends beyond the melting zone of the weld. Subsequent to welding, the smaller diameter extensions 68 are drilled from an outer surface and into the region 66. The tag gas can either be sealed within region 66 prior to making of the extensions 68, or the extensions 68 can be utilized for tag gas charging. In a system having a plurality of welds, such as with the reactor vessel and internal components, two or more chambers located in close proximity to each other within a given weld, each filled with a different tag gas, can be utilized to enhance reliability of the system. Evidence of only one of these gases would tend to indicate a spurious signal, such as leakage passed a plug, as opposed to actual weld area failure. Simultaneous detection of the tag gases from separate chambers would tend to indicate a true failure. Where different tag gases are utilized within a single weld, detection of a given tag gas further evidences not only the existence of a weld failure, but also the location of the failure within a specific weld. Additionally, plural tag gas filled chambers can be, and in nuclear applications preferably are, positioned through a welded region at intervals such that a crack which eliminates all of the structural capability of the material between the chambers would still be of a size such that the structural integrity of the weld for its intended function will be maintained. This can be compatibly arranged particularly in liquid metal cooled reactors as major components are typically comprised of austenitic stainless steel which has a relatively long critical crack length. The critical crack length is here defined as the length at which the energy liberated as a result of an incremental crack growth is greater than the energy required to cause that incremental growth such that the crack extends in an unstable manner. Thus, an adequate crack detection system for particularly the lengthy welds in a liquid metal cooled reactor, up to several hundred inches, can be obtained using a relatively small number of tag gas chambers in a given weld. Sealing of the tag gas within the chamber can be accomplished by a number of well known techniques such as sealing of the tag gas while the weld area is disposed in an environment of the desired gas at the desired pressure, or as shown in FIG. 5, by utilization of a container 72, having a rupturable end cap 74 which is pierced by a penetrator 76. Piercing can be accomplished upon insertion of the plug 54 or, for example, by electromagnetically moving the container 72 into contact with the penetrator 76, among other known techniques. Since numerous changes may be made in the above-described apparatus without departing from the spirit and scope thereof, it is intended that all matter contained in the foregoing description and shown in the accompanying drawings shall be interpreted as illustrative and not in a limiting sense.
	abstract	A detector array for detecting X-rays has a number of sensor elements that each have a scintillator element, which is sensitive to X-rays, and a photo-electrical transducer optically coupled thereto. An intermediate areas separating adjacent scintillator elements from one another is present between each two adjacent scintillator elements. Scintillator material is present in the intermediate area. In a production method for such a detector array for detecting X-rays, separating channels are introduced into a layer that is composed of scintillator material, which is sensitive to X-rays, without completely separating the layer.
	summary
	claims	1. A nuclear reactor core of the pressure tube kind in which a plurality of pressure tubes extend through a tank containing, in normal use, heavy water moderator, each of said plurality of pressure tubes comprising:a plurality of fuel elements held together in spaced relationship so as to permit the flow of coolant through spaces between adjacent fuel elements during normal operation, andeach of said fuel elements comprising a plurality of fuel pellets in a cladding comprising sapphire. 2. A nuclear reactor core according to claim 1, wherein the plurality of fuel elements form a bundle, said reactor core further comprising a plurality of support tabs secured to an outermost ring of the bundle of fuel elements to support the bundle in the pressure tube, the tabs comprising sapphire. 3. A nuclear reactor core according to claim 1, wherein each fuel element comprises fuel pellets in tubular cladding closed at each end with end caps. 4. A nuclear reactor core according to claim 3, wherein the tubular cladding and end caps are made of the same material. 5. A nuclear reactor core according to claim 4, wherein the end caps are joined to respective ends of the cladding by an aluminium oxy-nitride bond between respective juxtaposed surfaces of the cladding and end caps. 6. A nuclear reactor core according to claim 1, wherein the sapphire comprises single crystal sapphire. 7. The nuclear reactor core of claim 6 wherein the single crystal sapphire comprises an edge defined film fed growth formation. 8. The nuclear reactor core of claim 1 wherein each of said plurality of pressure tubes comprises a protective insulator sleeve comprising fused silica. 9. The nuclear reactor core according to claim 8, wherein the insulator sleeve has an interior liner or coating of sapphire. 10. A nuclear reactor core in which a plurality of pressure tubes extend through a tank containing, in normal use, heavy water moderator, each of said plurality of pressure tubes comprising: a plurality of fuel elements held together in spaced relationship so as to permit the flow of coolant through spaces between adjacent fuel elements during normal operation, and a protective insulator sleeve comprising fused silica, the fuel elements comprising fuel pellets in a cladding comprising sapphire. 11. A nuclear reactor core according to claim 10, wherein the insulator sleeve has an interior liner or coating comprising sapphire. 12. A nuclear reactor core according to claim 10, wherein the plurality of fuel elements form a bundle, said reactor core further comprising a plurality of support tabs secured to an outermost ring of the bundle of fuel elements to support the bundle in the pressure tube, the tabs comprising sapphire. 13. A nuclear reactor core according to claim 10, wherein the cladding is tubular and is closed at each end with end caps. 14. A nuclear reactor core according to claim 13, wherein the tubular cladding and end caps are made of the same material. 15. A nuclear reactor core according to claim 14, wherein the end caps are joined to respective ends of the cladding by an aluminium oxy-nitride bond between respective juxtaposed surfaces of the cladding and end caps. 16. A nuclear reactor core according to claim 15, wherein the sapphire comprises single crystal sapphire. 17. The nuclear reactor core of claim 16 wherein the single crystal sapphire comprises an edge defined film fed growth formation. 18. A fuel element for use in a nuclear reactor core in which a plurality of fuel elements are held together in spaced relationship so as to permit the flow of coolant through spaces between said plurality of fuel elements during normal operation, the fuel element comprising a plurality of fuel pellets in a cladding comprising sapphire. 19. A fuel element according to claim 18, wherein the cladding comprises tubular cladding closed at each end with end caps. 20. A fuel element according to claim 19, wherein the tubular cladding and end caps are made of the same said material. 21. A fuel element according to claim 20, wherein the end caps are joined to respective ends of the cladding by an aluminium oxy-nitride bond between respective juxtaposed surfaces of the cladding and end caps. 22. A fuel element according to claim 18, wherein the cladding material comprises single crystal sapphire. 23. The fuel element according to claim 22, wherein the single crystal sapphire of the cladding comprises an edge defined film fed growth formation.
	abstract	An anti-scatter grid for an x-ray imaging system. The grid has substrate having a first face and a second face. The substrate has a plurality of grooves opening onto the first face of the substrate and not opening onto the second face. The substrate has high X-ray absorption properties. Each groove has an orientation such that the planes of all the grooves are convergent and intersect along a line situated on the side of the second face where the grooves do not open.
	claims	1. A polymerization apparatus for dental technology which cures a light curing material to be used for a dental prosthesis, the polymerization apparatus for dental technology comprising:a housing;a partition wall disposed in the housing so as to form a polymerization space where polymerization is carried out;a rotating table, which is arranged on a lower side of the polymerization space and on which the dental prosthesis is disposed directly or via other members; anda plurality of LED light sources which irradiate a light for the curing to a polymerization region, which is formed on the upper side of the rotating table and where the dental prosthesis can be disposed,wherein at least one of the plurality of LED light sources is disposed such that a light axis thereof is inclined obliquely upward to be directed toward the polymerization region. 2. The polymerization apparatus for dental technology according to claim 1, which is provided with a plurality of LED light sources disposed such that a light axis thereof is inclined obliquely upward to be directed toward the polymerization region, wherein at least one of the plurality of LED light sources is a purple or near-ultraviolet LED. 3. The polymerization apparatus for dental technology according to claim 1, wherein a height adjustment member is disposed on the rotating table or the rotating table itself is configured to be movable up and down in order to adjust the height position of the dental prosthesis in the polymerization region. 4. The polymerization apparatus for dental technology according to claim 1, wherein at least one of the plurality of LED light sources is disposed such that a light axis thereof is inclined obliquely downward to be directed toward the polymerization region. 5. A polymerization apparatus for dental technology which cures a light curing material to be used for a dental prosthesis, the polymerization apparatus for dental technology comprising:a housing;a partition wall disposed in the housing so as to form a polymerization space where polymerization is carried out, the partition wall having a part formed of a light-transmissive material;a rotating table, which is arranged on a lower side of the polymerization space and on which the dental prosthesis is disposed directly or via other members; anda plurality of LED light sources which irradiate a light for the curing to a polymerization region, which is forme-d on the upper side of the rotating table and where the dental prosthesis can be disposed, the plurality of LED light sources being disposed on opposite side of the partition wall from the polymerization region,wherein at least one of the plurality of LED light sources is disposed such that a light axis thereof is inclined obliquely upward to be directed toward the polymerization region. 6. The polymerization apparatus for dental technology according to claim 5, which is provided with a plurality of LED light sources disposed such that a light axis thereof is inclined obliquely upward to be directed toward the polymerization region, wherein at least one of the plurality of LED light sources is a purple or near-ultraviolet LED. 7. The polymerization apparatus for dental technology according to claim 5, wherein a height adjustment member is disposed on the rotating table or the rotating table itself is configured to be movable up and down in order to adjust the height position of the dental prosthesis in the polymerization region. 8. The polymerization apparatus for dental technology according to claim 5, wherein at least one of the plurality of LED light sources is disposed such that a light axis thereof is inclined obliquely downward to be directed toward the polymerization region. 9. The polymerization apparatus for dental technology according to claim 5, wherein the part formed of a light-transmissive material of the partition wall comprises a lens to at least either diffuse or collect light from the LED light sources. 10. The polymerization apparatus for dental technology according to claim 5, wherein the partition wall is formed of two or more layers having different materials on a side of the polymerization region and the opposite side of the partition wall from the polymerization region.
	abstract	A nuclear fuel assembly having lateral support provided by a bimetallic spring that extends from a side of the fuel assembly under certain core conditions to pressure against an adjacent component and withdraws under other core conditions, such as shutdown, to enable the nuclear fuel assembly to be aligned or withdrawn from the core and repositioned.
043366147	description	DESCRIPTION OF THE PREFERRED EMBODIMENT The tube-in-shell heat exchanger shown in the drawing is for use in a nuclear reactor construction to effect heat exchange between a primary liquid sodium coolant flowing through the tubes and a secondary liquid sodium coolant flowing through the shell. The shell is generally cylindrical and the tubes arranged in a bundle extend substantially parallel to the longitudinal axis of the shell and are secured at their ends to transverse tube sheets. The bundle of tubes has a central tubular spine and to provide transverse support for the heat exchange tubes there is a series of longitudinally spaced grids. Referring now to FIG. 1 there is shown a segment of the heat exchanger wherein is depicted the tubular spine designated 1 and a typical grid 2. With reference now to FIGS. 1 and 2 the grid comprises an inner strap 3 and a coaxial outer strap 4 co-joined by six equally angularly spaced radially extending spider arms or spokes 5. Extending between adjacent spider arms there are arcuate strap members 6 which are arranged to form spaced coaxial intermediate support straps for the heat exchange tubes 7 disposed in substantially coaxial banks. The arcuate strap members 6 carry corrugated strip 8 (as shown in FIG. 3) on each side, the corrugations each serving to cradle a tube 7 and the strips have end fittings for the arcuate strap members welded to them. The arcuate strap members 6 are attached to the spider arms by co-operating end fittings 9 which are fastened together with bolts 10 penetrating the spider arms as shown in FIGS. 4, 5 and 6. The inner and outer banks of tubes 7 are supported from the inner and outer straps 3 and 4 by corrugated strips 8 welded thereto. The corrugated strips are angularly displaced relative to each other, the displacement being in opposite directions for successive grids so as to brace the tubes at points along their lengths thereby to reduce the tendency to vibrate due to fluid flow through the shell. The grids are resiliently supported from the tubular spine 1 each by means of an annular spring 11 co-axially disposed with the spine and engaging with two circular series of six radially extending forked brackets 12, 13 associated with the spine 1 and grid 2 respectively. The spring 11 comprises a stack of relatively slidable annular laminations or leaves designated 11a in FIG. 1. The six forked brackets 12 of a series each have a pair of radially outwardly extending arms and as shown in FIGS. 1 and 7 are welded to an annular stepped collar 14 which is itself welded to the tubular spine. Upper and lower flanges 15 and 16 carried by the collar 14 support sections of an inner wall 17 of an annular shroud 18 for the bundle of tubes. The six forked brackets 13 of a complementary series have radially inwardly extending forked arms which extend through slots in the inner strap 3 and as shown in FIGS. 1 and 2 are welded to recessed inner ends of the spider arms 5. Each forked bracket 13 is interposed between two neighbouring brackets 12. Each outer strip 4 is stepped to receive lower and upper ends of outer walls 19 of the tubular shroud 18 the lower end of each wall being welded to an outer strip 4 whilst the upper end for each wall is slidably fitted to an outer strap. The outer strap at the upper and lower grid plates of the series has an annular spring adapted to bear sealingly on the wall of the shell designated 20. The described construction provides flexibility in the mountings of the tube bundle on the central spine so that groups of heat exchange tubes 7 can be longitudinally displaced relative to the spine and to adjoining groups of tubes to accommodate differential linear thermal expansion. FIG. 8 illustrates a nuclear reactor of the liquid metal cooled fast breeder type having a fuel assembly 21 submerged in a pool 22 of liquid sodium coolant in a primary vessel 23. The primary vessel is suspended from the roof of a containment vault 24 and there is provided a plurality of coolant pumps 25 and heat exchangers 26 of the kind hereinbefore described, only one of each of the pumps and heat exchangers being shown. The fuel assembly 21 is mounted on a diagrid 27 and housed with the heat exchangers in a core tank 28 whilst the pumps, which deliver coolant to the diagrid, are disposed outside of the core tank. The core or fuel assembly 21 comprises a plurality of sub-assemblies which upstand from the diagrid in closely spaced side-by-side array and is surrounded by a neutron shield 29. Control rods 30 and instrumentation 31 are shown penetrating the roof of the vault and passing down towards the fuel assembly through a core cover plate structure 32. The core cover plate structure depends from a rotatable inner plug 33 which is mounted eccentrically in an outer plug 34 rotatable in an aperture in the roof of the vault. The plug 33 carries a fuelling machine 35 which, by rotation of the plugs 33, 34, can be arranged to command any fuelling position of the fuel assembly. In operation of the nuclear reactor relatively cold coolant drawn from the region of the pool which is outside of the core tank 28 is flowed upwardly through the fuel assembly 21 by the pumps 25 by way of the diagrid 27. The coolant flow impinging on the core cover plate is deflected radially to flow into the intermediate heat exchangers 26 through which it passes in heat exchange with a secondary liquid metal coolant associated with steam generating plant disposed externally of the vault.
	summary
	claims	1. A sample container for obtaining an environmental sample, the sample container comprising:an outer chamber having a base region;an outer container wall surrounding said outer chamber, said outer container wall having at least one passage opening formed therein, said at least one passage opening directly fluidically connecting said outer chamber to the environment for filling said outer chamber with a liquid at least in said base region;an inner chamber having a connection for a sampling line and a connection for a conveyor medium line and said inner chamber otherwise being sealed from the environment in a pressure-tight and media-tight manner;an inner container wall surrounding said inner chamber, said inner container wall having a closable passage opening disposed therein forming a fluidic connection between said inner chamber and said base region of said outer chamber, said closable passage opening between said outer chamber and said inner chamber being constructed as a Venturi nozzle; anda pneumatically or hydraulically actuatable closure device for said closable passage opening between said outer chamber and said inner chamber, said closure device having a connection for an actuation medium line. 2. The sample container according to claim 1, wherein said closure device includes a hollow cylinder and a closing piston being moveable in said hollow cylinder between an opening position and a closing position. 3. The sample container according to claim 2, which further comprises a spring element having a spring force and having a resting state setting said closing piston to said opening position, said closing piston moving against said force of said spring element into said closing position in an event of a sufficient pressure in an actuation medium in said actuation medium line. 4. The sample container according to claim 1, wherein said outer container wall has a base plate, and said closure device is disposed in said base plate. 5. The sample container according to claim 1, which further comprises a riser projecting into said inner chamber, said connection for said sampling line being connected to said riser. 6. The sample container according to claim 1, which further comprises at least one heat transfer pipe projecting into said inner chamber for transferring heat from the environment to said inner chamber. 7. A sampling system, comprising:a sample container according to claim 1;a vacuum pump;said sampling line connecting said vacuum pump to said inner chamber of said sample container;a compressed gas source;said conveyor medium line connecting said compressed gas source to said inner chamber of said sample container; andsaid actuation medium line disposed on said closure device of said sample container and connected to said compressed gas source. 8. The sampling system according to claim 7, wherein said actuation medium line branches off from said conveyor medium line permitting a conveyor medium to simultaneously function as an actuation medium for said closure device. 9. The sampling system according to claim 8, wherein the conveyor medium is nitrogen gas. 10. The sampling system according to claim 7, which further comprises a supercritical flow-through choke disposed in said sampling line for maintaining a constant volumetric flow through said sampling line. 11. The sampling system according to claim 7, wherein said vacuum pump is a jet pump. 12. A nuclear power plant, comprising:a safety containment having an interior;a process and analysis module disposed outside said safety containment; anda sampling system including:at least one sample container according to claim 1 disposed in said interior of said safety containment;a vacuum pump;said sampling line connecting said vacuum pump to said inner chamber of said sample container;a compressed gas source;said conveyor medium line connecting said compressed gas source to said inner chamber of said sample container; andsaid actuation medium line disposed on said closure device of said sample container and connected to said compressed gas source. 13. A method for operating a sampling system, the method comprising the following steps:providing a sample container according to claim 1;providing a vacuum pump;using said sampling line to connect said vacuum pump to said inner chamber of said sample container;providing a compressed gas source;using said conveyor medium line to connect said compressed gas source to said inner chamber of said sample container;providing said actuation medium line on said closure device of said sample container and connected to said compressed gas source; andextracting a gas sample by applying a negative pressure to said sampling line upon said closable passage opening between said outer chamber and said inner chamber being open. 14. The method according to claim 13, which further comprises using a scrubbing liquid to convey the gas sample upon a transition of the gas sample from said outer chamber to said inner chamber. 15. The method according to claim 14, which further comprises using said sampling line to feed the scrubbing liquid into said sample container prior to sampling. 16. The method according to claim 13, which further comprises extracting an aerosol sample by firstly conveying a gas sample into the sample container by using a scrubbing liquid and then extracting a liquid sample from the scrubbing liquid. 17. The method according to claim 13, which further comprises applying pressure to said actuation medium line to close said closure device. 18. A method for operating a sampling system, the method comprising the following steps:providing a sample container according to claim 1;providing a vacuum pump;using said sampling line to connect said vacuum pump to said inner chamber of said sample container;providing a compressed gas source;using said conveyor medium line to connect said compressed gas source to said inner chamber of said sample container;providing said actuation medium line on said closure device of said sample container and connected to said compressed gas source; andcompressing a fluid sample through said sampling line by applying an overpressure to said conveyor medium line upon said closable passage opening between said outer chamber and said inner chamber being closed. 19. The method according to claim 18, which further comprises extracting an aerosol sample by firstly conveying a gas sample into the sample container by using a scrubbing liquid and then extracting a liquid sample from the scrubbing liquid. 20. The method according to claim 18, which further comprises applying pressure to said actuation medium line to close said closure device.
049842587	abstract	An apparatus for slit radiography comprising discrete controllable attenuation elements coacting with a slit diaphragm, each attenuation element being affixed to damping means.
046876260	description	DETAILED DESCRIPTION OF THE INVENTION The present invention provides a safety device for dumping steam from a steam generator in a nuclear power reactor, removing residual heat from the steam generator and providing the steam generator with feedwater during emergencies such as steam line rupture and/or power failure. As shown in FIG. 1, reference numeral 10 generally refers to the safety device of the present invention which is located outside the containment wall 12 for steam generator 14 which is connected to a nuclear power reactor (not shown). Steam from generator 14 is fed through valve 16 and containment wall 12 to drive steam turbines (not shown) via main steam line 18. Outside containment wall 12, main steam line 18 is provided with steam safety valve 20. Downsteam from safety valve 20, steam isolation valve 22 is provided on main steam line 18. Between valves 18 and 22, branch pipe 24 feeds steam in main steam line 18 into cold water storage tank 26 through steam relief valve 28. The lower end of line 24 is formed into a nozzle or steam jet 30 for the ejection of high pressure steam into diffuser body 32. Inlet 34 of diffuser body 32 is in the form of an inverted bell for the suction of cold water in storage tank 26 into diffuser body 32. Discharge 34 of diffuser body 32 is in fluid communication with chamber 36 where the condensed steam in the form of a subcooled water is stored. Chamber 36 is provided with exit 38 which is connected to the entrance 40 of a heat exchange device. Typically, the heat exchange device is a cooling coil 42 which provides a large surface area for optimal heat transfer. The outlet 44 of cooling coil 42 is connected to inlet port 46 in storage tank 26. Thus, hot steam condensatae flows from chamber 36 through outlet 38 and inlet 40 into cooling coil 42, leaves the coil through outlet 44 and into storage tank 26. To prevent excessive build-up of cooling water in storage tank 26, a drain pipe controlled by valve 29 is provided in storage tank 26. Cooling coil 42 is submerged in cold water storage tank 50 for heat exchange between the hot condensate within the coil and the cold water stored in pool 46. The volume of coolant water stored in pool 46 should be sufficiently large so as to allow optimal cooling of the coil 42. In a preferred embodiment, pool 46 is in the shape of an inverted L, i.e. one comprising a horizontal portion 48 and vertical portion 50. In FIG. 1, cooling coil 42 is placed within the vertical section in pool 50 which is divided into two connecting chambers by baffle 52. Baffle 52 comprises a long vertical section 51, a short vertical section 53 and a horizontal section 55 connecting the two vertical sections 51 and 53. Baffle 52 is completely under the waterline in pool 46. The baffle forms two chambers 54 and 56 in pool 46 which are in fluid communication. Cooling coil 42 is placed in the vertical portion of chamber 56. In order to permit device 10 to function as an emergency feedwater source to steam generator 14, chamber 36 is optionally connected to main feedwater line 15 to steam generator 14 via pipe 58 and exit 60. Emergency feedwater valve 62 is located downstream from exit 60 in pipe 58. Check valves 64 prevent the flow of feedwater from line 15 into chamber 36. Emergency feedwater valve 62 is opened when the pressure within chamber 36 exceeds a specified limit. In operation, when there is a rupture in the main steam line downstream from steam isolation valve 22, valve 22 is closed to direct the flow of steam into device 10 of the present invention. When the pressure within line 18 exceeds a specific value, steam relief valve 28 opens, thus allowing high pressure steam to flow through nozzle 30 as a strong steam jet. The flow of the steam from nozzle 30 into diffuser body 32 causes cold water in storage tank 26 to be entrained and drawn into diffuser body 32. The steam is condensed and is collected in chamber 36 as subcooled water. As more steam enters diffuser body 32, the condensate is forced into inlet 40 of cooling coil 42 submerged in cooling water pool 48. The condensate flows up coil 42, through outlet 44 and inlet 46 into storage tank 26. Heat in the condensate is dissipated to the cooling water in chamber 56 of pool 54. As a result, when the condensate enters storage tank 26, its temperature is much lower than the condensate in chamber 36. Due to the transfer of heat into chamber 56, the water therein rises as a result of the increase in temperature. Consequently, the water in pool 48 is caused to circulate by flowing upwardly and then horizontally in chamber 56 over short vertical section of baffle 53 into chamber 54 where it flows horizontally and then vertically down past the lower end of long vertical section 51 of baffle 52 into chamber 56. Due to this circulation, the cooling water loses the heat it absorbs from cooling coil 42 so that when the water enters chamber 56, the temperature of the water has dropped significantly and can be used to absorb heat from cooling. coil 42 again. Accordingly, the steam from main steam line 18 can be safely dumped into safety device 10, a closed system, without the risk of harming the surrounding atmosphere. In addition, the residual heat in the steam generator can be dissipated through device 10 which requires no power source to operate, i.e. a passive system. Hence, in the event of a total alternating current power failure, the present safety device 10 offers a back-up cooling system which will operate without power. Valves 20, 22, 24 and 62 can be pressure-operated or direct current operated so as to render device totally independent of alternating current power. The present device can also provide emergency feedwater to the steam generator. When the pressure in chamber 36 exceeds a specified value, valve 62 opens, thus allowing the condensate stored therein to flow through check valves 64 and pipe 15 and into steam generator 14 as feedwater.
059873990	description	DETAILED DESCRIPTION OF PREFERRED EMBODIMENTS A system 10 constructed in accordance with the invention is set forth in general in the flow chart of FIG. 1A. In describing various preferred embodiments, specific reference will be made throughout to application of the surveillance methodologies to specific industrial systems, such as nuclear reactors; however, the inventions are equally applicable to any system which provides signals or other data over time which describe attributes or parameters of the system. Therefore, the inventions herein are, for example, applicable to analysis, modification and termination of processes and systems involving physical, chemical, biological and financial sources of data or signals. The system 10 is made up of three methodologies which, as appropriate, can be used separately, and possibly together, to monitor or validate data or signals. A series of logical steps can be taken to choose one or more of the methods shown in detail in FIGS. 1B-1D. Initialization of the system 10 is shown in FIG. 1A. The first step in the initialization is to obtain the user specified parameters; the Sample Failure Magnitude (SFM), the false alarm probability (.alpha.), and the missed alarm probability (.beta.). The next step in the initialization is to query the monitored system to obtain the sensor configuration information. If the system has a single sensor the method selected for monitoring will be the MONOSPRT approach described immediately hereinafter. For the single sensor case, that is all that needs to be done to complete the initialization. If the system has exactly two sensors, then information about the relationship between the two sensors is required. First, are the two sensors linearly related? If so, the regression SPRT algorithm is selected for monitoring, and this will be discussed in detail hereinafter. If the two sensors aren't linearly related, the next step is to check to see if they are non-linearly related. If so, the BART algorithm (described hereinafter) is used for monitoring. Otherwise, each sensor is monitored separately using the MONOSPRT method. In a first preferred embodiment (MONOSPRT) involving surveillance and analysis of systems having only one source of signals or data, such as, non-safety grade nuclear reactors and many industrial, biological and financial processes, a highly sensitive methodology implements a sequential analysis technique when the decision process is based on a single, serially correlated stochastic process. This form of the invention is set forth in detail in FIG. 1B on the portion of the flow diagram of FIG. 1A directed to "one sensor" which activates a MONOSPRT methodology. Serial correlation can be handled by a vectorized type of SPRT (sequential probability ratio test) method which is based on a time series analysis, multivariate statistics and the parametric SPRT test (see, for example, U.S. Pat. Nos. 5,223,207; 5,410,492; 5,586,066 and 5,629,872 which describe details of various SPRT features and are incorporated by reference herein for such descriptions). The MONOSPRT method is described in FIG. 1B. The method is split into two phases, a training phase and a monitoring phase. During the training phase N samples are collected from the single sensor (or data source) that are representative of normal operation. Next, a covariance matrix is constructed from the representative data that is p.sub.x p, where p is the user specified number of lags to consider when characterizing the autocorrelation structure of the sensor signal. The final steps in the training phase of the MONOSPRT method are to calculate the SPRT parameters; SDM, L and U. The SDM (System Disturbance Magnitude) is calculated by multiplying the standard deviation of the sensor signal with the SFM specified during the system initialization. The standard deviation of the sensor signal is the square root of the diagonal elements of the covariance matrix. L and U are the lower and upper thresholds used for to compare the MONOSPRT indexes to in order to make a failure decision. Both L and U are functions of .alpha. and .beta. specified during system initialization. During the monitoring phase of MONOSPRT a data vector of length p is acquired at each time step t and is used in the calculation of the MONOSPRT index .lambda.. The index is then compared to L and H. If the MONOSPRT index is greater than or equal to U, then the sensor signal is not behaving normally and a failure alarm is annunciated. If the MONOSPRT index is less than or equal to L then the decision that the sensor is good is made. In either case, after a decision is made in the MONOSPRT index is reset to zero and the process continues. In this vectorized SPRT methodology, (hereinafter "MONOSPRT"), suppose there exists the following stationary, a periodic sequence of serially correlated random variables: {X'}.sub.t where t=1, 2, 3 . . . , N. It is conventional that a periodic sequence can be handled by removing the periodic component of the structural time series model, and a non-stationary sequence can be differenced to produce a stationary sequence. The stationary assumption provides constant mean, constant variance and covariances that depend only on the separation of two variates in time and not the actual times at which they were recorded. The mean, .mu., is given by EQU .mu.=E[X'.sub.t ] where E[.multidot.] is the expectation operator. If we let EQU X.sub.t =X'.sub.t.sup.- X PA1 where, ##EQU1## and n.sub.s is the sample size, then E[X.sub.t ]=0. The autocovariance of two time points, X.sub.t and X.sub.s is .sigma..sub..vertline.t-s.vertline. =E[X.sub.t X.sub.s ], where s and t are integers in the set {[1, N]} and .sigma..sub.0 is the variance. Suppose there exists p<N such that for every m>p: .sigma..sub.m <.delta., where .delta. is arbitrarily close to 0. ##EQU2## PA1 Multivariate Normal: ##EQU6## where S is either 0 or A. Therefore: ##EQU7## The equation for .lambda..sub.t can be simplified into a more computationally efficient form as follows: ##EQU8## For the sequential test the equation is written as ##EQU9## PA1 H.sub.0 : D.sub.1,D.sub.2, . . . have Gaussian distribution with mean M.sub.0 and variance .sigma..sup.2 PA1 A) The similarity between the maximum and minimum values in the similarity domain is 0, and PA1 B) the similarity between equal values is 1. PA1 if x.sub.1 =[X.sub.11 X.sub.12 X.sub.13 . . . X.sub.1n ]and x.sub.2 =[X.sub.21 X.sub.22 X.sub.23 . . . X.sub.2n ] Therefore, we have constructed a stationary sequence of random vectors. The mean of the sequence is where is the zero vector with p rows. The variance of the sequence is the covariance matrix .SIGMA..sub.Y. ##EQU3## The SPRT type of test is based on the maximum likelihood ratio. The test sequentially samples a process until it is capable of deciding between two alternatives: H.sub.0: .mu.=0; and H.sub.A : .mu.=M. It has been demonstrated that the following approach provides an optimal decision method (the average sample size is less than a comparable fixed sample test). A test statistic, .lambda..sub.t, is computed from the following formula: ##EQU4## where 1n(.multidot.) is the natural logarithm, .function..sub.H.sbsb.s () is the probability density function of the observed value of the random variable Y.sub.i under the hypothesis H.sub.s and j is the time point of the last decision. In deciding between two alternative hypotheses, without knowing the true state of the signal under surveillance, it is possible to make an error (incorrect hypothesis decision). Two types of errors are possible. Rejecting H.sub.0 when it is true (type I error) or accepting H.sub.0 when it is false (type II error). We would like to control these errors at some arbitrary minimum value, if possible. We will call the probability of making a type I error, .alpha., and the probability of making a type II error .beta.. The well-known Wald's Approximation defines a lower bound, L, below which one accepts H.sub.0 and an upper bound, U beyond which one rejects H.sub.0. ##EQU5## Decision Rule: if .lambda..sub.t <L, then ACCEPT H.sub.0 ; else if .lambda., <U, then REJECT H.sub.0 ; otherwise, continue sampling. To implement this procedure, this distribution of the process must be known. This is not a problem in general, because some a priori information about the system exists. For our purposes, the multivariate Normal distribution is satisfactory. In practice, we implement two separate tests. One test is for M greater than zero and the second test for M less than zero. Here, M is chosen by the evaluating, EQU M=[1 1 1 . . . 1].sigma..sub.0 k (11) where k is a user specified constant that is multiplied by the standard deviation of y. M is then used in equation (10) to determine the amount of change in the mean of y that is necessary to accept the alternative hypothesis. FIGS. 2A-2F show results after applying the MONOSPRT embodiment to a sinusoid containing no disturbance, a step disturbance, and a linear drift. In these examples the noise added to the sinusoid is Gaussian and white with a variance of 2. The sinusoid has an amplitude of 1, giving an overall signal-to-noise ratio ("SNR" hereinafter) of 0.25 (for a pure sinusoid SNR=0.5A.sup.2 /.sigma..sup.2, where .sigma..sup.2 is the variance of the noise and A is the amplitude of the sinusoid). The autocorrelation matrix used in MONOSPRT for these examples were calculated using 30 lags. The false alarm probability .alpha. and missed alarm probability .beta. are both specified to be 0.0001 for MONOSPRT, and the sample failure magnitude (SFM) is set to 2.5. FIG. 2A shows the sinusoid with noise without any disturbance being present. FIG. 2B is the resulting MONOSPRT when applied to the signal. FIGS. 2C and 2D illustrate the response of MONOSPRT to a step change in the sinusoid. The magnitude of the step is 2.sigma..sub.s, where .sigma..sub.s is the standard deviation of the sinusoid plus the noise. The step begins at time 500 seconds. Due to the low SNR, MONOSPRT takes 25 samples to alarm, indicating that the signal is not at a peak in the sinusoid but rather that the mean of the overall signal has changed. In FIGS. 2E and 2F analogous MONOSPRT results are shown for a linear drift introduced into the noisy sinusoid signal. Here, the drift starts at time 500 seconds at a value of 0 and increases linearly to a final value of 4.sigma..sub.s at 1000 seconds. MONOSPRT detects the drift when it has reached a magnitude of approximately 1.5.sigma..sub.s. In FIGS. 3A-3F the results of running the same experiment are shown except this time the SNR is 0.5 and the SFM is changed to 1.5. The degree of autocorrelation is much higher in this case, but MONOSPRT can detect the disturbances more quickly due to the increased SNR. To test MONOSPRT on an actual sensor signal exhibiting non-white characteristics a sensor signal was selected from the primary pump #2 of the EBR-II nuclear reactor at Argonne National Laboratory (West) in Idaho. The signal is a measure of the pump's speed over a 1000 minute interval. FIG. 4A shows the sensor signal under normal operating conditions. The MONOSPRT results are shown in FIG. 4B. For this example .alpha. and .beta. are specified to be 0.0001 and the SFM is 2.5. The autocorrelation matrix was calculated using 10 lags. In FIGS. 5A and 5B MONOSPRT results are shown when a very subtle sensor drift is simulated. FIG. 5A is the sensor signal with a linear drift starting at time 500 min and continuing through the rest of the signal to a final value of -0.10011% of the sensor signal magnitude. MONOSPRT detects this very small drift after about only 50 min, i.e. when the drift has reached a magnitude of approximately 0.01% of the signal magnitude. The MONOSPRT plot is shown in FIG. 5B with the same parameter settings as were used in FIG. 4B. FIG. 5B illustrates the extremely high sensitivity attainable with the new MONOSPRT methodology. In another preferred embodiment (the regression SPRT method of FIG. 1C), a methodology provides an improved method for monitoring redundant process signals of safety- or mission-critical systems. In the flow diagram shown in FIG. 1C., the method is split into two phases, a training phase and a monitoring phase. During the training phase N data samples are collected from both sensors when the system is operating normally. The two data sets are then used to calculate the regression coefficients m and b using the means of both sensor signals (.mu..sub.1 and .mu..sub.2), the autocorrelation coefficient of one of the sensors (.sigma..sub.22), and the cross-correlation coefficient (.sigma..sub.12) between both sensors. The SPRT parameters are also calculated in the same manner as was calculation of the SDM is from the regression difference function. During the monitoring phase of the regression SPRT method, a regression based difference (D.sub.t) is generated at each time point t. The regression based difference is then used to calculate the SPRT index and to make a decision about the state of the system or sensors being monitored. The logic behind the decision is analogous to the decision logic used in the MONOSPRT method. Further details are described hereinafter. In this method, known functional relationships are used between process variables in a SPRT type of test to detect the onset of system or sensor failure. This approach reduces the probability of false alarms while maintaining an extremely high degree of sensitivity to subtle changes in the process signals. For safety- or mission-critical applications, a reduction in the number of false alarms can save large amounts of time, effort and money due to extremely conservative procedures that must be implemented in the case of a failure alarm. For example, in nuclear power applications, a failure alarm could cause the operators to shut down the reactor in order to diagnose the problem, an action which typically costs the plant a million dollars per day. In this preferred embodiment shown schematically in flow diagram FIG. 1C (two sensors, linearly related), highly redundant process signals can be monitored when the signals have a known functional relationship given by EQU X.sub.1 =f(X.sub.2) (12) where f() is some function determined by physical laws or by known (or empirically determined) statistical relationships between the variables. In principle, if either of the process signals X.sub.1 or X.sub.2 have degraded (i.e. fallen out of calibration) or failed, then (12) will no longer hold. Therefore, the relationship (12) can be used to check for sensor or system failure. In practice, both monitored process signals, or any other source of signals, contain noise, offsets and/or systematic errors due to limitations in the sensors and complexity of the underlying processes being monitored. Therefore, process failure cannot be detected simply by checking that (12) holds. More sophisticated statistical techniques must be used to ensure high levels of noise or offset do not lead to false and missed failure alarms. This preferred embodiment involves (a) specifying a functional relationship between X.sub.1 and X.sub.2 using known physical laws or statistical dependencies and linear regression when the processes are known to be in control, and (b) using the specified relationship from (a) in a sequential probability ratio test (SPRT) to detect the onset of process failure. For example, in many safety- or mission-critical applications, multiple identical sensors are often used to monitor each of the process variables of interest. In principle, each of the sensors should give identical readings unless one of the sensors is beginning to fail. Due to measurement offsets and calibration differences between the sensors, however, the sensor readings may be highly statistically correlated but will not be identical. By assuming that the sensor readings come from a multivariate normal distribution, a linear relationship between the variables can be specified. In particular, for two such sensor readings it is well-known that the following relationship holds EQU E[X.sub.1 .vertline.X.sub.2 ]=.sigma..sub.12 /.sigma..sub.12 (X.sub.2 -u.sub.2)+u.sub.1 (13) where E[X.sub.1 .vertline.X.sub.2 ] is the conditional expectation of the signal X.sub.1 given X.sub.2, .sigma..sub.12 is the square root of the covariance between X.sub.1 and X.sub.2. The .sigma..sub.22 is the standard deviation of X.sub.2, and u.sub.1 and u.sub.2 are the mean of X.sub.1 and X.sub.2, respectively. Equation (13) is simply a linear function of X.sub.2 and can therefore be written EQU X.sub.1 =mX.sub.2 +b (14) In practice, the slope m=.sigma..sub.12 /.sigma..sub.22 and intercept b=-.sigma..sub.12 /.sigma..sub.22 u.sub.2 +u.sub.1 can be estimated by linear regression using data that is known to have no degradation or failures present. Once a regression equation is specified for the relationship between X.sub.1 and X.sub.2, then the predicted X.sub.1 computed from (14) can be compared to the actual value of X.sub.1 by taking the difference EQU D.sub.1 =X.sub.1 -(mX.sub.2 +b) (15) Under normal operating conditions, D.sub.1, called the regression-based difference, will be Gaussian with mean zero and some fixed standard deviation. As one of the sensors begins to fail or degrade, the mean will begin to change. A change in the mean of this regression-based difference can be detected using the SPRT methodology. The SPRT approach is a log-likelihood ratio based test for simple or composite hypothesis (also see the incorporated patents cited hereinbefore). To test for a change in the mean of the regression-based difference signal D.sub.1, D.sub.2, . . . , the following two hypotheses are constructed: H.sub.F : X.sub.1,X.sub.2, . . . have Gaussian distribution with mean M.sub.F and variance .sigma..sup.2 where H.sub.0 refers to the probability distribution of the regression-based difference under no failure and H.sub.F refers to the probability distribution of the regression-based difference under system or process failure. The SPRT is implemented by taking the logarithm of the likelihood ratio between H.sub.0 and H.sub.F. In particular, let f.sub.0 (d.sub.i) represent the probability density function for D.sub.1, D.sub.2, . . . under H.sub.0, and f.sub.1 (d.sub.i) represent the probability density function for D.sub.1, D.sub.2, . . . under H.sub.F. Let Z.sub.i =log [.function..sub.1 (X.sub.i)/.function..sub.0 (X.sub.i)] the log-likelihood ratio for this test. Then ##EQU10## Defining the value S.sub.n to be the sum of the increments Z.sub.i up to time n where S.sub.n =.SIGMA..sub.1.ltoreq.i.ltoreq.n Z.sub.i, then the SPRT algorithm can be specified by the following: ##EQU11## The endpoints A and B are determined by the user specified error probabilities of the test. In particular, let .alpha.=P{ conclude H.sub.F .vertline. H.sub.0 true} be the type I error probability (false alarm probability) and .beta.=P{ conclude H.sub.0 .vertline. H.sub.F true} be the type II error probability (missed alarm probability) for the SPRT. Then ##EQU12## For real time applications, this test can be run repeatedly on the computed regression-based difference signal as the observations are collected so that every time the test concludes H.sub.0, the sum S.sub.n is set to zero and the test repeated. On the other hand, if the test concludes H.sub.F, then a failure alarm is sounded and either the SPRT is repeated or the process terminated. An illustration of this preferred form of bivariate regression SPRT method can be based on the EBR-II nuclear reactor referenced hereinbefore. This reactor used redundant thermocouple sensors monitoring a subassembly outlet temperature, which is the temperature of coolant exiting fuel subassemblies in the core of the reactor. These sensors readings are highly correlated, but not identical. The method of this embodiment as applied to this example system was performed using two such temperature sensors; X.sub.1 =channel 74/subassembly outlet temperature 4E1, and X.sub.2 =channel 63/ subassembly outlet temperature 1A1. For 24 minutes worth of data during normal operation on Jul. 7, 1993, a regression line is specified for X.sub.1 as a function of X.sub.2 according to equation (14). The predicted X.sub.1 from (14) is then compared to the actual X.sub.1 by taking the regression-based difference (15) in our new regression-SPRT algorithm. The results of this experiment are then compared to the results of performing a prior-art SPRT test on the difference X2-X1 according to U.S. Pat. No. 5,410,492. Plots of subassembly outlet temperature 1A1 and 4E1 under normal operating conditions are given in FIG. 6A and 6B. The relationship between the two variables when no failure is present is illustrated in FIG. 7. In FIG. 7, the slope and intercept of the regression line from equation (14) are given. FIGS. 8A and 8B illustrate the regression-based difference signal along with the difference signal of the prior art proposed by U.S. Pat. No. 5,223,207. It is easy to see that the regression-based difference signal tends to remain closer to zero than the original difference signal under normal operating conditions. FIGS. 9A and 9B plot the results of a SPRT test on both the regression-based difference signal and the original difference signal. In both cases, the pre-specified false- and missed-alarm probabilities are set to 0.01, and the threshold for failure (alternate hypothesis mean) is set to 0.5.degree. F. In both subplots, the circles indicate a failure decision made by the SPRT test. Note that under no failure or degradation modes, the new regression-based SPRT gives fewer false alarms than the original difference. The calculated false alarm probabilities are given in Table I for these comparative SPRT tests plotted in FIGS. 9A and 9B. TALBE I ______________________________________ Empirical False Alarm Probability for the SPRT test to Detect Failure of an EBR-II Subassembly Outlet Temperature Sensor Regression-Based Original Difference Difference ______________________________________ False Alarm Probability 0.025 0.0056 ______________________________________ The empirical false alarm probability for the SPRT operated on the regression-based difference (see FIG. 9A) is significantly smaller than the for the SPRT performed on the original difference signal (see FIG. 9B), indicating that it will have a much lower false-alarm rate. Furthermore, the regression-based difference signal yields a false alarm probability that is significantly lower than the pre-specified false alarm probability, while the original difference function yields an unacceptably high false alarm probability. To illustrate the performance of the regression-based difference method in a SPRT methodology under failure of one of the sensors, a gradual trend is added to the subassembly outlet temperatures 1A1. 4E1 to simulate the onset of a subtle decalibration bias in that sensor. The trend is started at 8 minutes, 20 seconds, and has a slope of 0.005.degree. F. per second. These EBR-II signals with a failure injected are plotted in FIGS. 10A and 10B. The respective regression-based difference signal and the original difference signal are plotted in FIGS. 11A and 11B. FIGS. 12A and 12B respectively, plot the results of the SPRT test performed on the two difference signals. As before, the SPRT has false and missed alarm probabilities of 0.01, and a sensor failure magnitude of 0.5.degree. F. In this case, the regression-based SPRT annunciated the onset of the disturbance even earlier than the conventional SPRT. The time of failure detection is given in Table II. TABLE II ______________________________________ Time to Detection of Gradual Failure of EBR-II Subassembly Outlet Temperature Regression-Based Original Difference Difference ______________________________________ Time to Failure Detection 9 min. 44 sec. 9 min. 31 sec. ______________________________________ These results indicate that the regression-based SPRT methodology yields results that are highly sensitive to small changes in the mean of the process. In this case, using the regression-based SPRT gave a failure detection 13 seconds before using the prior art method. A problem that is endemic to conventional signal surveillance methods is that as one seeks to improve the sensitivity of the method, the probability of false alarms increases. Similarly, if one seeks to decrease the probability of false alarms, one sacrifices sensitivity and can miss the onset of subtle degradation. The results shown here illustrate that the regression-based SPRT methodology for systems involving two sensors simultaneously improves both sensitivity and reliability (i.e. the avoidance of false alarms). It is also within the scope of the preferred embodiments that the method can be applied to redundant variables whose functional relationship is nonlinear. An example of this methodology is also illustrated in FIG. 1 branching off the "sensors are linearly related" to the "monitor separately" decision box which can decide to do so by sending each signal to the MONOSPRT methodology or alternatively to the BART methodology described hereinafter. In particular for a nonlinear relation, if the monitored processes X.sub.1 and X.sub.2 are related by the functional relationship EQU X.sub.1 =f(X.sub.2) (18) where f() is some nonlinear function determined by physical laws (or other empirical information) between the variables, then the relationship (18) can be used to check for sensor or system failure. In this case, the relationship (18) can be specified by using nonlinear regression of X.sub.1 on X.sub.2. The predicted X.sub.1 can then be compared to the actual X.sub.1 via the regression-based SPRT test performed on the resulting nonlinear regression-based difference signal. In another form of the invention shown in FIG. 1D in systems with more than two variables one can use a nonlinear multivariate regression technique that employs a bounded angle ratio test (hereinafter BART) in N Dimensional Space (known in vector calculus terminology as hyperspace) to model the relationships between all of the variables. This regression procedure results in a nonlinear synthesized estimate for each input observation vector based on the hyperspace regression model. The nonlinear multivariate regression technique is centered around the hyperspace BART operator that determines the element by element and vector to vector relationships of the variables and observation vectors given a set of system data that is recorded during a time period when everything is functioning correctly. In the BART method described in FIG. 1D., the method is also split into a training phase and a monitoring phase. The first step in the training phase is to acquire a data matrix continuing data samples from all of the sensors (or data sources) used for monitoring the system that are coincident in time and are representative of normal system operation. Then the BART parameters are calculated for each sensor (Xmed, Xmax and Xmin). Here Xmed is the median value of a sensor. The next step is to determine the similarity domain height for each sensor (h) using the BART parameters Xmed, Xmax and Xmin. Once these parameters are calculated a subset of the data matrix is selected to create a model matrix (H) that is used in the BART estimation calculations. Here, H is an N.times.M matrix where N is the number of sensors being monitored and M is the number of observations stored from each sensor. As was the case in both the MONOSPRT and regression SPRT method, the last steps taken during the training phase are the SPRT parameters calculations. The calculations are analogous to the calculations in the other methods, except that now the standard deviation value used to calculate SDI is obtained from BART estimation errors from each sensor (or data source) under normal operating conditions. During the BART monitoring phase a sample vector is acquired at each time step t, that contains a reading from all of the sensors (or data sources) being used. Then the similarity angle (SA) between the sample vector and each sample vector stored in H is calculated. Next an estimate of the input sample vector Y is calculated using the BART estimation equations. The difference between the estimate and the actual sensor values is then used as input to the SPRT. Each difference is treated separately so that a decision can be made on each sensor independently. The decision logic is the same as is used in both MONOSPRT and the regression SPRT methods. This method is described in more detail immediately hereinafter. In this embodiment of FIG. 1D of the invention, the method measures similarity between scalar values. BART uses the angle formed by the two points under comparison and a third reference point lying some distance perpendicular to the line formed by the two points under comparison. By using this geometric and trigonometric approach, BART is able to calculate the similarity of scalars with opposite signs. In the most preferred form of BART an angle domain must be determined. The angle domain is a triangle whose tip is the reference point (R), and whose base is the similarity domain. The similarity domain consists of all scalars which can be compared with a valid measure of similarity returned. To introduce the similarity domain, two logical functional requirements can be established: Thus we see that the similarity range (i.e. all possible values for a measure of similarity), is the range 0 to 16) inclusive. BART also requires some prior knowledge of the numbers to be compared for determination of the reference point (R). Unlike a ratio comparison of similarity, BART does not allow "factoring out" in the values to be compared. For example, with the BART methodology the similarity between 1 and 2 is not necessarily equal to the similarity between 2 and 4. Thus, the location of R is vital for good relative similarities to be obtained. R lies over the similarity domain at some distance h, perpendicular to the domain. The location on the similarity domain at which R occurs (Xmed) is related to the statistical distribution of the values to be compared. For most distributions, the median or mean is sufficient to generate good results. In or preferred embodiment the median is used since the median provides a good measure of data density, and is resistant to skewing caused by large ranges of data. Once Xmed has been determined, it is possible to calculate h. In calculating h, it is necessary to know the maximum and minimum values in the similarity domain. (Xmax and Xmin respectively) for normalization purposes the angle between Xmin and Xmax is defined to be 90.degree.. The conditions and values defined so far are illustrated in FIG. 13. From this triangle it is possible to obtain a system of equations and solve for h as shown below: ##EQU13## Once h has been calculated the system is ready to compute similarities. Assume that two points: X.sub.0 and X.sub.1 (X.sub.0 .ltoreq.X.sub.1) are given as depicted in FIG. 14 and the similarity between the two is to be measured. The first step in calculating similarity is normalizing X.sub.0 and X.sub.1 with respect to Xmed. This is done by taking the euclidean distance between Xmed and each of the points to be compared. Once X.sub.0 and X.sub.1 have been normalized, the angle .angle.X.sub.0 RX.sub.1 (hereinafter designated .theta.) is calculated by the formula: EQU .theta.=ArcTan(X.sub.1 .vertline.h)=ArcTan(X.sub.0 .vertline.h)(20) After .theta. has been found, it must be normalized so that a relative measure of similarity can be obtained that lies within the similarity range. To ensure compliance with functional requirements (A) and (B) made earlier in this section, the relative similarity angle (SA) is given by: ##EQU14## Formula (21) satisfies both functional requirements established at the beginning of the section. The angle between Xmin and Xmax was defined to be 90.degree., so the similarity between Xmin and Xmax is 0. Also, the angle between equal values is 0.degree.. The SA therefore will be confined to the interval between zero and one, as desired. To measure similarity between two vectors using the BART methodology, the average of the element by element SAs are used. Given the vectors x.sub.1 and x.sub.2 the SA is found by first calculating S.sub.i for i=1,2,3 . . . n for each pair of elements in x.sub.1 and x.sub.2 i.e., The vector SA .GAMMA. is found by averaging over the S.sub.i 's and is given by the following equation. ##EQU15## In general, when given a set of multivariate observation data from a process (or other source of signals), we could use linear regression to develop a process model that relates all of the variables in the process to one another. An assumption that must be made when using linear regression is that the cross-correlation information calculated from the process data is defined by a covariance matrix. When the cross-correlation between the process variables is nonlinear, or when the data are out of phase, the covariance matrix can give misleading results. The BART methodology is a nonlinear technique that measures similarity instead of the traditional cross-correlation between variables. One advantage of the BART method is that it is independent of the phase between process variables and does not require that relationships between variables be linear. If we have a random observation vector y and a known set of process observation vectors from a process P, we can determine if y is a realistic observation from a process P by combining BART with regression to form a nonlinear regression method that looks at vector SAs as opposed to Euclidean distance. If the know observation vectors taken from P are given by, ##EQU16## where H is k by m (k being the number of variables and m the number of observations), then the closest realistic observation vector to y in process P given H is given by EQU y=Hw (24) Here w is a weighting vector that maps a linear combination of the observation vectors in H to the most similar representation of y. The weighting vector w is calculated by combining the standard least squares equation form with BART. Here, .theta. stands for the SA operation used in BART. EQU w=(H'.sym.H).sup.-1 H'.sym.y (25) An example of use of the BART methodology was completed by using 10 EBR-II sensor signals. The BART system was trained using a training data set containing 1440 observation vectors. Out of the 1440 observation vectors 129 were chosen to be used to construct a system model. The 129 vectors were also used to determine the height h of the angle domain boundary as well as the location of the BART reference point R for each of the sensors used in the experiment. To test the accuracy of the model 900 minutes of one minute data observation vectors under normal operating conditions were run through the BART system. The results of the BART system modeling accuracy are shown in FIGS. 15A-15E and FIGS. 16A-16E (BART modeled). The Mean Squared Errors for each of the sensor signals is shown in Table III. TABLE III ______________________________________ BART System Modeling Estimation Mean Squared Errors for EBR-II Sensor Signals MSE of Normalized Normalized Sensor Estimation MSE MSE Channel Sensor Description Error (MSE/.mu..sub.3) (MSE/.sigma..sub.3) ______________________________________ 1. Primary Pump #1 0.0000190 0.0000002 0.0002957 Power (KW) 2. Primary Pump #2 0.0000538 0.0000004 0.0004265 Power (KW) 3. Primary Pump #1 0.0000468 0.0000001 0.0005727 Speed (RPM) 4. Primary Pump #2 0.0000452 0.0000001 0.0004571 Speed (RPM) 5. Reactor Outlet 8.6831039 0.0009670 0.1352974 Flowrate (GPM) 6. Primary Pump #2 0.0571358 0.0000127 0.0163304 Flowrate (GPM) 7. Subassembly Outlet 0.0029000 0.0000034 0.0062368 Temperature 1A1 (F) 8. Subassembly Outlet 0.0023966 0.0000027 0.0052941 Temperature 2B1 (F) 9. Subassembly Outlet 0.0025957 0.0000029 0.0050805 Temperature 4E1 (F) 10. Subassembly Outlet 0.0024624 0.0000028 0.0051358 Temperature 4F1 (F) ______________________________________ A second example shows the results of applying BART to ten sensors signals with three different types of disturbances with their respective BART estimates superimposed followed by the SPRT results when applied to the estimation error signals. The first type of disturbance used in the experiment was a simulation of a linear draft in channel #1. The drift begins at minute 500 and continues through to the end of the signal, reaching a value of 0.21% of the sensor signal magnitude and the simulation is shown in FIG. 17A. The SPRT (FIG. 17B) detects the drift after it has reached a value of approximately 0.06% of the signal magnitude. In FIG. 17C a simulation of a step failure in channel #2 is shown. Here the step has a height of 0.26% of the signal magnitude and begins at minute 500 and continues throughout the signal. FIG. 17D shows the SPRT results for the step failure. The SPRT detects the failure immediately after it was introduced into the signal. The last simulation was that of a sinusoidal disturbance introduced into channel #6 as shown in FIG. 17E. The sinusoid starts at minute 500 and continues throughout the signal with a constant amplitude of 0.15% of the sensor signal magnitude. The SPRT results for this type of disturbance are shown in FIG. 17F. Again the SPRT detects the failure even though the sinusoid's amplitude is within the operating range of the channel #6 sensor signal. In further variations on the above described embodiments a user can generate one or more estimated sensor signals for a system. This methodology can be useful if a sensor has been determined to be faulty and the estimated sensor signal can be substituted for a faulty, or even degrading, sensor or other source of data. This methodology can be particularly useful for a system having at least three sources of data, or sensors. While preferred embodiments of the invention have been shown and described, it will be clear to those skilled in the art that various changes and modifications can be made without departing from the invention in its broader aspects as set forth in the claims provided hereinafter.
	claims	1. A radiation shield characterised by comprising a hydride and a thermoplastic or thermosetting polymer binder wherein the radiation shield comprises greater than 14.4 wt % hydrogen. 2. A radiation shield as claimed in claim 1 characterised in that the hydride is selected from one or more of ammonia borane, ammonium borohydride, methylammonium borohydride, an ammoniate of lithium borohydride, a methyl amine borane, ammonia triborane, ammonium octahydrotriborane, lithium borohydride and beryllium hydride. 3. A radiation shield as claimed in claim 1 characterised in that the polymer binder is one or more of polyethylene, polypropylene, polyisobutylene, polybutadiene, poly(methyl methacrylate), polysulphone, polystyrene, poly(vinyl pyrrolidone), poly(vinylidene fluoride), poly(tetrafluoromethylene), polyethylene oxide, poly(vinyl acetate), polyester, a co-polymer comprising two or more polymers, polyepoxide, polyimide, polyamide, polyaramide and melamine formaldehyde. 4. A radiation shield as claimed in claim 1 characterised in that the polymer binder is one or more of polyethylene, polypropylene, poly(4-methyl-1-pente), nylon6,6, poly(vinyl alcohol) and poly(oxymethylene). 5. A radiation shield as claimed in claim 1 characterised in that the hydride has a hydrogen content in the range 14 to 25 wt %. 6. A radiation shield as claimed in claim 1 characterised in that it is prepared by polymerising a monomer corresponding to the polymer binder in the presence of the hydride and a polymerisation catalyst. 7. A radiation shield as claimed in claim 1 characterised in that it is prepared by mixing the polymer binder and the hydride before moulding. 8. A radiation shield as claimed in claim 6 characterised in that preparation is carried out in the presence of a solvent. 9. A radiation shield as claimed in claim 1 characterised by comprising a composite of alternating layers of polymer binder and hydride. 10. A radiation shield as claimed in claim 1 characterised by comprising a flexible film of the polymer binder and hydride. 11. A radiation shield as claimed in claim 1 characterised by a mat of fibres comprising the polymer binder and the hydride. 12. A radiation shield as claimed in claim 1 characterised by comprising the hydride held in the pores of a foam of the polymer binder. 13. A radiation shield as claimed in claim 1 characterised by comprising a solid block prepared by sintering a mixture of the thermoplastic polymer binder and the hydride at high pressures and high temperatures. 14. A radiation shield as claimed in claim 1 characterised by comprising a mixture of the hydride and polymer binder in a polymer box. 15. A radiation shield as claimed in claim 1 characterised in that the hydride does not melt or release hydrogen at temperatures below 100° C. 16. A radiation shield as claimed in claim 1 characterised by further comprising a polymer layer or metallised polymer layer, said layer being gas impermeable and enveloping the hydride. 17. A radiation shield as claimed in claim 1 characterised in that the radiation shield comprises a neutron absorber, wherein the neutron absorber is one or more of lithium isotopically enriched with lithium-6 and boron isotopically enriched with boron-10.
	description	FIGS. 1 and 2 generally depict a spent nuclear fuel assembly pool 10 defined by a floor 12 and sidewalls 14 and containing water up to a water line 16. The pool 10 contains existing first spend nuclear fuel assembly storage racks 20 that are supported by structural members such as support pads 21 on the floor 12 of the pool 10. The storage racks 20 may alternatively be supported via a frame (not shown) that in turn is supported by the floor 12. Floor supported storage systems are discussed in U.S. Pat. No. 5,152,958 and the patents it references, including U.S. Pat. Nos. 4,889,681; 4,666,660; 4,336,103; 4,069,923 and 4,029,968, which patents are incorporated by this reference for their disclosures of the structures of such systems. As is discussed in U.S. Pat. No. 5,152,958 and referenced patents, including U.S. Pat. No. 4,287,426 and West German Patent No. 29 43 455, the spent nuclear fuel assembly storage racks may be wall supported instead of floor supported. These patents are incorporated by this reference for their disclosures of the structures of such systems. In order to increase the capacity of the spent fuel assembly fuel pools such as pool 10, one or more second spent nuclear fuel assembly storage racks 30 with support pads 31 are installed in the pool 10 over the previously installed racks 20 as shown in FIG. 1. Importantly, the second storage racks 30 are structurally supported independently of the previously installed storage racks 20. Thus, the second storage racks 30 are not directly or indirectly supported by the support pads 21, support frames or other structural support members supporting the previously installed storage racks 20. Advantageously, the second spent nuclear fuel assembly storage racks 30 may be installed while spent nuclear fuel assemblies are being stored in the previously installed racks 20. As shown in FIG. 1, the second storage racks 30 may be supported by a platform 32 suspended from hanger rods 34 extending from hanger brackets 36 bolted or otherwise mounted to the existing operating deck 38 of a spent fuel pool 10. Advantageously, the second storage racks 30 and their structural supports may be designed to support the static load of the second storage racks 30 and the dynamic loads of earthquakes independently of the previously installed storage racks 20. FIG. 2 shows an alternative two tier storage arrangement where one or more second spent fuel assembly storage racks 30 having support pads 31 are supported on a platform 32. The platform 32 is supported by support members such as floor supported posts 42 that are structurally independent of support pads 21 or other structural members supporting the spent fuel assemblies 20 of the lower tier. The platform 32 may be stationary or, alternatively, may be horizontally mobile (not shown). The previously referenced U.S. Pat. No. 5,152,958 discloses structure of a horizontally mobile spent nuclear fuel assembly storage platform or bridge. As depicted in FIGS. 1 and 2, the spent nuclear fuel assemblies 30 of the upper tier, like the spent nuclear fuel assemblies 20 of the lower tier, are stored while vertically oriented, i.e., the lengths of the tubes in the assemblies extend vertically. Because fuel assemblies are generally from about eight feet to about fourteen feet in length, the water level 16 of the pool 10 will be no more than about 10 feet above the top of the spent fuel assemblies 30 of the upper tier in many existing spent nuclear fuel assembly pools. This depth is adequate to shield spent fuel rods that are at least about five years old. If additional shielding is desired, then additional shielding (not shown) may be provided over the top of the spent fuel assemblies 30. In the practice of the present invention, spent nuclear fuel assemblies are introduced into the spent nuclear fuel assembly storage racks 30 while vertically oriented. Accordingly, at least a portion of the spent nuclear fuel assemblies will be raised above the water level 16 in plants where the fuel assemblies are longer than about ten feet in length. In some practices, the spent nuclear fuel assemblies may be entirely raised above the water level 16 of the pool 10. FIG. 3 depicts a handling assembly 52 in the process of introducing a vertically oriented spent nuclear fuel assembly 50 into an upper tier spent nuclear fuel assembly storage rack 30. The handling assembly 52 may be similar to the design of assemblies carried by refueling machines to introduce fuel assemblies 50 into a reactor pressure vessel (not shown). See, e.g., U.S. Pat. Nos. 4,511,531 and 4,929,413, which show telescoping gripper assemblies carried by refueling machines. Thus, the handling assembly 52 may include a handling tool with air-operated gripper fingers (not shown) disposed in an elongated gripper tube 54 also having an integral shield plug (not shown) disposed above the fingers. The gripper tube 54 may be telescoping operated within a mast 56 by an electric winch 58 to raise and lower the fuel assembly 50 between a position in the storage racks 20 and 30 and a position in a vertically oriented transfer cask 62 mounted on the mast 56. Advantageously, the transfer cask 62 will provide radiation shielding for the spent fuel assembly 50 when near to and above the water surface 18. A support platform 60 supports the gripper tube 54, mast 56, winch 58 and transfer cask 62. The support platform 60 may also support an air hose reel 64 for the handling tool and an electric reel 66 for the electric winch 58. The platform 60 may be suspended over the spent fuel pool 10 via a cable 68 from the generating plant""s overhead spent fuel pool crane (not shown). While a present preferred embodiment of the present invention has been shown and described, it is to be understood that the invention may be otherwise variously embodied within the scope of the following claims of invention.
	description	This application claims priority to U.S. Provisional Patent Application Ser. No. 62/640,683 to Robert Kamen filed on Mar. 9, 2018 and entitled “COMPRESSION MEMBER FOR BIOHAZARDOUS MATERIAL TRANSPORTING PIG”, the contents of which are incorporated herein by reference. This invention relates to hazardous materials, for example radiopharmaceuticals. In particular this invention relates to a compression member for a pig for storing, transporting and dispensing of liquid and capsules formulations of biohazardous products and substances in liquid and solid form, for example radiopharmaceuticals. The transportation of biohazardous materials and substances, for example radioactive materials or biological substances such as pathogens, presents a potentially dangerous situation and must be subject to strict controls. For example, radioactive pharmaceutical products, commonly known as “radiopharmaceuticals,” are prepared for patient injection, ingestion or other forms of administration in specially equipped and controlled facilities. Radiopharmaceuticals are well known for use as markers in nuclear medicine diagnostic procedures, and to treat certain diseases. Unless properly shielded, such products become a radiation hazard for individuals handling the product. For example, radioiodine pills or capsules that can be used for treating certain pathologies such as thyroid diseases or in conjunction with a diagnostic procedure to diagnose certain types of illnesses, are stored before use in a container typically made of plastic, for example a polyethylene pill bottle. In the case of a liquid radiopharmaceutical the container is typically a glass vial. Neither of these containers have any radioactivity-shielding properties. Therefore the storage, transportation and dispensing of radiopharmaceuticals is carefully controlled by rules designed to regulate the handling of such materials in a manner that reduces the radiation hazard. Each metered (for example assayed or calibrated) dose of the radiopharmaceutical product, for example in the case of a treatment for thyroid issues a radioiodine pill, or in the case of isotopes used in Nuclear Medicine (SPECT) and positron emission tomography (PET) diagnostic procedures a liquid, is placed by the manufacturer into the container to be shipped to a qualified facility for administration to a particular patient or patient category. At the radiopharmacy stock vials of different radiopharmaceuticals are dispensed as unit doses. This represents the first opportunity for hazardous exposure to the radioactive contents, and accordingly is effected at the manufacturer in a shielded booth or other enclosure, or under other radioactivity-shielded conditions. The container containing the radiopharmaceutical must then be shipped to the destination hospital or clinic for administration to the patient. To effect this safely, the container is dropped into a radioactivity-shielding container commonly known as a “pig” for interim storage and delivery to the destination. A conventional pig comprises a two-part vessel which is either formed from a radioactivity-shielding material, for example lead or tungsten, or has an exterior shell encasing a radiopharmaceutical container compartment that is lined with a radioactivity-shielding material such as lead or tungsten. A non-limiting example is described and illustrated in U.S. Pat. No. 6,586,758 issued Jul. 1, 2003 to Martin, which is incorporated herein by reference in its entirety. When the pig is assembled, the radiopharmaceutical container compartment is sealed in order to contain the radiation and thus minimize human exposure to the radioactive contents of the radiopharmaceutical compartment. The compartment is sized to accommodate the radiopharmaceutical product, in the ingestible radioiodine example a pill or dissolving capsule, or in the case of a liquid of radiopharmaceutical a vial, syringe, ampule or other glass container. In each case the radiopharmaceutical compartment would be dimensioned accordingly. Once the radiopharmaceutical container has been placed into the radiopharmaceutical compartment and the pig assembled, the pig is ready to be shipped to the patient's location. Because this part of the delivery process occurs entirely within the confines of the manufacturing plant, which is specifically designed and staffed so as to meet all regulatory guidelines and procedures, there is less chance of human exposure to the radioactive radiopharmaceutical product up to the point that the pill, capsule, vial, syringe or the like is sealed in the radiopharmaceutical container compartment of the pig. As is well known, the pig is designed to provide optimal shielding so as to reduce exposure during shipping. The transportation phase is a second opportunity for exposure to the radioactive contents of the radiopharmaceutical container, posing an occupational exposure opportunity for the driver/courier. At the destination staff trained in handling radioactive substances, for example a nuclear medicine technologist or technician, opens the pig and then removes the closure from the radiopharmaceutical container to vent the container bottle. This is the third opportunity for exposure to the radioactive contents of the radiopharmaceutical container, in the presence of hospital or clinic staff. The technologist must transfer the radiopharmaceutical to a Dose Calibrator to assay (measure) the activity of the radiopharmaceutical, which must be within 10% of prescribed activity. After recording the assay, the technologist must retrieve container containing the radiopharmaceutical and return the radiopharmaceutical container to the pig's radiopharmaceutical container compartment, which is the third opportunity for exposure to radioactivity. The technologist then applies the lid to the pig for delivery to the patient. The pig is opened in the patient's presence in order to gain access to the radiopharmaceutical container and remove the container closure for administration of the radiopharmaceutical product to the patient, providing a fourth opportunity for exposure to the radioactive contents of the radiopharmaceutical container. In this step exposure of radioactivity to the ambient environment is unavoidable in order to access the radiopharmaceutical product for administration to the patient, so great care must be taken to handle the unshielded radiopharmaceutical product using proper safety equipment and procedures. However, the assaying process, and the venting of the container in the case of certain volatile radioactive substances which produce radioactive iodine vapours such as 131Iodine capsules, can present unnecessary points of risk of exposure to the technologist and other staff. Although the types of destination facilities to which these products are shipped are equipped to properly handle radiopharmaceutical products and the staff at such facilities are well trained in safety policies and procedures, this step in particular can increase the risk of human exposure to the radioactive contents of the radiopharmaceutical product. There is accordingly a need for a radiopharmaceutical pig that reduces opportunities for human exposure to the contents of the container when the pig reaches a hospital or clinic setting and the product in the container is exposed to the ambient environment. Such a pig is disclosed in PCT/CA2017/0505689 to Kamen, the contents of which are incorporated herein. While the inventions disclosed in the above-noted application to Kamen are useful, improvements are desirable. For example, improvements to the compression member to be disposed intermediate at least a container closure and the pig are desirable. According to an aspect there is provided a compression member for insertion into a pig for transporting a container of biohazardous materials, the compression member comprising a flange maintained in spaced relation with an annulus by pillars; and spaced apart pivotable grip components supported by the annulus and extending downwards from the annulus between respective ones of the pillars towards, but not into contact with, the flange, the pivotable grip components resiliently compressible inwardly against the container when the container is received within the compression member. In an embodiment, the compression member further comprises a ramp on an outward-facing surface of each of the grip components. In an embodiment, the compression member further comprises a buttress at the interface between each pillar and the annulus. In an embodiment, the flange of the compression member further comprises a sloped edge about its periphery for snap retention within the complementary annulus. In an embodiment, the flange, annulus, pillars and grip components are formed as a unitary structure. In an embodiment, the compression member is formed of a thermoplastic material. A pig 20 for transporting a container 10 containing a biohazardous product is shown in FIG. 1. The advantages of the pig 20 are particularly applicable in the case of radiopharmaceuticals, whether in solid or liquid form. However, the pig 20 may be configured to be suitable for transporting virtually any type of radiopharmaceutical product, and is also suitable for transporting other types of biohazardous products or substances such as biological pathogens. One or more advantages can be obtained in the use of a pig according to the invention for storing and transporting any kind of biohazardous product where access to the internal (non-protective) container holding the biohazardous product is required intermittently. The embodiments of the invention described herein are for purposes of example only and the invention is not intended to be limited to the specific embodiments described. A biohazardous materials container, for example a radiopharmaceutical container 10 as shown, comprises a bottle 12 and a closure 14 for sealing the bottle 12. The container 10 may be made of any suitable material, typically plastic or glass depending upon the type and form of radiopharmaceutical contained therein. For example in the embodiment shown in FIG. 2 the container 12 is a glass vial containing a liquid radiopharmaceutical 2. The cap 30 of the pig 20 is configured 1) to allow the container 10 to be removed from the body 22 of the pig 20 while secured to (and thus in part shielded by) the cap 30, and 2) to allow the closure 14 to be removed from the bottle 12 without opening the pig 20 in order to avoid exposing the user to the radioactive contents of the product, as described in detail below. In the embodiment shown the bottle 12 comprises a bead 12a about its neck, and the closure 14 is a stopper-type closure having a body 14a which closes the neck of the bottle 12 in an interference fit. In other containers 10 the closure may be clinched to the neck of the bottle 12. In the case of liquids the closure 14 is typically provided with a generally central septum 14b (see FIG. 12) for penetration by a syringe in order to extract the contents of the bottle 12. The pig 20 in the embodiment illustrated a radiopharmaceutical pig 20, comprises a cylindrical body 22 and a complementary cylindrical cap 30 for attachment to the body 22. The components of the radiopharmaceutical pig 20 shown may be formed from a radioactivity-shielding material such as lead or tungsten, or may be formed from any suitably strong metal or plastic. In the case of the radiopharmaceutical pig 20 shown the portions surrounding the compartment 24 are lined with a suitably radioactivity-resistant liner formed from a material such as lead or tungsten. If the pig is used to transport toxins, biological pathogens or other non-radioactive products or substances, the compartment 24 may be hermetically sealed when the pig 20 is closed to prevent exposure to the ambient environment. The body 22 comprises a recess concentric with and overlying the radiopharmaceutical container compartment 24, forming a throat 23 which provides projecting cams 25 along its interior wall, as best seen in FIG. 4. The cap 30 comprises a two-stage closure for sealing the biohazardous container compartment 24 against radioactivity leakage. The first body closure stage comprises an outer collar 30a that fits within the throat 23 of the body, which when secured to the body 22 extends into and sealingly engages with the throat 23. In the embodiment illustrated the collar 30a comprises a projecting collar neck portion 31 that provides external projecting cams 31a, best seen in FIG. 5, which are complementary to the cams 25 about the throat 23 and positioned so that when the neck 31 of the collar 30a is secured into the throat 23 above the biohazardous materials container compartment 24 by partial (e.g. 60 degree) rotation in a ‘bayonet’ connection, the lower edge 31b of the neck 31 sealingly engages against the floor 27 of the throat 23 around its periphery and prevents radioactivity from escaping around the collar 30a. The collar 30a comprises an orifice 29 extending through the body and neck 31 of the collar 30a, in communication with the biohazardous materials container compartment 24. The upper portion of the orifice 29 provides a larger diameter and projecting cams 31d (see FIG. 7) disposed about its interior surface, for receiving the cap closure 30b as described below. The orifice 29 narrows as it approaches the neck 31, creating a ledge 31c at an intermediate point for sealing engagement by the cap closure 30b. In some embodiments the narrower lower portion of the orifice 29 is adapted to receive a compression, or “grip”, member 50 that functions to grip closure 14 as will be described below. The cap closure 30b provides a cap closure neck 33 that fits into the orifice 29. In the embodiment illustrated the cap closure 30b comprises a projecting closure neck portion 33 that provides external projecting cams 33a, best seen in FIG. 6, that are complementary to the cams 31d and positioned so that when the closure neck 33 is secured into the orifice 29 by partial (e.g. 60 degree) rotation in a ‘bayonet’ connection, the lower surface 33b of the neck 33 sealingly engages against the ledge 31c of the orifice 29 around its periphery and prevents radioactivity from escaping through the orifice 29. The cap closure 30b attaches to the collar 30a in a compressive motion, such that the container closure 14 is gripped by the annulus 35 of the closure 30b. Although a bayonet fitting arrangement is a particularly convenient means of compressively attaching the cap closure 30b to the collar 30a, these components may be attached together in any other suitable manner that provides a compressive motion of the cap closure 30b relative to the collar 30a, for example by threading. Also, in the embodiment shown the body 22 and cap 30 have a cylindrical exterior, which simplifies the provision of a bayonet connection, however any other convenient configuration may be used with a closure mechanism suitable for substantially preventing leakage of radioactivity from the pig 20. To improve the gripping action of the cap closure 30b compressed against the collar 30a, the somewhat resilient grip 50 may be disposed in the orifice. In the embodiment shown the grip 50 comprises a flange 51 supporting spaced apart fingers 54 that form a circle complementary to the inner wall of the annulus 35, as best seen in FIG. 6. The fingers 54 each have a substantially vertical component extending upwards from the flange 51 and a substantially horizontal component extending inwards from the end of the substantially vertical component thereby to overlap the container closure 14 to a degree as illustrated. In this embodiment the annulus 35 projects from the lower edge 33b of the closure neck 33 into the narrower portion of the orifice 29 in a clearance fit, as shown in FIG. 6, and instead of engaging the container closure 14 directly the annulus 35 defines a recess 35a adapted to engage the grip 50, best seen in FIGS. 6 to 10. In particular, when the cap closure 30b is attached to the collar 30a the annulus 35 compressively engages the fingers 54 of grip 50 to collapse the fingers 54 toward each other against their tendency to remain substantially vertical (that is, to tilt fingers 54 inwardly against their bias) and grip the container closure 14, as shown in FIG. 12. When the cap closure 30b is disengaged from the collar 30a the annulus 35 does not compress the fingers 54 inwards against the container closure 14 thus permitting fingers 54 to spread apart again as per the resiliency to remain substantially vertical (that is to enable fingers 54 to tilt outwardly again to the substantially vertical orientation to which they are biased) enabling the top of container closure 14 to be more exposed through the orifice. The grip 50 may be formed from a semi-compressible material such as plastic (such as a thermoplastic such as Delrin™ available from Dupont Corporation of Wilmington, Del., U.S.A. or polypropylene) or silicone, and has an external profile allowing it to fit snugly within the recess 35a of the annulus 35, and an internal profile allowing the closure 14 of the biohazardous container 10 to fit snugly within the grip 50, as shown in FIG. 12. The grip 50 may be provided with a pattern of openings, increasing the overall compressibility of the grip 50 and reducing its cost. The lower end of the annulus 35 has a slightly diverging wall which is drawn downwardly against the grip 50 as the collar 30a is engaged to the body 22, compressing the grip 50 slightly. The grip 50 thus provides a buffer between the incompressible interior surface of the annulus 35 and the container closure 14, which in the example shown is a stopper engaged with the neck of the container 12 in an interference fit thereby capping the container 12. This both allows the closure 14 to be held securely by the cap 30 and, where the biohazardous container 10 is made of glass, potentially avoids breakage. As in the embodiment illustrated the grip 50 may be frictionally and secured to the collar by lugs 52 projecting into complementary bores 31e formed in the lower edge of the neck 31 of the collar 30a thereby to inhibit rotation and translational exit from the bores 31e. In other embodiments (not shown) the periphery of the flange 51 may snap-fit onto the recess 37 formed in the bottom surface of the collar 30a (see FIG. 6), for example by proving a slight reverse-chamfer in the recess wall so it converges toward the lower limit of the collar 30a, retaining the flange 51, which avoids having to line up the lugs 52 with bores 31e. The grip 50 can be supplied in a single-use sterile package for the plastic piece, or can be pre-loaded to vial and both sterilized together. Different sizes of vial would dictate a corresponding change in the diameter of the compartment 24, but such vials tend to have a standard neck and same septum circumference and in such cases the same size of cap 30 and grip 50 can be used. In the case of the radiopharmaceutical pig 20 shown, the assembled cap 30 and body 22 thus provide a radioactively-shielded compartment 24, for shielding the radioactive contents of the radiopharmaceutical container 10 contained when sealed into the radiopharmaceutical compartment 24. In the embodiment shown the compartment 24 is defined by a cavity formed largely within the body 22 which is sized to receive the bottle 12 in a close fit, preferably a clearance fit but alternatively an interference fit, however the compartment 24 may be formed by defined by suitably sized and aligned adjoining cavities formed respectively in the body 22 and the cap 30. Thus, when the closure remover 34 is seated over the compartment 24 it closes the cap opening 32 in order to radioactively seal the radiopharmaceutical compartment 24. Also, when the cap 30 is removed from the body 22 it is possible to manipulate the sealed container 10 by handling only the cap 30, thereby shielding the technologist's extremities from radiation. To preserve a radiopharmaceutical pill (not shown), the bottle 12 optionally may be provided with fins (not shown) that confine the pill 2 to an axially central portion of the container 10 and thus reduce the amount of pill surface touching the bottle 12. In use of the embodiment shown, a radiopharmaceutical liquid or solid material (e.g. a pill) is placed into the bottle 12 using conventional techniques and equipment to avoid exposure to staff. A radioisotope solution 2 in a glass bottle 12 is illustrated in FIG. 2. In the case of a liquid radiopharmaceutical product the vial typically arrives already filled with the radioactive liquid. The closure 14 may optionally be designed to accommodate a desiccant or other product-stability material or method (not shown) in order to control the humidity within the container 10. The closure 14 is applied to the container 10 which is then placed into the container compartment 24. The cap 30 is placed on the body 22 of the pig 20 and rotated in the closing direction to engage the cams 25, 31a and to seal the cap 30 tightly to the body 22, confining radioactivity from the pill 2 within the container compartment 24. The pig 20 can then be transported to the patient's facility for administration of the biohazardous material, in the example shown a liquid radioisotope. When the pig 20 arrives at the destination, the pig 20 is taken to a room designed to contain the radioactivity and protect staff, as is conventional. The technician grasps the collar 30a and ensures that the cap closure 30b is fully rotated in the direction that locks it to the collar 30a, clockwise in the embodiment illustrated as indicated by the ‘pick up vial’ arrow in FIG. 1. This lodges the container closure 14 into the annulus 35, where a grip 50 is used squeezing the grip 50 against the container closure 14, to lock the container 10 to the cap 30. The technician then grasps the body 22 and rotates the cap 30 collar (30a and cap closure 30b together) to remove the cap 30 from the body 22 with the container closure 14 lodged in the annulus 35 (or where a grip 50 is used, in the grip 50), and lifts the cap 30 off the body 22 as shown in FIG. 3. Where the biohazardous material is a liquid and the cap 14 of the bottle (typically a vial) 12 provides a septum 14b or other entry orifice for a syringe (not shown), the closure 30b can be removed from the collar 30a to expose the top of the container closure 14 and allow the insertion of a syringe without releasing the vial from the collar 30a. A tungsten insert 60, for example as shown in FIG. 13A, may be provided to replace the cap closure 30b. The insert 60 comprises a head 62 and a neck 64 that fits into the orifice 29 in the collar 30a. In the embodiment illustrated the neck 64 of the insert 60 provides external projecting cams 66 that are complementary to the cams 31d and positioned so that when the insert 60 is secured into the orifice 29 by partial (e.g. 60 degree) rotation in a ‘bayonet’ connection, the lower surface of the neck 64 sealingly engages against the ledge 31c of the orifice 29 around its periphery. The syringe may be inserted into the septum through an injection port 68 extending fully through the insert 60 in axial alignment with the compartment 24 of the body 22. In this embodiment, the injection port 68 is cylindrical and has a single diameter throughout its length. The insert 60 provides enhanced radiation protection while dispensing from multi dose vial (stock) due to its smaller-diameter injection port 68 through a head 62 and neck 64 of tungsten, as well as guidance for a syringe to be inserted centrally into the container 10 through the container closure 14. In alternative embodiments, the injection port may be frustoconical. An alternative tungsten insert 60A is shown in FIG. 13B. In this embodiment, tungsten insert 60A has an injection port 68A that has an upper portion 68A_U extending partway through the insert 60A (substantially the height of head 62A) with a larger maximum diameter than does injection port 68 of insert 60, and a lower portion 68A_L extending from the upper portion 68A_U through the rest of the insert 60A (substantially the height of neck 64A) with a smaller diameter (in this embodiment, similar to the diameter of injection port 68 of insert 60). This larger diameter of the upper portion 68A_U permits the ease of insertion and angling of multiple outlet or inlet conduits (such as other syringes or needles thereof) while also permitting a user sufficient room to insert a syringe for withdrawing contents of the container 10. It will be noted that the thickness of a tungsten neck 64A is suitable for sufficient radiation protection in many instances such that there need not be significant concern about the head 62A accommodating the larger upper portion 68A_U of the injection port 68A rather than providing the additional shielding. In this embodiment, each of upper portion 68A_U and lower portion 68A_L are cylindrical. However, in an alternative embodiment, one or both of upper portion 68A_U and lower portion 68A_L of injection port 68A may be frustoconical in shape. Still further, in another alternative embodiment, the upper and lower portions 68A_U and 68A_L of injection port 68A may be replaced by a single, frustoconical injection port with the widest end having a diameter similar to that shown in FIG. 18B at the upper end of the insert 60A. The container 10 can be released by grasping the collar 30a and fully rotating the cap closure 30b in the direction that unlocks it from the collar 30a, counter-clockwise in the embodiment illustrated as indicated by the ‘release vial’ arrow in FIG. 1. In use, the biohazardous material is placed in the container 10 by the manufacturer, placed in the container compartment 24 of the pig 20, and shipped to the destination. A technician at the destination removes the cap 30 with the container 10 attached, moves the container 10 to a dose calibrator (not shown) and, while grasping the collar 30a, rotates the cap closure 30b to release the container closure 14 and (typically using tongs) insert the container 10 into the dose calibrator to measure (assay) amount of radioactivity. The bottle 12 is vented in the dose calibrator, if required (typically only in the case of radioiodine capsules). The container 10 can then be re-sealed and the closure 14 reinserted into the grip 50. The technician while grasping the collar 30a rotates the cap closure 30b in the locking direction to secure the container closure 14 to the grip 50. The cap 30 is then replaced in the manner described above, and delivered to the patient for administration by a qualified professional. At the patient site, in the case of a liquid the technician removes the cap closure 30b from the collar 30a and secures the insert 60 or insert 60A to the collar 30a by interlocking cams 66 and 25 in a bayonet fashion. The technician then inserts a syringe through the orifice 80 and the septum 14b to aspirate the liquid 2 from the bottle 12. The insert 60 or 60A can then be removed and the cap closure 30b replaced on the collar 30a to shield the residual radioactivity in the bottle 12. The pig according to the invention can be used for any type of radioisotope, including those used for so-called “theranostics.” Although tungsten shields gamma rays effectively, optionally a Lucite (Trademark) or Aluminum tube can be used to line the compartment 24 for materials having high beta emissions, for example to shield beta emissions from a radioisotope such as I-131. Bremsstrahlung occurs as beta particles strike a dense material like tungsten or steel, and the Lucite tube thus serves as a ‘pillow’ to reduce or eliminate bremsstrahlung x-rays. FIG. 14 is a front perspective view of a pig 200 according to an alternative embodiment and a handle assembly 300 for the pig 200. In this embodiment, pig 200 is very similar to pig 20 described above, but the outer dimensions (in this embodiment, diameter) of the body 220 of pig 200 is larger than the outer dimensions of the collar 30a of the cap 30 of pig 200 and thereby presents a ledge extending laterally outwards from below collar 30a to the periphery of body 220. As will be described, handle assembly 300 is configurable for carrying pig 200, for supporting pig 200 during extraction of contents of bottle contained within, and for inhibiting unintended removal of cap 30 particularly during transportation of pig 200. In this embodiment, handle assembly 300 includes an upper collar 310 and a lower collar 320 maintained in a fixed spaced relationship by two struts 330a, 330b located opposite each other with respect to pig 200 and extending between the upper collar 310 and the lower collar 320. Upper collar 310 includes a ring 312 with a central opening 314 and an outer diameter that is slightly larger than the outer diameter of body 220 of pig 200, and a wall 316 depends downwards at right angles to the ring 312 about its periphery. The diameter of the central opening 314 is slightly larger than the diameter of collar 30a so that the upper collar 310 can be associated with the body 220 of pig 200 by being placed atop the body 220 such that the ring 312 of upper collar 310 directly faces the ledge of body 220 with the wall 316 of the upper collar 310 extending down a short distance along the exterior of body 220. In this embodiment, lower collar 320 is identical to upper collar 310, but is oriented upward thereby to be associated with the bottom of body 220 by receiving the bottom of body 220 within its peripheral wall 326. It will be understood that, while upper and lower collars 310, 320 are identical in this embodiment, the lower collar 320 in this embodiment does not really need its own central opening 322 to fulfil its function since the bottom of body 220 does not have a corresponding feature. In this embodiment, upper collar 310 and lower collar 320 are made of Delrin™. Each of struts 330a, 33b is connected at a proximate end to the wall 316 of upper collar 310 and at a distal end to the wall 326 of lower collar 320. In this embodiment, channels 318a, 318b, 328a and 328b in the outer face of the peripheral walls 316, 326 of each of upper and lower collars 310, 320 receives corresponding proximate and distal ends of a strut 330a or 330b, and the proximate and distal ends of the strut 330a or 330b are locked within the corresponding channels 318a, 318b, 328a, 328b with fasteners F. In this way, the upper and lower collars 310, 320 contain body 220 of pig 200 such that it is not separable from the upper and lower collars 310, 320 unless these fasteners F are removed. Each of struts 330a, 330b has an outward-facing threaded aperture along its outward-facing surface and intermediate its proximate and distal ends for receiving the threaded end of a corresponding knob 340a or 340b via a corresponding washer 341a, 341b. A U-shaped handle 350 has elongate arms 352a and 352b each depending from a cross member 354, and each of the elongate arms 352a, 352b has therethrough an elongate channel 356a, 356b. The handle 350 is connectable to the struts 330a, 330b by passing knob 340a, 340b through a respective elongate channel 356a, 356b threading the knobs 340a, 340b into its corresponding threaded aperture in the strut 330a, 330b. In this configuration, if both of the knobs 340a, 340b are not fully threaded into corresponding threaded apertures, they do not compress respective arms 352a, 352b against the corresponding strut 330a, 330b, such that the channel 356a, 356b and correspondingly the handle 350 can be both freely rotated about and freely slid along the corresponding knob 340a, 340b while remaining generally connected to the rest of the handle assembly 300. In this way, the handle 350 can be moved between various rotational and extensional orientations with respect to the body 220 of pig 200. If any or both of the knobs 340a, 340b are tightened so as to press the arms 352, 352b against the struts 330a, 330b, the handle is held frictionally in position and is thereby prevented from rotating or sliding with respect to the struts 330a, 330b. It is preferred that the operator tighten both knobs 340a, 340b when intending to maintain the handle 350 in a particular fixed position with respect to the body 220, since the body 220 and the closure 30, being formed with dense, thick walls of tungsten, can be quite heavy. FIG. 15 is a perspective view of the pig 200 and handle assembly 300 of FIG. 14, with the handle 350 having been slid along knobs 340a, 340b to a position in which the cross member 354 is resting atop the cap 30 of the pig 200. In this position, the handle 350 serves to further inhibit removal of the cap 30 thereby providing an extra measure of security for transportation. Cap 30 cannot be lifted from body 220 while handle 350 is in this position (and knobs 340a, 340b are tightened), even if it is rotated somewhat with respect to body 220. In this respect, body 220 can be rotated somewhat within collars 310 and 320 if urged to do so either manually or during jostling in transportation, because, while handle assembly 300 encapsulates body 220, it is not fastened directly to it in this embodiment. The surface of cross member 354 facing the top of cap 30 is generally smooth such that cap 30 is free to rotate along with body 220 even when handle 350 is in the position shown in FIG. 15. In this way, handle 350 is not easily positioned with respect to cap 30 in a way that will result in handle 350 inadvertently loosening cap 30. In an alternative embodiment, body 220 is non-cylindrical such as square-based and handle assembly 300 is of a complementary shape, thus inhibiting any rotation of one with respect to the other. FIG. 16 is a perspective view of the pig 200 and handle assembly 300 of FIG. 14, with the handle 350 having been slid and rotated along knobs 340a, 340b to a position in which the cross member 354 is underneath and spaced from the bottom of lower collar 320. In this position, handle 350 can be used to hold pig 200 either manually or on a hook (not shown) in preparation for removal of the contents of pig 200. FIG. 17 is an exploded perspective view of the handle assembly 300 for the pig 200 in isolation. In this view, compression washers 341a and 341b, in this embodiment formed of rubber, are viewable. These are positioned adjacent to the threaded apertures in struts 330a, 330b for knobs 340a and 340b in order to improve their grip against handle arms 352a, 352b via their channels 356a, 356b, particularly during jostling in transport but also for handling. FIG. 18 is a perspective top view of an alternative compression member, or grip 500, for assisting in securing a container closure 14 to the cap 30. In the embodiment shown the grip 500 comprises a flange 510 supporting a sleeve 505 that is integrated with and encompasses spaced apart fingers 540 that form a circle complementary to the inner wall of the annulus 35. The fingers 540 each have a substantially vertical component extending vertically with the sleeve 505 from the flange 510 and a substantially horizontal component extending inwards with the sleeve 505 from the end of the substantially vertical component thereby to overlap the container closure 14 to a degree in a similar manner as has been described above with respect to grip 50. Extending between each pair of fingers 540 of grip 500, however, is a respective web 542 integrated also with sleeve 505 that is made of a material as will be described that permits flexibility of the fingers 540 inwards and outwards and accordingly towards and away from each other, while providing a more unitary overall structure for surrounding a container closure 14. In this embodiment, flange 510 is formed of a semi-compressible material such as plastic (such as a thermoplastic such as Delrin™ or polypropylene). In this embodiment, flange 510 is not circular, but is instead substantially a square with significantly rounded corners 512. Furthermore, flange 510, as best seen in the side elevation view of FIG. 19, has a sloped edge S spanning the entire periphery of the flange 510. Both the rounded corners 512 and the sloped edge S contribute to permit flange 510 to be snapped into, and retained frictionally within, corresponding sloped structure at a correspondingly sloped lower edge of the neck 31 of collar 30a of the cap 30. While flange 510 is retained within such a correspondingly sloped lower edge of neck 31, when desired, flange 510 may be manually snapped out of the lower edge of neck 31 of collar 30a for disposal of grip 500 and a new grip 500 snapped into place as a replacement. It will be noted that, unlike grip 50, grip 500 does not have posts 52. However, in an alternative embodiment the combination of such posts and the sloped edge S of flange 510 may be employed. In this embodiment, fingers 540 are formed of the same rigid material as flange 510, while sleeve 505 and webs 542 are formed of a more flexible but resilient material such as silicone that is fused at its boundaries with flange 510 and fingers 540. While a grip 500 of two integrated materials exhibiting the two different properties (rigid and flexible) can be very useful, it can be expensive to manufacture. As such, in alternative embodiments grip 500 may be manufactured from a single material for the sleeve 505, fingers 540 and webs 542 with the relative rigidity and flexibility produced through differing thicknesses at different points throughout the grip 500 of the one material rather than necessarily from different materials. For example, the interfaces between the webs 542 and the fingers 540 and flange 510 may incorporate less of the material than between the fingers 540 and the flange 510 thereby to permit webs 542 to be flexed relative to the flange 510 and fingers 540 more than the fingers 540 can flex relative to the flange 510. In this way, the resilience of fingers 540 with respect to flange 510 can be maintained while reducing the rigidifying effect of the webs 542 between the fingers 540. FIG. 20 is a top plan view of the grip 500, FIG. 21 is a bottom plan view of the grip 500, FIG. 22 is a perspective bottom view of the grip 500, FIG. 23 is a perspective top view, partially sectioned, of the grip 500, FIG. 24 is a perspective bottom view, partially sectioned, of the grip 500, FIG. 25 is another perspective top view, partially sectioned below the horizontal components of the sleeve 505, the fingers 540 and the webs 542, of the grip 500, FIG. 26 is another perspective bottom view, partially sectioned, of the compression member of FIG. 18. The radiopharmaceutical pigs 20 and 200 described and illustrated are particularly suitable for transporting radioactive substances such as liquid and solid radiopharmaceuticals due to the radioactivity-shielding character of the container 24, but can be adapted to transport other biohazardous products and materials without the use of radioactivity shielding by hermetically sealing the container 24. Various embodiments of the present invention comprising been thus described in detail by way of example, it will be apparent to those skilled in the art that variations and modifications may be made without departing from the invention. The invention includes all such variations and modifications as fall within the scope of the appended claims. For example, while embodiments described herein involve the compartment 24 of body 22 or body 220 being dimensioned to receive only a container of the biohazardous material, embodiments are contemplated in which the compartment 24 is dimensioned to receive a container in addition to a sponge, such as a cellulose sponge, for physically absorbing liquid originally contained within the received container should it escape from the container during transportation or other handling. Some regulators require that there be provided a quantity of sponge that is capable of absorbing twice the volume of liquid to be contained within the container. Such a cellulose sponge may be formed as a slab and positioned at the bottom of compartment 24 underneath the container, but may alternatively be formed as a cup having a bottom and a sleeve dimensioned to receive the container and, in turn, to be received within compartment 24. The cellulose sponge slab or sleeve would be a consumable. Furthermore, while handle assembly depicted and describe herein has two struts, alternatives are contemplated having more than two struts, or other structures for encapsulating the body within the handle assembly. Still further, very thin layers of rubber or other frictional material may be placed at the interfaces between collar 30a and cap closure 30b and collar 30a and body 22 in order to resist inadvertent relative movements when being transported to thereby resist inadvertent exposure to the contents of the container 10. FIG. 27 is a perspective top view of another alternative compression member, or grip 600, for assisting in securing the container closure 14 to the cap 30. In the embodiment shown the grip 600 comprises a flange 610 being maintained in a spaced relation with an annulus 620 by pillars 630 extending between the annulus 620 and the flange 610. Spaced-apart pivotable grip components 640 are supported by annulus 620 and extend downwards from the annulus 620 between respective pillars 630 towards, but not into contact with, the flange 610. The pivotable grip components 640 are resiliently compressible inwardly against a container 10 and its closure 14 by compressive engagement of a complementary annulus 35 of the pig 20 into which the compression member 600 is dimensioned to be inserted, since the ends of the grip components 640 terminate between the annulus 620 and the flange 610 and are thus unattached. The outward-facing sides of the pivotable grip components 640 each incorporate ramps 642 that engage the complementary annulus and progressively urge the pivotable grip components 640 inwards towards a container 10 as grip 600 is, along with a container 10, urged further into annulus 35. Pillars 630 include additional buttresses 632 at their interfaces to flange 610 in order to strengthen their interconnection. The pivotable grip components 640 and pillars 630 encircle a closure 14 and part of a neck 12 of a container 10 received within the interior of flange 610. This interaction with a closure 14 and part of a neck 12 of a container 10 is similar to that shown between grip 50 and container 10 in FIG. 11, except that, with grip 50, fingers 54 overlie the closure 14 whereas with grip 600, annulus 620 overlies the vial cap and the compressible grip components 640 do not. Also, because pivotable grip components 640 depend from annulus 620 between the pillars 630 only partway, compression member 600 is, via the pivoting of the pivotable grip components 640, therefore able to apply more even force along the surface of a crimped vial closure 14. This provides an improved grip, and improved handling and radiation safety. In use, the grip 600 is placed atop of a crimped-top of a vial, and the corresponding portions of the pig in the annulus 35 of the pig that interact with the ramps 642 along the outward-facing sides of the pivotable grip components 640 as the compression member 600 (along with the container 10) are inserted therein cause the pivotable grip components 640 to move inwardly to grip the closure 14 of the container 10 and, in some embodiments, also contact the glass of the container 10. Grip 600 also serves as a spacer for between these portions of the container 10 and the pig 20. In this embodiment, the above-described components of grip 600 are formed of a semi-compressible material such as plastic (such as a thermoplastic such as Delrin™ or polypropylene). In this embodiment, grip 600 is a single-piece component—a unitary structure—formed by machining. Furthermore, in this embodiment, flange 610 is not circular, but is instead substantially a square with significantly rounded corners 612. Furthermore, flange 610, as best seen in the side elevation view of FIG. 28, has a sloped edge S spanning the entire periphery of the flange 610. Both the rounded corners 612 and the sloped edge S contribute to permit flange 610 to be snapped into, and retained frictionally within, corresponding sloped structure at a correspondingly sloped lower edge of the neck 31 of collar 30a of the cap 30. While flange 610 is retained within such a correspondingly sloped lower edge of neck 31, when desired, flange 610 may be manually snapped out of the lower edge of neck 31 of collar 30a for disposal of grip 600 and a new grip 600 (or grip 500, or grip 50) snapped into place as a replacement. It will be noted that, unlike grip 50, and like grip 500, grip 600 does not have posts 52. However, in an alternative embodiment the combination of such posts and the sloped edge S of flange 610 may be employed. FIG. 28 is a side elevation view of the compression member 600 of FIG. 27. FIG. 29 is a top plan view of the compression member 600 of FIG. 27. FIG. 30 is a perspective top view, partially sectioned, of the compression member 600 of FIG. 27. FIG. 31 is a perspective bottom view of the compression member 600 of FIG. 27. The alternative compression member 600 of FIGS. 27 through 31 is a single-use disposable insert that, when correctly positioned, prevents shield elements from contacting a contained vial septum/crimp (such as container 10), so as to enhance aseptic technique. Furthermore, the design using an annulus 620 concentrates a needle towards the centre of the vial septum in the closure 14. This avoids or prevents coring of the septum and piercing of the aluminum crimp of a closure 14, thereby to preserves integrity (sterility) of the radiopharmaceutical contained within a container 10. An article regarding challenges due to coring and fragmentation may be found at https://www.apsforg/newsletters/html/2013/winter/12coring.htm.
	description	This application is a divisional application of U.S. Application Serial No. 12/537,452, filed Aug. 7, 2009, the contents of which are incorporated herein by reference. The present application claims priority from Japanese Patent application serial no. 2008-205061, filed on Aug. 8, 2008, the content of which is hereby incorporated by reference into this application. The present invention relates to a core of a light water reactor and a fuel assembly, and more particularly, to a core of a light water reactor and a fuel assembly suitable for a boiling water reactor. When actinide nuclide, which has many isotopes and is included in a nuclear fuel material in a fuel assembly loaded in a core of a light water reactor, burns in a core, the actinide nuclide to transfers among isotopes in succession by nuclear transmutation such as nuclear fission and neutron absorption. Since odd-numbered nucleus that has a large nuclear fission cross section with respect to a resonance and thermal neutrons, and even-numbered nucleus that undergoes fission only for fast neutrons are present as the actinide nuclide, in general, present ratios of the isotopes present in the actinide nuclides included in the fuel assembly largely change as the actinide nuclides burn. It is known that this present ratio change depends on the neutron energy spectrum at the position at which the fuel assembly is loaded in the core. Current light water rectors use slightly enriched uranium as nuclear fuel. However, since the natural uranium resource is finite, it is necessary to successively replace fuel assemblies used in the light water reactor with recycled fuel assemblies including a nuclear fuel material which is formed by enriching depleted uranium, which is a residual after uranium enrichment, with the transuranic nuclide (hereinafter referred to as TRU) extracted from spent fuel assemblies in the light water reactor. TRU needs to be recycled as a useful resource over a very long period predicted to be necessary for commercial reactors, and during this period, the amount of TRU needs to always increase or to be maintained nearly constant. JP 3428150 B describes technology to implement a breeder reactor in which the amount of fissionable Pu is increased or maintained nearly constant in light water reactors that occupy most of the current commercial reactors. In a light water reactor in which the breeder reactor described in JP 3428150 B and R. TAKEDA et al., Proc. of International Conference on Evaluation of Emerging Nuclear Fuel Cycle Systems. GLOBAL '95 Versailles, France, September, 1995, P. 938 is became a reality, a plurality of fuel assemblies, each of which has a hexagonal transverse cross section, are disposed in the core, each fuel assembly being formed by closely arranging a plurality of fuel rods in a triangular grid. In the core of this light water reactor, the amount of water around the fuel rods is lessened due to the close arrangement of the fuel rods, and thereby the ratios of resonant energy neutrons and fast energy neutrons are increased. In addition, the height of a mixed oxide fuel section of the TRU is reduced and blanket zones loaded with depleted uranium are disposed above and below the mixed oxide burning part so as to maintain a negative void coefficient, which is a safety criterion. The core is formed in two stacked stages by applying the concept of a parfait-type core described in G. A. Ducat et al., Evaluation of the Parfait Blanket Concept for Fast Breeder Reactors, MITNE-157, January, 1974, thereby a breeding ratio of 1 or more is ensure, keeping the economy. To recycle TRU, the reprocessing of spent fuel is indispensable. Due to a fear that consumer TRU is diverted to weapons of mass destruction, there has been an increasing demand for nuclear non-proliferation and thereby restrictions on TRU recycling have been severe. It is certain that an electric power generating system superior to a fission reactor is put into practical use on some day in the future. At that time, the value of TRU is lowered from a very useful fuel equivalent to enriched uranium to is a cumbersome long-lived waste material. Accordingly, the most important object in nuclear power development is to establish a TRU disposal method. R. TAKEDA et al., Proc. of International Conference on Advanced Nuclear Fuel Cycles and Systems. GLOBAL '07 Boise, USA, September, 2007, P. 1725 suggests a TRU disappearance reactor and a light water breeder reactor for recycling TRU while the present ratio of each isotope of the TRU is maintained nearly constant to achieve multiple recycling, in which recycling, the TRU obtained by reprocessing a spent nuclear fuel is reused as a new nuclear fuel to repeat recycling. This light water breeder reactor can recycle in a state in which the amount of TRU is maintained constant or increased, and it has a core loaded with fuel assemblies with a high burnup and high nuclear proliferation resistance. The TRU disappearance reactor is a reactor for reducing TRU through nuclear fission by successively reducing and aggregating the TRU until all the TRU is reduced by undergoing fission to the amount of the TRU being loaded in the last one core to prevent the TRU from becoming a long-life radioactive waste at the end of the light water reactor's purpose. A light water reactor for recycling TRU was achieved in R. TAKEDA et al., Proc. of International Conference on Evaluation of Emerging Nuclear Fuel Cycle Systems. GLOBAL '95 Versailles, France, September, 1995, P. 938; such that a recycle reactor effectively uses a seed fuel, in which the amount of TRU is maintained constant, with a sufficient safety margin to meet design standards for abnormal transience and accidents; and that the recycle reactor can stabilize supply of energy for a long period of time by burning all depleted uranium, and by making all the TRU undergo fission, the recycle reactor prevents the TRU from becoming a long-life radioactive waste at the end of the nuclear fission reactor's purpose when the TRU has become no longer necessary. On the other hand, there has recently been a movement to tighten up the attitude toward safety; consequently, a core having a high safety potential is expected, having a sufficient safety margin for responding to an accident outside the design standards (Anticipated Transient Without Scram, or ATWS) such as, for example, a compound event where the core flow rate suddenly drops for some reason and all control rods cannot be inserted into the core. An object of the present invention is to provide a core of a light water reactor and a fuel assembly which can further increases a safety margin without sacrificing an economic efficiency of the light water reactor. The present invention for attaining the above object is characterized in that, in a fuel assembly, which was loaded in a core, with a burnup of 0, a ratio of Pu-239 in all transuranium nuclides included in the fuel assembly is in a range of 40 to 60%, sum of heights of a lower blanket zone and an upper blanket zone formed in the core is in a range of 250 to 600 mm, and the height of the lower blanket zone is in a range of 1.6 to 12 times the height of the upper blanket zone. According to the present invention, a safety margin can be sufficiently maintained even with the occurrence of a compound event, in which core flow rate is suddenly dropped for some reason and all control rods cannot be inserted into the core at the same time during the operation of the light water reactor, beyond design standards. When such a compound event occurs, void fraction in the core rapidly increases, a boiling start point of coolant being slightly sub-cooled, being supplied into the core from below the core, shifts toward a lower end of the core, and the power distribution in the axial direction of the core also shifts toward the lower end of the core. Therefore, excess neutrons are shifted toward the lower end of the core. These shifted excess neutrons can be absorbed by neutron absorber in a neutron absorber filling-zone, an upper end of which is positioned in the vicinity of the lower end of the core, of a safety rod. As a result, power of the light water reactor can be automatically reduced to the power at which the fuel assemblies in the core can be cooled by the capacity of the coolant suppliable from an emergency high-pressure core flooder. Thus, a sufficient safety potential can be maintained upon the occurrence of a compound event beyond the design standards. In this way, the present invention can improve the safety margin without sacrificing the economic efficiency of the light water reactor even with the occurrence of the compound event. The above object can also be achieved when, in a fuel assembly, which was loaded in a core, with a burnup of 0, a ratio of Pu-239 in all transuranium nuclides included in the fuel assembly is at least 5% but less than 40%, a height of an upper blanket zone is in a range of 20 to 100 mm, and a lower end of a fissile zone matches a lower end of the core. According to the present invention, a safety margin can be further increased without sacrificing an economic efficiency of a light water reactor. Cooling water (coolant) for cooling fuel assemblies in a core of a BWR is supplied into a core from below as sub-cooled water at approximately 10° C., and as cooling the fuel assemblies, it becomes a two-phase flow including saturated water and steam-and-water. This cooling water becomes a two-phase flow with a void volume fraction of approximately 60 to 80% at an outlet of the core. Thus, a distribution of hydrogen atoms, which significantly contribute to moderate neutrons, in the axial direction of the core decreases from the lower portion toward the upper portion of the core. For this reason, if the fuel assemblies having one zone of axial enrichment distribution are loaded into the core of the BWR, a large power peak is formed in the lower region of the core. When the flow rate of the cooling water in the core is reduced for some reason, the boiling start point of the cooling water is shifted further downward than that of when the reactor is operated at its rated power and rated core flow rate, and the power peak is shifted further downward in the core as well. The inventors have thoroughly considered the above characteristics of a core of a BWR and found out that, in a core of a light water reactor loaded with a plurality of fuel assemblies including nuclear fuel material obtained by reprocessing, the ratio of Pu-239 in all TRU included in fuel assemblies at the time of zero burnup is in a range of 40 to 60%, when an upper end of a neutron absorber filling-zone of a control rod is disposed in the vicinity of a lower end of the core, a safety margin can be further increased without sacrificing the economic efficiency of the light water reactor by applying any of the following constitutions, i.e., (1) making the sum of the heights of an upper blanket zone and a lower blanket zone in the core in a range of 250 to 600 mm and at the same time, making the height of the lower blanket zone in a range of 1.6 to 12 times that of the upper blanket zone, (2) making the height of the lower blanket zone higher than that of the upper blanket zone and at the same time, making the height of the upper blanket zone in a range of 30 to 105 mm, and (3) making the height of the lower blanket zone higher than that of the upper blanket zone and at the same time, making the height of an upper fissile zone including Pu for the core higher than the height of a lower fissile zone including Pu within a range of 10 to 25 mm. In other words, by applying any of the constitutions (1), (2), and (3), even with the occurrence of a compound event, which is beyond design standards, such as the coolant in the core is lost for some reason and all the control rods cannot be into the core for some reason, excess neutrons in the core are automatically absorbed by the neutron absorber disposed in the lower end of the core, because a void fraction in the core is rapidly increased when the flow rate of the coolant supplied to the core (a core flow rate) is suddenly decreased, and the power distribution in the axial direction of the core is shifted toward the lower end of the core. For this reason, reactor power is automatically reduced to the reactor power at which cooling can be achieved by the flow rate of the coolant supplied to the core by an emergency high-pressure core flooder that is automatically activated in the case of an emergency. As described above, the inventors have newly found out that a safety potential in a core of a light water reactor can be enhanced by applying any of the constitutions (1), (2), and (3). In addition, the inventors have found out that the safety margin can also be increased, as done in above, by (4) making the height of the lower blanket zone higher than that of the upper blanket zone and at the same time, disposing neutron absorbing material to the position where excess neutrons generated at the time of the accident gather. The safety margin can be further improved by combining some of the constitutions (1), (2), (3), and (4). For example, when the constitutions (1) and (2) are combined, the safety margin will be larger than that of the constitution (1) alone; and when the constitution (3) is additionally combined to the combination of the constitutions (1) and (2), the safety margin will be further improved than that of the constitutions (1) and (2) combined. This can be said to the other combinations in which the constitution (2), (3), or (4) is combined to other two constitutions. When (4) is additionally combined to the combination of the constitutions (1), (2), and (3), the safety margin will be the largest among the combinations including some of the constitutions (1) to (4). Furthermore, the inventors have newly found out that, in a core of a light water reactor loaded with a plurality of fuel assemblies including nuclear fuel material obtained by reprocessing, the ratio of Pu-239 in all TRU included in fuel assemblies at the time of zero burnup is at least 5% but less than 40%, when the upper end of a neutron absorber filling-zone of a control rod is disposed in the vicinity of the lower end of the core, a safety margin can be further increased without sacrificing economic efficiency and design target performance of TRU multi-recycling by applying any of the following constitutions, i.e., (5) matching a lower end of a fissile zone in the core to the lower end of the core and at the same time, making the height of an upper blanket zone in a range of 20 to 100 mm, and (6) having the upper blanket zone and at the same time, making a height of an upper fissile zone in the core higher than a height of a lower fissile zone within a range of 10 to 25 mm. The safety margin is further improved by combining the constitution (6) to the constitution (5) compared to the light water reactor core having either constitution (5) or (6) alone. A goal of the present invention is to improve safety of a recycling-type light water reactor which utilizes nuclear fuel material containing TRU obtained by reprocessing. Such present invention is made to maintain safety even with the occurrence of multiple accidents beyond design standards and to allow TRU multi-recycling to continue, when the performance as a breeder reactor (a light water breeder reactor) is to be improved in the light water reactor shown in JP 3428150 B, and when the TRU considered to be disposed of as a long-life radioactive waste when it is no longer necessary, is to be utilized as nuclear fuel material until all the TRU, except for those being loaded in the last one core, are made to undergo fission. A reactor core of a light water reactor having improved performance as a breeder reactor is described. For example, a light water breeder reactor that yields a fissile Pu residual rate of 1 or more in a BWR was first achieved in JP 3428150B. To realize the breeder reactor in a light water reactor, neutron energy in the core must be maintained at a high level. However, since the mass of a hydrogen atom forming the water used as a coolant in the light water reactor is small compared to that of Na generally used as a coolant in a breeder reactor, the energy loss of neutrons at one collision becomes large in the light water reactor. Thus, it is necessary to reduce a ratio of coolant per unit volume of nuclear fuel material in the light water reactor. When a nuclear fuel material having a ratio of Pu-239 in all TRU in a range of more than 60% is recycled, the following problems may arise, i.e., (a) a capacity for cooling the nuclear fuel material in the core is not enough, (b) a burnup of the fuel assemblies is reduced, impairing the economic efficiency of the fuel, and (c) a gap between fuel rods disposed in the fuel assembly become too narrow, causing the production of the fuel assembly to be difficult. When a nuclear fuel material having a ratio of Pu-239 in all TRU in a range of less than 40% is recycled, the following problems may arise, i.e., (d) a ratio of odd-numbered nuclides having a larger nuclear fission cross section is reduced compared to that of even-numbered nuclides having a smaller nuclear fission cross section, causing a fissile Pu residual rate of 1 or more to be difficult to attain, and (e) the core becomes large and the void coefficient, which is a safety indicator, is worsened. Therefore, in a light water breeder reactor, the ratio of Pu-239 contained in all the TRU should be within a range of 40 to 60%. Next described is a reactor core of a light water reactor (a TRU disappearance reactor) which allows the TRU being considered to be disposed of as a long-life radioactive waste when it is no longer necessary, to be utilized as nuclear fuel material until all the TRU except for those being loaded in the last one core is made to undergo fission. The inventors have thought out to reduce the TRU by nuclear fission when the TRU is no longer needed, by integrating the TRU dispersed in many cores based on the amount of reduction of TRU, and leaving the TRU only in one core at the end. At this time, when a nuclear fuel material having a ratio of Pu-239 in all the TRU in a range of at least 40% is recycled to prevent the TRU from becoming a long-life radioactive waste, it takes too long to integrate the TRU in one core since the speed of the TRU reduction is slow. When a nuclear fuel material having a ratio of Pu-239 in all the TRU in a range of less than 5% is used for recycling, the core becomes large and the void coefficient is worsened. Therefore, in a TRU disappearance reactor, the ratio of Pu-239 contained in all the TRU should be set to a range of at least 5% but less than 40%. Now, an overview of a parfait-type reactor core is described. The parfait-type reactor core uses a fuel assembly having a lower blanket zone, a lower fissile zone, an inner blanket zone, an upper fissile zone, and an upper blanket zone disposed in this order from the lower end portion to the upper end portion, as a new fuel assembly (having a burnup of 0) for loading. In the parfait-type reactor core, a lower blanket zone, a lower fissile zone, an inner blanket zone, an upper fissile zone, and an upper blanket zone are formed from the lower end portion to the upper end portion as well. The lower and upper fissile zones include TRU oxide fuel (or mixed oxide fuel of TRU oxide and uranium oxide). The present invention is intended for the above recycling-type light water reactor and the light water reactor core. The results of study done by the inventors are described below. First of all, the results of the study by the inventors regarding a core of a light water breeder reactor are described below. In the description, a BWR core with an electric power of 1350 MW and a breeding ratio of 1.01, loaded with 720 fuel assemblies, each of which having 271 fuel rods, in the core is used as an example of the core of the light water breeder reactor. While the cores disclosed in JP 3428150 B; R. TAKEDA et al., Proc. of International Conference on Evaluation of Emerging Nuclear Fuel Cycle Systems. GLOBAL '95 Versailles, France, September, 1995, P. 938; and R. TAKEDA et al., Proc. of International Conference on Advanced Nuclear Fuel Cycles and Systems. GLOBAL '07 Boise, USA, September, 2007, P. 1725 can safely and sufficiently respond to abnormal transience and accidents within design standards, this BWR core cannot always respond in a sufficient manner upon the occurrence of a compound event such as the core flow rate being suddenly decreased for some reason and in addition, all control rods being inoperable, when such an event is currently regarded as beyond the design standards. In some cases, TRU recycling may have to be stopped in the middle. In other words, multiple recycling may not be continued. In order to continue TRU recycling while maintaining a sufficient safety potential in the above BWR core, the void coefficient must be maintained within to a predetermined range. The inventors have studied a method to improve a safety margin for the core of the light water reactor which has the lower and upper blanket zones, is loaded with a plurality of fuel assemblies including nuclear fuel material obtained by reprocessing. In each of these fuel assemblies, the ratio of Pu-239 in all TRU included in the fuel assembly at the time of zero burnup is in a range of 40 to 60%. As a result of the study, the inventors have newly found out a way to realize TRU multi-recycling such that, when the core flow rate is suddenly dropped for some reason specific to the function of the BWR, the void fraction in the core rapidly rises and the boiling start point of the coolant being slightly sub-cooled and flowing into the core from below, shifts to the lower end side of the core, which makes the power distribution in the axial direction of the core shift to the lower end side of the core; thus, by disposing neutron absorbing material in the vicinity of the lower end of the core, a sufficient safety potential can be maintained upon the occurrence of multiple accidents. Based on this knowledge, the inventors have newly found out that the safety potential can be enhanced while the breeding ratio of the TRU is maintained by adapting any of the above (1), (2), (3), and (4). In the core of the light water reactor discussed here, the control rods are inserted into the core from below. In FIG. 1, a property 1 shows an average power distribution in the axial direction of the core in the core having a fissile Pu breeding rate of 1.01 during its rated power operation, and a property 2 shows an average power distribution in the axial direction of the core when the core flow rate is reduced to 4 kt/h which is a flow rate of the coolant from an emergency high-pressure core flooder. In FIG. 2, a property 3 shows an average void fraction distribution in the axial direction of the core corresponding to the property 1, and a property 4 shows an average void fraction distribution in the axial direction of the core corresponding to the property 2. Due to the sudden drop of the core flow rate from a rated value of 21 kt/h to 4 kt/h, the void fraction distribution rapidly rises from the property 3 to the property 4, and at the same time, the boiling start point shifts to the lower end side of the core. This also shifts the power distribution in the axial direction of the core to the lower end side of the core from the property 1 to the property 2, as can be seen. When the core flow rate drops in such an extreme way, a large power peak may be generated in a reflector (cooling water) in the lower portion of the core, and positive reactivity may be introduced into the core in some cases. In the reactor cores disclosed in JP 3428150 B; R. TAKEDA et al., Proc. of International Conference on Evaluation of Emerging Nuclear Fuel Cycle Systems. GLOBAL '95 Versailles, France, September, 1995, P. 938; and R. TAKEDA et al., Proc. of International Conference on Advanced Nuclear Fuel Cycles and Systems. GLOBAL '07 Boise, USA, September, 2007, P. 1725, each safety rod, which is one type of control rod being withdrawn from the core during the rated power operation, is held, while being withdrawn, at the position which does not affect the core by introducing negative reactivity (a position 20 to 30 cm below the lower end of the core) as usually done in a relatively low-height core having a height of 2 m or less. Thermal neutron flux distribution in the axial direction of the core in this state is shown as a property 5 in FIG. 3. Thus, an upper end of a neutron absorber filling-zone of the safety rod being withdrawn below the lower end of the core during the reactor operation as described above, is positioned at the lower end of the core, so that this safety rod can absorb excess neutrons shifting to the lower portion of the core when the core flow rate is suddenly decreased. The thermal neutron flux distribution in the axial direction of the core at this time is shown as a property 6 in FIG. 3. However, positioning the upper end of the safety rod to the lower end of the core reduces reactivity of the core. As a countermeasure to solve a lack of this reactivity, the height of fissile zones may be increased. However, in this proposal, the volume ratio of blanket zones per unit volume of the fissile zones is reduced which reduces the breeding ratio of fissile Pu, consequently, the core fails to meet a design target for the fissile Pu breeding ratio. In order to increase the fissile Pu breeding ratio, the height of each of the upper and lower blanket zones in the core must be further increased. The increase in the heights of these zones causes a neutron leak ratio in the axial direction of the core to decrease and the void coefficient, which is an important safety indicator, is worsened. The results of the study done by the inventors regarding a core of a light water reactor loaded with a plurality of fuel assemblies, in which the ratio of Pu-239 in all TRU included in each of these fuel assemblies at the time of zero burnup is in a range of 40 to 60%, including nuclear fuel material obtained by reprocessing, showed that when neutron absorbing material is disposed in the vicinity of the lower end of the core, the void coefficient can be prevented from getting worse by decreasing the height of an upper blanket zone and increasing the height of a lower blanket zone. In the light water breeder reactor, the above-mentioned vicinity of the lower end of the core means an area between the lower end of the core and a position, for example, 5 mm below the lower end of the core, and when the lower blanket zone is formed in the core, this lower blanket zone is also included in the vicinity of the lower end. By making the height of the lower blanket zone higher than that of the upper blanket zone, that is, by making sum of the heights of the upper and lower blanket zones 250 mm or higher and at the same time, making the height of the lower blanket zone 1.6 times or more that of the upper blanket zone, a breeding ratio of 1.01 can be maintained, all restrictive conditions are met, and at the same time, even with the occurrence of a compound event beyond design standards such as the core flow rate is substantially dropped for some reason and all control rods become inoperable, power can be automatically reduced to the power at which the fuel assemblies in the core can be cooled by the capacity of the coolant suppliable to the core by an emergency high-pressure core flooder as shown in FIG. 4. For this reason, a safety margin can be improved in the core of the light water reactor loaded with the fuel assemblies including the nuclear fuel material obtained by reprocessing, the ratio of Pu-239 in all the TRU included in each of these fuel assemblies at the time of zero burnup is in a range of 40 to 60%. When the sum of the heights of the upper and lower blanket zones is over 600 mm or when the height of the lower blanket zone is more than 12 times that of the upper blanket zone, the ratio of Pu-239 contained in all the TRU in the nuclear fuel material in the spent fuel assembly to be taken out from the core will be higher than the ratio of that contained in a new fuel assembly with a burnup of 0. For this reason, when the core flow rate is increased to keep these values in the same range, a pressure loss in the core will exceed the design standard, which will make the structure design of the fuel assemblies difficult. Thus, the sum of the heights of the upper and lower blanket zones should be within a range of 250 to 600 mm. In FIG. 5, a property 31 shows a change in the void coefficient when the height of the upper blanket zone is varied in the core of the light water reactor having a fissile Pu breeding ratio of 1.01, and a property 32 shows a ratio of the height of the lower blanket zone to the height of the upper blanket zone. As shown in FIG. 5, it became clear that when the height of the upper blanket zone was 105 mm or lower, the height of the lower blanket zone would become 1.6 times or more the height of the upper blanket zone and the void coefficient would become more negative than −2×10−4 Δk/k/% void. By making the negative absolute value of the void coefficient larger, power can be automatically reduced to the power at which the fuel assemblies can be cooled by the capacity of the coolant suppliable by the emergency high-pressure core flooder even with the occurrence of a compound event beyond design standards such as the core flow rate being substantially reduced, that is, the void fraction in the core being substantially increased and all control rods being inoperable. When the height of the upper blanket zone is less than 30 mm, the power of fuel pellets located near the upper end of the upper blanket zone, being substantially affected by thermal neutron flux in the upper reflector, will exceed the design standard. Thus, the height of the upper blanket zone is set within a range of 30 to 105 mm. In FIG. 6, a property 33 shows the sum of the heights of the upper and lower blanket zones when the height of the upper blanket zone is varied. It became clear that by making the height of the upper blanket zone 105 mm or lower, the sum of the heights of the upper and lower blanket zones would be 250 mm or higher. Furthermore, when the reactor is operated while the upper end of the neutron absorber filling-zone of the safety rod is positioned in the vicinity of the lower end of the core, there may be a case that boron-10, which is a neutron absorber included in the safety rod, is used up very quickly. For this reason, in some cases, it is also useful to dispose pellets including a neutron absorbing material such as boron, gadolinia, Dy, Sm, Eu, etc. below the lower blanket zone in the fuel rod included in each fuel assembly. The height of the lower blanket zone is made higher than that of the upper blanket zone and the height of the upper fissile zone including TRU in the core is made higher than the height of the lower fissile zone including TRU within a range of 10 to 25 mm. By making the height of the upper fissile zone at least 10 mm higher than the height of the lower fissile zone, the safety margin of the core can be improved even with the occurrence of the above compound event. When the height of the upper fissile zone is more than 25 mm higher than the height of the lower fissile zone, the power in the upper fissile zone will become too high, exceeding the design standard for the power. Next, the results of the study done by the inventors regarding the core of the TRU disappearance reactor disclosed in R. TAKEDA et al., Proc. of International Conference on Advanced Nuclear Fuel Cycles and Systems. GLOBAL '07 Boise, USA, September, 2007, P. 1725 is described. As an example of the core of the TRU disappearance reactor, another BWR core with an electric power of 1350 MW, loaded with 720 fuel assemblies, each of which having 397 fuel rods, is described. When TRU recycling is repeated for the purpose of decreasing TRU, in other words, when loading of the fuel assemblies into the core is repeated for every operation cycle, each of which fuel assemblies includes nuclear material obtained by reprocessing and the ratio of Pu-239 in all the TRU included in each of which fuel assembly at the time of zero burnup is at least 5% but less than 40%, fast neutrons leaked out from the core are moderated due to the neutron moderation effect by hydrogen atoms forming the water in a reflector region in the lower portion of the core, causing a large power peak of thermal neutrons. In order to avoid a problem of the power of fuel pellets, which are located in the vicinity of the lower end of a lower fissile zone of each of the fuel assemblies adjoining each other in the core, exceeding the value of the design standard due to the neutron current continuity condition, a lower blanket zone of about 20 mm high is constructed below the lower fissile zone in the TRU disappearance reactor disclosed in R. TAKEDA et al., Proc. of International Conference on Advanced Nuclear Fuel Cycles and Systems. GLOBAL '07 Boise, USA, September, 2007, P. 1725. By applying any of the above-described methods (5) and (6) for improving the safety potential, which the inventors have found out, it is no longer necessary to provide the lower blanket zone in the lower portion of the core since the occurrence of the power peak of thermal neutrons in the reflector (cooling water) in the vicinity of the lower end of the core can be controlled by positioning the upper end of a neutron absorber filling-zone of a control rod in the vicinity of the lower end of the core. In other words, the fissile zone, or particularly the lower end of the lower fissile zone, matches the lower end of the core. In the TRU disappearance reactor, the above-mentioned vicinity of the lower end of the core for positioning the upper end of the neutron absorber filling-zone of the control rod means an area between the lower end of the core and a position, for example, 5 mm below the lower end. In FIG. 7, a property 11 shows an average power distribution in the axial direction of the core during the rated power operation, and a property 12 shows an average power distribution in the axial direction of the core when the core flow rate is dropped to 4 kt/h which is a flow rate of the cooling water suppliable by the emergency high-pressure core flooder. In FIG. 8, a property 13 shows an average void fraction distribution in the axial direction of the core corresponding to the property 11, and a property 14 shows an average void fraction distribution in the axial direction of the core corresponding to the property 12. Due to the sudden drop in the core flow rate from a rated value of 20 kt/h to 4 kt/h, the average void fraction distribution in the axial direction of the core rapidly increases from the property 13 to the property 14 shown in FIG. 8. At the same time, the boiling start point shifts to the lower end side of the core, causing the power distribution in the axial direction of the core to shift from the property 11 to the property 12 shown in FIG. 7. When the core flow rate drops in such an extreme way, a large power peak is generated in the reflector in the lower portion of the core and positive reactivity may be introduced into the core in some cases. Each safety rod which is being withdrawn from the core during the rated power operation, is held, while being withdrawn, at a position where the safety rod does not affect the core by introducing negative reactivity (example for, a position 30 cm below the lower end of the core), as usually done in a relatively low-height core having a height of 2 m or less. In FIG. 9, a property 15 shows thermal neutron flux distribution in the axial direction of the core in the core provided with a 20-mm-high lower blanket zone without a neutron absorber filling-zone of a safety rod disposed in the vicinity of the lower end. While the upper end of the neutron absorber filling-zone of the safety rod, which is being withdrawn below the lower end of the core during the reactor operation, is positioned at the lower end of the core and when the core flow rate is suddenly decreased, the safety rod can absorb excess neutrons shifting to the lower portion of the core. A property 16 in FIG. 9 shows the thermal neutron flux distribution in the axial direction of the core at that time. As shown in FIG. 10, even with the occurrence of a compound event beyond design standards such as the core flow rate substantially dropping for some reason and all control rods becoming inoperable, power can be automatically reduced to the power at which the fuel assemblies in the core can be cooled by the capacity of the coolant suppliable to the core from the emergency core flooder. For this reason, a safety margin can be improved in the core of the light water reactor loaded with the fuel assemblies including the nuclear fuel material obtained by reprocessing, the ratio of Pu-239 in all the TRU included in each of these fuel assemblies at the time of zero burnup is at least 5% but less than 40%. In the core of the TRU disappearance reactor, a safety margin of the core can be improved by making the height of the upper blanket zone 100 mm or less. However, when the height of the upper blanket zone is less then 20 mm, the power of fuel pellets located near the upper end of the upper blanket zone, being substantially affected by thermal neutron flux in the upper reflector, will exceed the design standard as in the light water breeder reactor. Thus, the height of the upper blanket zone is set within a range of 20 to 100 mm. The reason for making the height of the upper fissile zone higher than the height of the lower fissile zone within a range of 1.0 to 25 mm in the core of the TRU disappearance reactor is the same as the reason for setting these values in the light water breeder reactor. Various embodiments of the present invention are described below in detail with reference to the figures. A core of a light water reactor core according to embodiment 1, which is a preferred embodiment of the present invention, is described below in detail with reference to FIGS. 11 to 19 and Table 1. TABLE 1NuclideComposition (wt %)Np-2370.5Pu-2382.9Pu-23944.0Pu-24036.2Pu-2415.0Pu-2424.9Am-2413.6Am-242M0.2Am-2431.3Cm-2441.0Cm-2450.3Cm-2460.1 A core 20 of a light water reactor in the present embodiment is for generating an electric power of 1350 MW; however, the power scale is not limited to this value. A core having a different power scale, to which the present embodiment can be applied, can be achieved by changing the number of fuel assemblies loaded into the core 20. An overview of a BWR, which is a light water reactor for generating an electric power of 1350 MW and to which the core 20 of the present embodiment is applied, is described based on FIG. 11. A BWR 19 has the core 20, a steam separator 21, and a stream dryer 22 disposed in a reactor pressure vessel 27. The core 20 is a parfait-type core, which is surrounded by a core shroud 25 in the reactor pressure vessel 27. A plurality of control rods 2 is disposed at the positions which allow the control rods 2 to be inserted into the core 20. These control rods 2 are inserted into the core 20 from below. The steam separator 21 is disposed above the core 20, and the steam dryer 22 is disposed above the steam separator 21. A plurality of internal pumps 26 is provided at the bottom portion of the reactor pressure vessel 27, and impellers of each internal pump 26 are disposed in a downcomer formed between the reactor pressure vessel 27 and the core shroud 25. A main steam pipe 23 and a feed water pipe 24 are connected to the reactor pressure vessel 27. The BWR 19 is equipped with a low-pressure core flooder 31 and a high-pressure core flooder 32 as an emergency core cooling system in case the coolant to be supplied to the core is lost for some reason. As shown in FIG. 12, the core 20 is loaded with 720 fuel assemblies 1. One Y-shaped control rod 2 is provided for every three fuel assemblies 1, and 223 control rods 2 are disposed. Approximately ⅙ of the 223 control rods 2 are control rods for adjusting the reactor power (power adjustment control rods) by being inserted into or withdrawn from the core 20 in the BWR 19 during the operation, and the approximately ⅚ remaining are control rods 2 for inserting into the core 20 when the reactor is shutdown (hereinafter referred to as safety rods), which are being withdrawn from the core 20 in the BWR 19 during the operation. The fuel assembly 1 sequentially forms five zones, i.e., an upper blanket zone 5, an upper fissile zone 6, an inner blanket zone 7, a lower fissile zone 8, and a lower blanket zone 9, from the upper end to the lower end in a portion of an active fuel length (see FIG. 17). In the core 20 loaded with the plurality of fuel assemblies 1, five zones are sequentially formed from the upper end to the lower end, i.e., an upper blanket zone 5A formed by the upper blanket zones 5, an upper fissile zone 6A formed by the upper fissile zones 6, an inner blanket zone 7A formed by the inner blanket zones 7, a lower fissile zone 8A formed by the lower fissile zones 8, and a lower blanket zone 9A formed by the lower blanket zones 9 (see FIG. 1). The zones 5A, 6A, 7A, 8A, and 9A are located at the same positions in the axial direction of the core 20 as the zones 5, 6, 7, 8, and 9 of each fuel assembly 1 respectively. In the fuel assembly 1, as shown in FIG. 13, 271 fuel rods 3, each having a diameter of 10.1 mm, are disposed in a regular triangle lattice in a channel box 4 which is a hexagonal tube. The transverse cross-sectional shape of the fuel assembly 1 is hexagonal. A gap between the fuel rods 3 disposed in the fuel assembly 1 is 1.3 mm. A plurality of fuel pellets (not shown) composed of nuclear fuel material, arranged in the axial direction, is disposed in a cladding tube 36 of each fuel rod 3. Nine fuel rods 3 are disposed in a fuel rod row in an outermost peripheral layer. In the fuel rod 3, as shown in FIG. 14, the plurality of fuel pellets prepared by using the nuclear fuel material obtained by reprocessing is filled in the cladding tube 36 whose a lower end portion and an upper end portion are hermetically sealed with a lower end plug 33 and an upper end plug 35 respectively. An active fuel length 14 is a zone filled with these fuel pellets. A gas plenum 34 is formed between an upper end of the active fuel length 14 and the upper end plug 35 in the hermetically sealed cladding tube 36. In the active fuel length 14 in each fuel rod 3, the above-mentioned five zones, i.e., the upper blanket zone 5, the upper fissile zone 6, the inner blanket zone 7, the lower fissile zone 8, and the lower blanket zone 9 are sequentially formed from the upper end to the lower end. The control rod 2 having a Y-shaped cross section has three blades extending outward from a tie rod located in the center. Each blade is provided with a plurality of neutron absorbing rods filled with B4C, which is a neutron absorber, and disposed around the tie rod at intervals of 120 degrees. The control rod 2 is provided with a follower portion 16 composed of carbon, which is a material having a smaller moderating power than light water, in the insertion end portion to be inserted into the core 20 first. Below the follower portion 16 in the control rod 2 is a neutron absorber filling-zone 15 formed by the neutron absorber filled in each neutron absorber rod (see FIG. 14). When the BWR 19 is in operation at its rated power, the safety rods, which are control rods 2 being completely withdrawn, are withdrawn from the core 20 such that the upper end of the neutron absorber filling-zone 15 is positioned at the lower end of the active fuel length 14 in the fuel rod 3 (see FIG. 14). When the BWR 19 is in operation, the coolant in the downcomer is pressurized by rotation of the internal pumps (coolant supplying apparatuses) 26 and then supplied into the core 20. The coolant supplied into the core 20 is introduced to each fuel assembly 1, and heated by heat generated by nuclear fission of the nuclear fission material, causing part of the coolant to turn into steam. The coolant in a gas-liquid two-phase flow state is introduced from the core 20 to the steam separator 21, where the steam is separated. Moisture in the separated steam is further removed by the steam dryer 22. The steam from which the moisture has been removed is supplied to a turbine (not shown) through the main steam pipe 23 and rotates the turbine. A power generator (not shown) linked to the turbine rotates and generates electric power. The steam exhausted from the turbine is condensed in a condenser (not shown) and turns into condensed water. This condensed water (feed water) is introduced into the reactor pressure vessel 27 through the feed water pipe 24. The liquid coolant separated by the steam separator 22 is mixed with the above feed water in the downcomer and pressurized by the internal pumps 26 again. The rated flow rate of the BWR 19 is 22 kt/h. An arrangement of the fuel assemblies 1 in the core 20 in the state of being an equilibrium core is described with reference to FIG. 15. Fuel assemblies 1E (four-times burned fuel assemblies 1E) in the operation cycle of which is the fifth cycle and staying in the core for the longest time in the in-core fuel dwelling time, are disposed in the outermost peripheral region of the core having a low neutron importance. In a core outer region internally adjacent to the outermost peripheral region, fuel assemblies 1A (fresh fuel assemblies 1A) staying in the core in a first cycle in the in-core fuel dwelling time and having the highest neutron infinite multiplication, are loaded to flatten the power distribution in the radial direction of the core. In a core inner region, fuel assemblies 1B, 1C, and 1D (once-burned fuel assemblies 1B, twice-burned fuel assemblies 10, and three-times burned fuel assemblies 1D) are disposed, the operation cycles of which are respectively second cycle, third cycle, and fourth cycle in the in-core fuel dwelling time. Such an arrangement is made to flatten the power distribution in the core inner region. Each of the fuel assemblies 1A, 1B, 1C, 1D, and 1E is a fuel assembly 1 shown in FIG. 13, and FIGS. 17 and 18 given later. A plurality of fuel supports (not shown) are provided to a core plate (not shown) disposed to the lower end portion of the core 20. Lower tie-plates (not shown) of four fuel assemblies 1 are supported by one fuel support. Four coolant passages for introducing the coolant to four fuel assemblies are formed in each fuel support, and an orifice (not shown) provided to each fuel support is disposed at the inlet portion of each coolant passage. The core 20 forms two regions in the radial direction, an outermost peripheral region 6 and an inner region 7 located inside the outermost peripheral region 6 (see FIG. 16). Each orifice located in the outermost peripheral region 6, where the power of the fuel assembly 1 is small, has a smaller bore diameter than that of the orifice located in the inner region 7. As shown in FIG. 17, the fuel assembly 1 has five zones, i.e., the upper blanket zone 5, the upper fissile zone 6, the inner blanket zone 7, the lower fissile zone 8, and the lower blanket zone 9, sequentially formed in the portion of the active fuel length from the upper end to the lower end. The height of each zone is as follow: the upper blanket zone 5 (the upper blanket zone 5A) is 70 mm high, the upper fissile zone 6 (the upper fissile zone 6A) is 241 mm high, the inner blanket zone 7 (the inner blanket zone 7A) is 520 mm high, the lower fissile zone 8 (the lower fissile zone 8A) is 225 mm high, and the lower blanket zone 9 (the lower blanket zone 9A) is 280 mm high. When the fuel assembly 1 is a new fuel assembly with a burnup of 0, all the fuel rods 3 in the fuel assembly 1 are filled with depleted uranium oxide pellets in the three blanket zones. The upper fissile zone 6 and the lower fissile zone 8 are filled with mixed oxide fuel having a mixture ratio of 100 parts by average weight of TRU to 172 parts by weight of depleted uranium. A weight ratio of fissile Pu to the total weight of the TRU and the depleted uranium in the mixed oxide fuel, that is, an average enrichment of the fissile Pu is 18 wt %. The TRU is a material extracted, by reprocessing, from the nuclear fuel material contained in the spent fuel assemblies 1. None of the blanket regions are filled with the mixed oxide fuel. Instead of the depleted uranium, the oxide pellets of natural uranium or of the depleted uranium recovered from a spent fuel assembly may be used in each blanket zone. The fuel assembly 1 includes five types of fuel rods 3 shown in FIG. 18. These fuel rods 3 are fuel rods 3A to 3E. The fuel rods 3A to 3E are disposed in the fuel assembly 1 as shown in FIG. 18. In the mixed oxide fuel filled in each of the upper fissile zone 6 and the lower fissile zone 8 of each of the fuel rods 3A to 3E, a fissile Pu enrichment is 10.7 wt % in the fuel rod 3A, 13.5 wt % in the fuel rod 3B, 16.8 wt % in the fuel rod 3C, 18.2 wt % in the fuel rod 3D, and 19.5 wt % in the fuel rod 3E, when the fuel assembly is new fuel assembly having a burnup of 0. The average enrichment of the fissile Pu is 18 wt % for both the upper and lower fissile zones 6 and 8. None of the blanket zones of each fuel rod 3 includes TRU, but the mixed oxide fuel in the upper fissile zone 6 and the lower fissile zone 8 of each fuel rod 3 includes TRU with the composition shown in Table 1 when the burnup is 0. When the fuel assembly 1 is a new fuel assembly, the ratio of Pu-239 in all the TRU is 44 wt %. Table 1 shows a composition of TRU in the nuclear fuel material obtained by reprocessing the nuclear fuel material in a spent fuel assembly, included in a fuel assembly 1, which was originally taken out of the core 20, stayed outside the core for the total of three years, that is, two years in a fuel storage pool and a fuel reprocessing facility and one year in a fuel manufacturing facility, and then loaded again into the core as a new fuel assembly. A plurality of TRU isotopes of the TRU shown in Table 1 is included in the nuclear fuel material in the new fuel assembly 1 obtained by reprocessing. The present embodiment achieves TRU multi-recycling in which, the composition of the TRU in the fuel assembly taken out from the core at the completion of an operation cycle and the composition of the TRU in the fuel assembly newly loaded to the core ready to start the operation cycle are practically uniform. According to the present embodiment in which, the sum of the heights of the upper blanket zone and the lower blanket zone is 350 mm and the height of the lower blanket zone is 4 times the height of the upper blanket zone, a sufficient safety margin can be maintained even with the occurrence of a compound event beyond design standards during the operation of the BWR 19 such as the core flow rate suddenly dropping for some reason and all the control rods being inoperable, by positioning the upper end of the neutron absorber filling-zone 15 of the safety rod, which is a control rod 2 being completely withdrawn at the starting time of the rated operation of the reactor, to the lower end of the active fuel length 14 of the fuel rod 3 (the lower end of the core 20) (see FIG. 14). On the occurrence of such a compound event, the void fraction in the core rapidly rises, the boiling start point of the coolant being slightly sub-cooled and flowing into the core from below the core, shifts to the lower end side of the core, and the power distribution in the axial direction of the core shifts to the lower end side of the core. For this reason, B4C in the neutron absorber filling-zone 15 whose upper end is positioned at the lower end of the core, that is, the lower end of the lower blanket zone 9A, can absorb excess neutrons shifting to the lower end of the core. As a result, in the present embodiment, power can be automatically reduced to the power at which the fuel assemblies 1 in the core 20 can be cooled by the capacity of the coolant suppliable by the emergency high-pressure core flooder 32, and a sufficient safety potential can be maintained even with the occurrence of the compound event beyond design standards. The present embodiment such as this can improve a safety margin without sacrificing the economic efficiency of the light water breeder reactor, which is a light water reactor, even with the occurrence of the above composite event. In the present embodiment, since the height of the lower blanket zone is higher than the height of the upper blanket zone and the height of the upper blanket zone is 70 mm, which is no more than 105 mm, the safety margin of the to core upon the occurrence of the above compound event can be further improved. In the present embodiment, since the height of the lower blanket zone is higher than the height of the upper blanket zone and the height of the upper fissile zone is 16 mm, which is at least 10 mm, higher than that of the lower fissile zone, the safety margin of the core upon the occurrence of the above compound event can is be further improved. In order to suppress a decrease in reactor reactivity when the upper end of the neutron absorber filling-zone 15 of the safety rod, which is a control rod 2 being completely withdrawn during the rated power operation of the BWR 19, is positioned at the lower end of the active fuel length 14 of the fuel rod 3, the height of the upper fissile zone 6 is set to 241 mm and the height of the lower fissile zone 8 to 225 mm. In addition, in order to maintain a breeding ratio to 1.01 while keeping an impact to the void fraction minimum, the height of the upper blanket zone 5 is set to 70 mm and the height of the lower blanket zone 9 to 280 mm which is 1.6 times more that of the upper blanket zone 5. The present embodiment can meet all the restrictive conditions, maintain a breeding ratio of 1.01, and at the same time, automatically reduce power to the power at which the fuel assemblies can be cooled by the capacity of the coolant suppliable to the core by the emergency high-pressure core flooder 32 even with the occurrence of a compound event beyond design standards such as the core flow rate significantly dropping for some reason and all the control rods being inoperable. For this reason, the safety margin of the BWR 19, which is a light water breeder reactor, can be improved (see FIG. 4). In the BWR 19 to which the core 20 is applied and which generates the same electric power of 1350 MW as a current ABWR by using a reactor pressure vessel 27 of approximately the same size as that in the ABWR, a higher discharge burnup can be achieved in a core zone which includes the upper fissile zone 6A, the lower fissile zone 8A, and the inner blanket zone 7A, but excludes the upper blanket zone 5A and the lower blanket zone 9A, than a burnup of 45 GWd/t in the light water breeder reactor stated in JP 3428150B. The discharge burnup of the core zone in the core 20 becomes 53 GWd/t and the discharge burnup of the core 20 including the upper blanket zone 5A and the lower blanket zone 9A becomes 45 GWd/t. According to the present embodiment, MCPR is 1.3 and the void coefficient is −3×10−4 Δk/k/% void, the absolute value of which is one digit higher than the void coefficient −2×10−5 Δk/k/% void of the light water breeder reactor stated in R. TAKEDA et al., Proc. of International Conference on Advanced Nuclear Fuel Cycles and Systems. GLOBAL '07 Boise, USA, September, 2007, P. 1725. Furthermore, according to the present embodiment, a breeding rate of 1.01 can be achieved while the ratios of TRU isotopes are maintained practically constant as described above. In the present embodiment, the same effect can be obtained by disposing pellets 21 including a neutron absorbing material such as boron, gadolinia, Dy, Sm, Eu, etc. below the active fuel length 14 of the fuel rod 3 included in each fuel assembly (see FIG. 19) instead of positioning the upper end of the neutron absorber filling-zone 15 of the safety rod being completely withdrawn, below the lower end of the active fuel length 14 of the fuel rod 3 (the lower end of the lower blanket zone 9A) (see FIG. 14). A core of a light water reactor core according to embodiment 2, which is another embodiment of the present invention, is described below in detail with reference to FIGS. 20 to 22 and Table 2. TABLE 2NuclideComposition (wt %)Np-2370.1Pu-2384.8Pu-2398.5Pu-24039.1Pu-2414.5Pu-24226.0Am-2414.5Am-242M0.2Am-2434.8Cm-2444.5Cm-2451.4Cm-2461.1Cm-2470.2Cm-2480.3 A core 20A of a light water reactor in the present embodiment has a structure in which the fuel assembly 1 in the embodiment 1 is replaced with a fuel assembly 1K shown in FIGS. 20 and 22, and other components are the same as in the embodiment 1. In the present embodiment, only components different from the embodiment 1 are described, and the descriptions of the same components as in the embodiment 1 are omitted. The core 20A is also a parfait-type core. The light water reactor to which the core 20A is applied is a BWR 19 shown in FIG. 11, in which the core 20 is replaced with the core 20A. This BWR 19 to which the core 20A is applied, has the same structure, except for the core 20, as the BWR 19 to which the core of the embodiment 1 is applied. The core 20A is a core to be applied to a TRU disappearance reactor. In the fuel assembly 1K (see FIG. 20) disposed in the core 20A, 397 fuel rods 3K, each having a diameter of 7.2 mm, are disposed in a regular triangle lattice in a channel box 4. A gap between the fuel rods 3K is 2.2 mm, and 11 fuel rods 3K are disposed in a fuel rod row in an outermost peripheral layer. As shown in FIG. 21, fuel assemblies 1A to 1D which have experienced a different number of operation cycles are disposed in the core 20A in the state of being an equilibrium core. The fuel assemblies 1D, the operation cycle of which is the fourth cycle, are disposed in the outermost peripheral region of the core. The fuel assemblies 1A, the operation cycle of which is the first cycle, are disposed in a core outer region, and the fuel assemblies 1B, 1C, and 1D, the operation cycles of which are respectively the second cycle, third cycle, and fourth cycle, are dispersedly disposed in a core inner region. There is an intermediate region between the core inner region and the core outer region, in which intermediate region, a plurality of fuel assemblies 1B is disposed in an annular shape. In such core 20A, the power distribution in the radial direction is more flattened. Each of the fuel assemblies 1A to 1D shown in FIG. 21 is a fuel assembly 1K. The fuel assembly 1K has a structure in which the lower blanket is removed from the fuel assembly 1 (see FIG. 22), thus its active fuel length portion is divided into four zones. An upper blanket zone 5 is 30 mm high, an upper fissile zone 6 is 228 mm high, an inner blanket zone 7 is 560 mm high, and a lower fissile zone 8 is 215 mm high. When the fuel assembly 1K is a new fuel assembly with a burnup of 0, the two blanket zones are filled with depleted uranium oxide pellets and the upper fissile zone 6 and the lower fissile zone 8 are filled with TRU oxide fuel in all the fuel rods 3K in the fuel assembly 1K. The enrichment of the fissile Pu in this TRU oxide fuel is 13.0 wt %. The TRU for the fuel assembly 1K can be obtained by reprocessing nuclear fuel material in a spent fuel assembly. Neither blanket zone is filled with the mixed oxide fuel. Each TRU fuel in the upper fissile zone 6 and the lower fissile zone 8 contains TRU with the composition shown in Table 2. When the fuel assembly 1K has a burnup of 0, the ratio of Pu-239 in all the TRU is 8.5 wt %. In the core 20A, an upper blanket zone 5A formed by the upper blanket zones 5, an upper fissile zone 6A formed by the upper fissile zones 6, an inner blanket zone 7A formed by the inner blanket zones 7, and a lower fissile zone 8A formed by the lower fissile zones 8 are sequentially disposed from the upper end to the lower end. In the core 20A, the lower end of the lower fissile zone 8A matches the lower end of the core 20A, and no lower blanket zone is formed. In the present embodiment, as in FIG. 14 of the embodiment 1, a safety rod, which is a control rod 2 being completely withdrawn during the rated power operation of the BWR 19, is withdrawn from the core 20A such that the upper end of a neutron absorber filling-zone 15 filled with B4C is positioned at the lower end of an active fuel length of the fuel rod 3K. The control rod 2 is provided, above the neutron absorber filling-zone 15, with a follower portion 16 composed of carbon which is a material having smaller moderating power than light water. According to the present embodiment in which the height of the upper blanket zone is 30 mm, which is no more than 100 mm, the lower end of the lower fissile zone matches the lower end of the core 20A, and no lower blanket zone is provided, the upper end of the neutron absorber filling-zone 15 of each of the plurality of safety rods being completely withdrawn is positioned at the lower end of the active fuel length 14 of the fuel rod 3, that is, the lower end of the lower fissile zone 8A (see FIG. 14); thus upon the occurrence of a compound event beyond design standards such as the core flow rate suddenly dropping for some reason and all the control rods being inoperable during the operation of to the BWR 19, which is a TRU disappearance reactor, the void fraction in the core 20A rapidly rises, the boiling start point of the coolant being slightly sub-cooled and flowing from below the core 20A, shifts to the lower end side of the core 20A, and the power distribution in the axial direction of the core 20 shifts to the lower end side of the core. Therefore, B4C in each neutron absorber filling-zone 15 whose upper end is positioned at the lower end of the lower fissile zone 8A can absorb excess neutrons shifting to the lower end side of the core. As a result, power can be automatically reduced to the power at which the fuel assemblies 1 can be cooled by the capacity of the coolant suppliable to the core 20A from an emergency high-pressure core flooder 32. Even with the occurrence of a compound event beyond design standards, a sufficient safety potential can be maintained in the TRU disappearance reactor. The present embodiment such as this can improve the safety margin without sacrificing the economic efficiency of the TRU disappearance reactor, which is a light water reactor, even with the occurrence of the above compound event. The present embodiment can further improve the safety margin of the core upon the occurrence of the above compound event since it has the upper blanket zone 5A and the height of the upper fissile zone 6A is 13 mm, which is more than 10 mm, higher than that of the lower fissile zone 8A. The height of the upper blanket zone 5 is set to 30 mm and the height of the upper fissile zone 6 is set to 13 mm higher than the height of the lower fissile zone 8 so that when the upper end of the neutron absorber filling-zone 15 of the safety rod being completely withdrawn during the operation of the BWR 19 is positioned at the lower end of the active fuel length of the fuel rod 3 (the lower end of the lower fissile zone 8A), a decrease in core reactivity can be prevented to as well as an impact to the void coefficient can be kept to a minimum. According to the present embodiment in which the height of the upper blanket zone is 30 mm, which is no more than 100 mm, the lower end of the lower fissile zone matches the lower end of the core 20A, and no lower blanket zone is provided, all the restrictive conditions can be met and at the same time, even upon the occurrence of a composite event beyond design standards such as the core flow rate substantially dropping for some reason and all the control rods being inoperable, power can be automatically reduced to the power at which the fuel assemblies can be cooled by the capacity of the coolant suppliable to the core 20A from the emergency high-pressure core flooder 32 (see FIG. 10). For this reason, even with the occurrence of such a compound event, the safety margin of the core 20A can be improved. The core 20A can reduce the amount of TRU included in the fuel assembly 1K to less than that of when the burnup of the fuel assembly is 0. In the BWR 19 to which the core 20A is applied, generating the same electric power of 1350 MW as a current ABWR using the reactor pressure vessel of approximately the same size as that in the ABWR, a discharge burnup of 65 GWd/t for the core 20A can be obtained. According to the present embodiment, MCPR is 1.3 and the void coefficient is −4×10−4 Δk/k/% void, the absolute value of which is one digit higher than the void coefficient of −2×10−5 Δk/k/% void in the TRU disappearance reactor stated in R. TAKEDA et al., Proc. of International Conference on Advanced Nuclear Fuel Cycles and Systems. GLOBAL '07 Boise, USA, September, 2007, P. 1725. Furthermore, according to the present embodiment, TRU can be decreased while the ratios of TRU isotopes are maintained. In the present embodiment, as in the embodiment 1, the same effect can be obtained by disposing pellets 21 including a neutron absorbing material such as boron, gadolinia, Dy, Sm, Eu, etc. below the active fuel length 14 of each fuel rod 3 included in each fuel assembly (see FIG. 19) instead of positioning the upper end of the neutron absorber filling-zone 15 of the safety rod being completely withdrawn, at the lower end of the active fuel length 14 of the fuel rod 3 (the lower end of the lower fissile zone 8A) (see FIG. 14). A core of a light water reactor according to embodiment 3, which is another embodiment of the present invention, is described below in detail with reference to FIGS. 23 to 25 and Table 3. TABLE 3NuclideComposition (wt %)Np-2370.2Pu-2385.0Pu-23913.4Pu-24040.8Pu-2414.6Pu-24221.1Am-2414.7Am-242M0.2Am-2434.1Cm-2443.6Cm-2451.1Cm-2460.8Cm-2470.2Cm-2480.2 A core 20B of a light water reactor in the present embodiment has a structure in which the fuel assemblies 1K in the core 20A in the embodiment 2 are replaced with fuel assemblies 1L described in FIGS. 24 and 25, and other components are the same as in the embodiment 2. The light water reactor to which the core 20B is applied is a BWR 19 shown in FIG. 11, in which the core 20 is replaced with the core 20B. This BWR 19 to which the core 20B is applied, has the same components, except for the core 20, as the BWR 19 to which the core of the embodiment 1 is applied. The core 20B is a core applied to a TRU disappearance reactor. The components of the present embodiment which are different from the embodiment 2 are described, and the descriptions of the components that are the same as the embodiment 2 are omitted. In the fuel assembly 1L used in the present embodiment (see FIG. 24), 397 fuel rods 3L, each having a diameter of 7.6 mm, are disposed in a regular triangle lattice in a channel box 4. A gap between the fuel rods 3L is 1.8 mm, and 11 fuel rods 3L are disposed in a fuel rod row in an outermost peripheral layer. As shown in FIG. 23, fuel assemblies 1A to 1D which have experienced a different number of operation cycles are disposed in the core 20B in the state of being an equilibrium core. The fuel assemblies 1D, the operation cycle of which is the fourth cycle, are disposed in the outermost peripheral region of the core. The fuel assemblies 1A, the operation cycle of which is the first cycle, are disposed in a core outer region, and the fuel assemblies 1B, 1C, and 1D, the to operation cycles of which are respectively the second cycle, third cycle, and fourth cycle, are dispersedly disposed in a core inner region. There is an intermediate region between the core inner region and the core outer region, in which intermediate region, a plurality of the fuel assemblies 1B is disposed in an annular shape. In such core 20B, the power distribution in the radial direction is more flattened. Each of the fuel assemblies 1A to 1D shown in FIG. 23 is a fuel assembly 1L. In the fuel assembly 1L, as in the fuel assembly 1K, its active fuel length portion is divided into four zones (see FIG. 25). An upper blanket zone 5 is 50 mm high, an upper fissile zone 6 is 183 mm high, an inner blanket zone 7 is 560 mm high, and a lower fissile zone 8 is 173 mm high. When the fuel assembly 1L is a new fuel assembly with a burnup of 0, the two blanket zones are filled with depleted uranium oxide pellets and the upper fissile zone 6 and the lower fissile zone 8 are filled with TRU oxide fuel in all the fuel rods 3L in the fuel assembly 1L. The enrichment of the fissile Pu in this TRU fuel is 18.0 wt %. Neither blanket zone is filled with the mixed oxide fuel. Each TRU oxide fuel in the upper fissile zone 6 and the lower fissile zone 8 contains TRU with the composition shown in Table 3. This TRU is a material obtained by reprocessing nuclear fuel material in a spent fuel assembly. When the fuel assembly 1L has a burnup of 0, the ratio of Pu-239 in all the TRU is 13.4 wt %. In the core 20B as well, an upper blanket zone 5A formed by the upper blanket zones 5, an upper fissile zone 6A formed by the upper fissile zones 6, an inner blanket zone 7A formed by the inner blanket zones 7, and a lower fissile zone 8A formed by the lower fissile zones 8 are sequentially disposed from the upper end of the core 20B to the lower end of the core 20B. In the core 20B, the lower end of the lower fissile zone 8A matches the lower end of the core 20B, and no lower blanket zone is formed. In the present embodiment, as in FIG. 14 of the embodiment 1, when the BWR 19 is operated at the rated power, the upper end of a neutron absorber filling-zone 15 with each of safety rods (some of control rods 2) being completely withdrawn, is positioned at the lower end of an active fuel length of the fuel rod 3L (the lower end of the lower fissile zone 8A). The Y-shaped control rod 2 is provided, above the neutron absorber filling-zone 15, with a follower portion 16 composed of carbon which is a material having a smaller moderating power than light water. According to the present embodiment in which the height of the upper blanket zone is 50 mm, which is no more than 100 mm, the lower end of the lower fissile zone matches the lower end of the core 20B, and no lower blanket zone is provided, the upper end of the neutron absorber filling-zone 15 of each safety rod being completely withdrawn, is positioned at the lower end of the active fuel length 14 of the fuel rod 3, that is, the lower end of the lower fissile zone 8A (see FIG. 14); thus upon the occurrence of a compound event beyond design standards such as the core flow rate suddenly dropping for some reason and all the control rods being inoperable during the operation of the BWR 19, which is a TRU disappearance reactor, the void fraction in the core 20B rapidly rises, the boiling start point of the coolant being slightly sub-cooled and flowing from below the core 20B, shifts to the lower end side of the core 20B, and the power distribution in the axial direction of the core shifts to the lower side of the core 20B. Therefore, B4C in each neutron absorber filling-zone 15 whose upper end is positioned at the lower end of the lower fissile zone 8A can absorb excess neutrons shifting to the lower end side of the core 20B. As a result, power can be automatically reduced to the power at which the fuel assemblies can be cooled by the capacity of the coolant suppliable to the core 20A from an emergency high-pressure core flooder 32. Even with the occurrence of a compound event beyond design standards, a sufficient safety potential can be maintained in the TRU disappearance reactor. The present embodiment such as this can improve the safety margin without sacrificing the economic efficiency of the TRU disappearance reactor even with the occurrence of the above compound event. The present embodiment can further improve the safety margin of the core upon the occurrence of the above compound event since it has the upper blanket zone and the height of the upper fissile zone is 10 mm higher than that of the lower fissile zone. The height of the upper blanket zone 5 is set to 50 mm and the height of the upper fissile zone 6 is set to 10 mm higher than the height of the lower fissile zone 8 so that when the upper end of the neutron absorber filling-zone 15 of the safety rod being completely withdrawn during the operation of the BWR 19 is positioned at the lower end of the active fuel length of the fuel rod 3 (the lower end of the lower fissile zone 8A), a decrease in core reactivity can be prevented as well as an impact to the void coefficient can be kept to a minimum. According to the present embodiment, all the restrictive conditions can be met and at the same time, even upon the occurrence of a composite event beyond design standards such as the core flow rate substantially dropping for some reason and all control rods being inoperable, power can be automatically reduced to the power at which the fuel assemblies can be cooled by the capacity of the coolant suppliable to the core 20B from the emergency high-pressure core flooder 32. For this reason, the safety margin of the core 20B can be improved upon the occurrence of such a compound event. The core 20B can reduce the amount of TRU included in the fuel assembly 1L to less than that of when the burnup of the fuel assembly is 0. In the BWR 19 to which the core 20B is applied, generating the same electric power of 1350 MW as a current ABWR using a reactor pressure vessel of approximately the same size as that in the ABWR, a discharge burnup of 65 GWd/t can be achieved for the core 20B. According to the present embodiment, the void coefficient is −6×10−4 Δk/k/% void and MCPR is 1.3 and TRU can be decreased while the ratios of TRU isotopes are maintained. A core of a light water reactor core according to embodiment 4, which is another embodiment of the present invention, is described below in detail with reference to FIGS. 26, 27, and Table 4. TABLE 4NuclideComposition (wt %)Np-2370.2Pu-2384.9Pu-2397.0Pu-24035.2Pu-2414.6Pu-24229.4Am-2413.9Am-242M0.2Am-2435.2Cm-2445.7Cm-2451.6Cm-2461.5Cm-2470.3Cm-2480.3 A core 20C of a light water reactor in the present embodiment has a structure in which the fuel assemblies 1K in the core 20A in the embodiment 2 are replaced with fuel assemblies 1M shown in FIGS. 26 and 27, and other components are the same as in the embodiment 2. The light water reactor to which the core 20C is applied is a BWR 19 shown in FIG. 11, in which the core 20 is replaced with the core 20C. This BWR 19 to which the core 20C is applied, has the same components, except for the core 20, as the BWR 19 to which the core of the embodiment 1 is applied. The core 20C is a core applied to a TRU disappearance reactor. The components of the present embodiment which are different from the embodiment 2 are described, and the descriptions of the components that are the same as the embodiment 2 are omitted. In the fuel assembly 1M used in the present embodiment (see FIG. 26), 397 fuel rods 3M, each having a diameter of 7.1 mm, are disposed in a regular triangle lattice in a channel box 4. A gap between the fuel rods 3M is 2.3 mm, and 11 fuel rods 3M are disposed in a fuel rod row in an outermost peripheral layer. The arrangement of the fuel assemblies in an equilibrium core in the present embodiment is the same as that shown in FIG. 21 in the embodiment 2. In the fuel assembly 1M, as in the fuel assembly 1K, its active fuel length portion is divided into four zones (see FIG. 27). An upper blanket zone 5 is 30 mm high, an upper fissile zone 6 is 240 mm high, an inner blanket zone 7 is 560 mm high, and a lower fissile zone 8 is 227 mm high. When the fuel assembly 1M is a new fuel assembly with a burnup of 0, the two blanket zones are filled with depleted uranium oxide pellets and the upper fissile zone 6 and the lower fissile zone 8 are filled with TRU oxide fuel in all the fuel rods 3M in the fuel assembly 1M. The enrichment of the fissile Pu in this TRU fuel is 11.6 wt %. Neither blanket zone is filled with the mixed oxide fuel. Each TRU fuel in the upper fissile zone 6 and the lower fissile zone 8 contains TRU with the composition shown in Table 4. This TRU is a material obtained by reprocessing nuclear fuel material in a spent fuel assembly. When the fuel assembly 1M is a new fuel assembly, the ratio of Pu-239 in all the TRU is 7.0 wt %. In the core 20C as well, an upper blanket zone 5A formed by the upper blanket zones 5, an upper fissile zone 6A formed by the upper fissile zones 6; an inner blanket zone 7A formed by the inner blanket zones 7, and a lower fissile zone 8A formed by the lower fissile zones 8 are sequentially disposed from the upper end of the core 20C to the lower end of the core 20C. In the core 20C, the lower end of the lower fissile zone 8A matches the lower end of the core 20C, and no lower blanket zone is formed. In the present embodiment, as in FIG. 14 of the embodiment 1, when the BWR 19 is operated at the rated power, the upper end of a neutron absorber filling-zone 15 of each of the safety rods (some of Y-shaped control rods 2) being completely withdrawn, is positioned at the lower end of the active fuel length of the fuel rod 3M (the lower end of the lower fissile zone 8A). The control rod 2 is provided, above the neutron absorber filling-zone 15, with a follower portion 16 composed of carbon which is a material having a smaller moderating power than light water. According to the present embodiment in which the height of the upper blanket zone is 30 mm, which is no more than 100 mm, the lower end of the lower fissile zone matches the lower end of the core 20C, and no lower blanket zone is provided, the upper end of the neutron absorber filling-zone 15 of each safety rod being completely withdrawn, is positioned at the lower end of the active fuel length 14 of the fuel rod 3, that is, the lower end of the lower fissile zone 8A (see FIG. 14); thus upon the occurrence of a compound event beyond design standards such as the core flow rate suddenly dropping for some reason and all the control rods being inoperable during the operation of the BWR 19, which is a TRU disappearance reactor, the void fraction in the core 20C rapidly rises, the boiling start point of the coolant being slightly sub-cooled and flowing from below the core 20C, shifts to the lower end side of the core 20C, and the power distribution in the axial direction of the core shifts to the lower side of the core 20C. Therefore, B4C in each neutron absorber filling-zone 15 whose upper end is positioned at the lower end of the lower fissile zone 8A can absorb excess neutrons shifting to the lower side of the core 20C. As a result, power can be automatically reduced to the power at which the fuel assemblies can be cooled by the capacity of the coolant suppliable to the core 20C from an emergency high-pressure core flooder 32. Even with the occurrence of a compound event beyond design standards, a sufficient safety potential can be maintained in the TRU disappearance reactor. The present embodiment such as this can improve the safety margin without sacrificing the economic efficiency of the TRU disappearance reactor even with the occurrence of the above compound event. The present embodiment can further improve the safety margin of the core upon the occurrence of the above compound event since it has the upper blanket zone and the height of the upper fissile zone is 13 mm, which is more than 10 mm, higher than that of the lower fissile zone. The height of the upper blanket zone 5 is set to 30 mm and the height of the upper fissile zone 6 is set to 13 mm higher than the height of the lower fissile zone 8 so that when the upper end of the neutron absorber filling-zone 15 of the safety rod being completely withdrawn during the operation of the BWR 19 is positioned at the lower end of the active fuel length of the fuel rod 3 (the lower end of the lower fissile zone 8A), a decrease in core reactivity can be prevented as well as an impact to the void coefficient can be kept to a minimum. According to the present embodiment, all the restrictive conditions can be met and at the same time, even upon the occurrence of a composite event beyond design standards such as the core flow rate substantially dropping for some reason and all the control rods being inoperable, power can be automatically reduced to the power at which the fuel assemblies can be cooled by the capacity of the coolant suppliable to the core 20C from the emergency high-pressure core flooder 32. For this reason, the safety margin of the core 20C can be improved even with the occurrence of such a compound event. The core 20C can reduce the amount of TRU included in the fuel assembly 1M to less than that of when a burnup of the fuel assembly is 0. In the BWR 19 to which the core 20C is applied, generating the same electric power of 1350 MW as a current ABWR using a reactor pressure vessel of approximately the same size as that in the ABWR, a discharge burnup of 65 GWd/t can be achieved. In the present embodiment, the void coefficient is −3×10−4 Δk/k/% void and MCPR is 1.3 and TRU can be decreased while the ratios of TRU isotopes are maintained. A core of a light water reactor core according to embodiment 5, which is another embodiment of the present invention, is described below in detail with reference to FIG. 28. In the light water reactor core of the present embodiment, each fuel assembly 1 loaded to the core 20 in the embodiment 1 is structured as shown in FIG. 28, and other components are the same as in the embodiment 1. In an active fuel length portion of the fuel assembly, as shown in FIG. 28, five zones, i.e., an upper blanket zone 5, an upper fissile zone 6, an inner blanket zone 7, a lower fissile zone 8, and a lower blanket zone 9 are sequentially formed from the upper end to the lower end. The height of each zone is as follows: the upper blanket zone 5 is 105 mm high; the upper fissile zone 6 is 248 mm high; the inner blanket zone 7 is 520 mm high; the lower fissile zone 8 is 232 mm high; and the lower blanket zone 9 is 280 mm high. In the core 20 loaded with a plurality of fuel assemblies 1 forming each zone shown in FIG. 28, an upper blanket zone 5A formed by the upper blanket zones 5, an upper fissile zone 6A formed by the upper fissile zones 6, an inner blanket zone 7A formed by the inner blanket zones 7, a lower fissile zone 8A formed by the lower fissile zones 8, and a lower blanket zone 9A formed by the lower blanket zones 9 are sequentially disposed from the upper end to the lower end. Each effect generated by the embodiment 1 can be obtained by the core of the present embodiment. In the present embodiment, which considers safety within the design standards only, a higher burnup can be achieved than in the embodiment 1, and in the BWR 19 to which the core of the present embodiment is applied, generating the same electric power of 1350 MW as a current ABWR using a reactor pressure vessel of approximately the same size as that in the ABWR, a discharge burnup of 66 GWd/t for the core zone and a discharge burnup of 55 GWd/t for the core including the upper and lower blanket zones can be achieved. In the present embodiment, the void coefficient is −5×10−5 Δk/k/% void and MCPR is 1.3, and a breeding ratio of 1.01 can be achieved while the ratios of TRU isotopes are maintained practically constant as described above. A core of a light water reactor core according to embodiment 6, which is another embodiment of the present invention, is described below in detail with reference to FIGS. 29 and 30. In the core of the present embodiment, each fuel assembly 1 loaded to the core 20 in the embodiment 1 is structured as shown in FIG. 29, and other components are the same as in the embodiment 1. In the present embodiment, when the light water reactor to which the core of the present embodiment is applied, is operated at the rated power, the upper end of a neutron absorber filling-zone 15 of each of the safety rods (some of Y-shaped control rods 2) being completely withdrawn, is positioned at ⅕ the height of a lower blanket zone 9 from the lower end of the lower blanket zone 9 (see FIG. 30). In an active fuel length portion of this fuel assembly, as shown in FIG. 29, five zones, i.e., an upper blanket zone 5, an upper fissile zone 6, an inner blanket zone 7, a lower fissile zone 8, and the lower blanket zone 9 are sequentially formed from the upper end to the lower end. The height of each zone is as follows: the upper blanket zone 5 is 60 mm high; the upper fissile zone 6 is 235 mm high; the inner blanket zone 7 is 450 mm high; the lower fissile zone 8 is 219 mm high; and the lower blanket zone 9 is 280 mm high. In the core of the present embodiment, as in the core 20, an upper blanket zone 5A, an upper fissile zone 6A, an inner blanket zone 7A, a lower fissile zone 8A, and a lower blanket zone 9A are formed at the same axial positions as the upper blanket zone 5, the upper fissile zone 6, the inner blanket zone 7, the lower fissile zone 8, and the lower blanket zone 9 formed in each fuel assembly shown in FIG. 29. The core of the present embodiment can be obtained each effect generated by the embodiment 1. In the BWR 19 to which the core of the present embodiment is applied, generating the same electric power of 1350 MW as a current ABWR using a reactor pressure vessel 27 of approximately the same size as that in the ABWR, a discharge burnup of 54 GWd/t for the core zone and a discharge burnup of 45 GWd/t for the core 20 including the upper and lower blanket zones can be achieved. In addition, in the present embodiment, the void coefficient is −3×10−4 Δk/k/% void and MCPR is 1.3, and a breeding ratio of 1.01 can be achieved while the ratios of TRU isotopes are maintained practically constant as described above.
	summary
	description	This is a divisional of U.S. patent application Ser. No. 10/609,187, filed Jun. 26, 2003, the disclosure of which is incorporated by reference. A portion of the disclosure of this patent document contains material which is subject to copyright protection. The copyright owner has no objection to the facsimile reproduction by any one of the patent disclosure, as it appears in the Patent and Trademark Office patent files or records, but otherwise reserves all copyright rights whatsoever. The present invention relates to techniques and tools for rating computer product capabilities or requirements and for matching computer software requirements with computer hardware capabilities. Matching system requirements of software with the specifications of the computer system on which it will be used is essential to ensure proper functioning of the software on the system. This is especially true for applications that place significant demands on the computer system, such as computer games and computer-assisted drafting tools. Purchasing software for use on a computer system currently requires an understanding of the system requirements of the software and technical details of the computer system. Unfortunately, the average consumer is often unable to match software requirements to system specifications due to the level of technical knowledge required. A significant portion of software places great demands on the computing resources of even the newest computer systems. Furthermore, users of resource-intensive applications upgrade their PCs more frequently than those who do not. Therefore, the ability of consumers to match software requirements with system specifications is important to PC manufacturers, component manufacturers, software manufacturers, and hardware and software retailers. The entertainment software market provides an example of the effects of software requirements/system specifications matching problems facing PC hardware and software manufacturers today. Games created for console platforms (e.g., Sony Playstation® 2, Nintendo® GameCube, or Microsoft® Xbox) tend to have more consistent sales volumes than games created for general-purpose personal computers in part because, with a better guarantee of matching software requirements with system specifications, customers are more willing to buy console titles. Keeping the PC ecosystem vibrant for both games and other performance demanding applications requires a system that enables consumers to purchase such applications with the same surety provided by the game console platforms. Previous attempts to categorize computer systems were based only on a list of required components for a system. In the 1990s, the Multimedia PC Marketing Council introduced the Multimedia PC (MPC) specifications to provide PC specification guidelines for consumers wishing to run multimedia applications. For example, to achieve the Level 1 MPC specification, a computer system was required to have a processor, RAM, hard disk drive, CD-ROM drive, sound card, and video display with certain characteristics. Previous attempts to categorize software include Microsoft's “Designed for Windows 95” logo program. Software vendors were granted permission to use Microsoft's “Designed for Windows 95” logo if the vendors met certain software design requirements set by Microsoft. The design requirements ensured that software carrying the logo would work effectively and reliably in a Microsoft Windows 95 operating system environment. Whatever the benefits of previous techniques, however, they do not have the advantages of the techniques and tools presented below. In summary, the described techniques and tools facilitate assigning ratings to computer products based on the capabilities of computer hardware and the requirements of computer software. In one aspect, a computer system's ability to run software applications (or a computer hardware component's ability to function in a computer system running software applications) is rated. The rating facilitates matching of software application requirements with computer system capabilities. An inventory of the computer system (or computer hardware component) is performed, yielding a set of features. The inventory can include an inventory of system components and features of the components. Performance is tested, yielding performance results for the set of features. The set of features and the performance results are compared with a required set of features and required performance criteria, and a capability rating (e.g., an integer number) is determined based on the comparison. The capability rating can be based on a lowest rating among components of a system or features of a component. The capability rating can be further based on a category of software applications (e.g., entertainment, computer-assisted drafting software, operating system software, image processing software, etc.). The capability rating can be associated with a vendor identifier for the computer system. The rating process can be performed in an operating system environment of the computer system as a feature of the operating system environment. For example, a computer running a capability tool rates a computer system's ability to run software applications. The capability tool can be a software program downloaded for execution locally or run from a remote computer via a web interface. Capability rating requirements can be obtained locally from the computer system or over a network from a database. Capability rating requirements can be obtained from a device driver signature for a computer hardware component. In another aspect, communicating a capability rating for a computer product (e.g., a hardware product or software product) involves assigning a capability rating to the computer product. The assigning involves associating a standard presentation of the capability rating with the computer product. The capability rating is determined by a computer-executed capability tool. The associating can be performed under a licensing agreement. While systems and hardware components are rated based on an objective set of tests, software ratings are based on largely subjective evaluations performed by the software vendor, determining what capability level is required from a computer system to provide a positive user experience with the software product. In some aspects, the software vendor may choose to provide several capability requirement levels, comprising a minimum rating, a recommended rating, and a “works-best” rating. Thus the consumer can determine not only if the software will work adequately on their computer system, but also if it will work exceptionally well. In another aspect, proposed capability rating level requirements comprising computer system performance criteria are created, and finalized capability rating level requirements are then created (e.g., by a ratings board comprising voting members) based on the proposed capability rating level requirements to establish one or more capability rating levels. A capability rating level is determined (e.g., by a testing organization) for one or more computer hardware devices based on the finalized capability rating level requirements and analysis of the one or more computer hardware devices. The analysis is performed on a computer running a capability tool. In some embodiments, analysis of the proposed capability rating level requirements is received (e.g., from a review group) prior to creating the finalized capability rating level requirements, and the finalized requirements are further based on the analysis of the proposed requirements. In another aspect, proposed rating levels for software applications are created based on computer system performance requirements for running the software applications on a computer system. Analysis of proposed rating levels is received and finalized rating levels are created based on proposed rating levels and the analysis. A finalized rating level is assigned to software applications based on computer-executed performance testing of the one or more software applications. The finalized rating level matches a capability rating level assigned to the computer system. In some embodiments proposed rating levels are based on performance testing of the one or more software applications. In another aspect, a software system for rating a computer system's ability to run software applications comprises an inventory module, a performance testing module, and an inventory and performance evaluator module. The inventory and performance evaluator module compares a set of computer system features obtained by the inventory module, performance results obtained by the performance testing module for the set of computer system features, and rating level requirements comprising a required set of features and required performance criteria. The inventory and performance evaluator module determines a rating level for the computer system based on the comparison. The software system can display rating information (e.g., via a graphical user interface) including, for example, rating level requirements, rating levels of individual components in the computer system, recommended computer system upgrade information, and/or a comparison of the performance results and the rating level requirements. Techniques and tools for implementing a capability rating system are described. The techniques and tools expedite the process of matching software requirements with computer system specifications. The techniques and tools provide a capability rating system that allows prospective purchasers of all knowledge levels to confidently purchase software for a PC, and to understand the value they would receive from purchasing new PC components or a new PC capable of running a broader set of applications. Consumers, hardware vendors, and software vendors will benefit from techniques and tools that allow a person with a limited understanding of computer system capabilities and software requirements to make informed software and hardware purchases. The described techniques and tools help consumers by, for example, allowing them to purchase demanding multimedia software applications (e.g., games, CAD programs, photo-editing programs, feature-rich operating systems, etc.) without having an unsatisfactory experience due to substandard performance on their PC. Consumers can use capability ratings on software or hardware to determine if their computer system is becoming substandard for the applications they wish to use. Consumers then can use capability rating information to make an accurate determination of what capability level to seek in a new system purchase or system upgrade. For example, computer games often require a high level of computer capabilities and performance. Computer game players can use capability rating information to determine the best computer system or component for a game they wish to play, or to find a computer game well-suited to the capabilities of their current system. Different capability rating systems can be used for different kinds of applications. For example, a low rating on a sound card might reduce a capability rating for games, but might not reduce a capability rating for running CAD applications, where sound is less important to application performance. The described techniques and tools help OEMs by, for example, instilling in consumers increased confidence in the quality of purchased PCs, and increasing motivation to purchase new PCs through consumer desires for key entertainment and other applications. Consumers may purchase higher-end PCs (which typically have higher profit margins than lower-end systems) because of a desire for and increased ability to recognize the value of more highly capable systems. Further, since the capability rating system clarifies software compatibility, the burden placed on customer support resources by frustrated customers is reduced. Similarly, the described techniques and tools help IHVs by, for example, encouraging consumers to purchase more high-end components as part of new higher-end PC sales or through PC component upgrades. Furthermore, the described techniques and tools help software manufacturers through, for example, increased consumer purchasing confidence, increased sales of software that typically comes bundled with new PC purchases (e.g., operating systems, personal productivity software, and utilities), and, similar to OEMs, a reduction of the burden on customer support resources. Accordingly, the described techniques and tools allow hardware and software vendors to communicate, in a simple and easily understood manner, capability ratings for computer hardware and software. The described techniques and tools, individually or in combination, include (but are not limited to) the following features: 1. Capability tools (e.g., for OEMs or end users) 2. Metrics board for proposing and deciding rating level requirements 3. Metrics review group for reviewing proposed rating level requirements 4. Standardized rating presentation for hardware (systems and components) 5. Standardized rating presentation for software Capability tools facilitate assigning capability ratings to computer products. For example, in some embodiments, a computer user can use a capability tool to determine a capability rating level for a computer system. In other embodiments, a computer hardware vendor, such as a PC manufacturer or PC component manufacturer, can use a capability tool to determine capability rating levels for its products. There are several differences between capability tools and benchmarking applications, these differences are made apparent from the descriptions and examples; herein. In some embodiments, criteria for assigning capability ratings are determined by a board comprising voting members. The board proposes capability rating level requirements and receives feedback on the proposed requirements from a review group. After receiving feedback, the board decides on capability rating level requirements, which can then be used to assign capability ratings of various levels to computer hardware or software. In some embodiments, a standardized rating presentation is used to associate computer hardware or software with capability ratings. For example, a rating presentation indicating the capability rating (e.g., an integer number) for a product is used by hardware manufacturers under a licensing agreement. Standard presentations can include several rating levels. For example, the software presentation can include a minimum rating level, a recommended rating level, and/or a “best experience” rating level. Some embodiments of the described techniques and tools operate in an operating system environment. For example, an operating system environment running on a PC includes a capability tool which facilitates determining a capability rating for the PC. I. Computing Environment FIG. 1 illustrates a generalized example of a suitable computing environment 100 in which some of the described techniques and tools may be implemented. FIG. 1 also illustrates a generalized example of a personal computer system configuration and hardware components that can be evaluated using a capability rating system. The computing environment 100 is not intended to suggest any limitation as to scope of use or functionality of described embodiments, which may be implemented in diverse general-purpose or special-purpose computing environments. With reference to FIG. 1, the computing environment 100 includes at least one processing unit 110 and memory 120. In FIG. 1, this most basic configuration 130 is included within a dashed line. The processing unit 110 executes computer-executable instructions and may be a real or a virtual processor. In a multi-processing system, multiple processing units execute computer-executable instructions to increase processing power. The memory 120 may be volatile memory (e.g., registers, cache, RAM), non-volatile memory (e.g., ROM, EEPROM, flash memory, etc.), or some combination of the two. In some embodiments, the memory 120 stores software 180 implementing capability rating system tools. A computing environment may have additional features. For example, the computing environment 100 includes storage 140, one or more input devices 150, one or more output devices 160, and one or more communication connections 170. An interconnection mechanism (not shown) such as a bus, controller, or network interconnects the components of the computing environment 100. Typically, operating system software (not shown) provides an operating environment for other software executing in the computing environment 100, and coordinates activities of the components of the computing environment 100. In some embodiments, operating system software includes capability rating system tools. The storage 140 may be removable or non-removable, and includes magnetic disks, magnetic tapes or cassettes, CD-ROMs, CD-RWs, DVDs, or any other medium which can be used to store information and which can be accessed within the computing environment 100. In some embodiments, the storage 140 stores instructions for the software 180 implementing capability rating system tools. The input device(s) 150 may be a touch input device such as a keyboard, mouse, pen, trackball, or game controller, a voice input device, a scanning device, a digital camera, or another device that provides input to the computing environment 100. The output device(s) 160 may be a display, printer, speaker, or another device that provides output from the computing environment 100. The communication connection(s) 170 enable communication over a communication medium to another computing entity. The communication medium conveys information such as computer-executable instructions, audio or video information, or other data in a modulated data signal. A modulated data signal is a signal that has one or more of its characteristics set or changed in such a manner as to encode information in the signal. By way of example, and not limitation, communication media include wired or wireless techniques implemented with an electrical, optical, RF, infrared, acoustic, or other carrier. The invention can be described in the general context of computer-readable media. Computer-readable media are any available media that can be accessed within a computing environment. By way of example, and not limitation, within the computing environment 100, computer-readable media include memory 120, storage 140, communication media, and combinations of any of the above. The invention can be described in the general context of computer-executable instructions, such as those included in program modules, being executed in a computing environment on a target real or virtual processor. Generally, program modules include routines, programs, libraries, objects, classes, components, data structures, etc. that perform particular tasks or implement particular abstract data types. The functionality of the program modules may be combined or split between program modules as desired in various embodiments. Computer-executable instructions for program modules may be executed within a local or distributed computing environment. II. Capability Rating System As explained above, the described techniques and tools allow computer hardware and software manufacturers to communicate the value of their products to consumers and stimulate more frequent computer hardware and software purchases. The described techniques and tools, individually or in combination, include (but are not limited to) the following features: 1. Capability tools (e.g., for OEMs or end users) 2. Metrics board for proposing and deciding rating level requirements 3. Metrics review group for reviewing proposed rating level requirements 4. Standardized rating presentation for hardware (systems and components) 5. Standardized rating presentation for software A. Capability Rating Levels A capability rating level represents the capability of the rated item to perform functions for which the rated item is designed. For example, the capability rating of a computer system represents the capability of the computer system to run software applications. The capability rating of a computer system is typically a function of the capability ratings of its components. However, it may not be necessary to increase ratings for all components of a system in order to increase the rating for the system as a whole. In some embodiments, a capability rating level is determined (e.g., by a capability tool) using a “weakest link” analysis—the capability rating level for a system is determined by identifying the component within the system having the lowest capability. For example, the capability rating level for a computer system is determined to be “3” by finding that the component within the primary device set of system with the lowest capability has a capability rating level of “3.” The “weakest link” analysis also can be applied to determining capability ratings for PC components, by determining the feature of the component with the lowest capability. FIG. 2 shows a technique 200 for determining a capability rating for a computer system or component. At 210, an inventory is taken of the computer system or component. For example, when determining a capability rating for a computer system, a capability tool takes an inventory of components in a computer system that are relevant to the capability of the system to run software applications and, in addition, takes an inventory of features of components within the computer system. At 220, the features and the performance of the computer system or component are evaluated to determine a rating. A capability rating is determined based on a required set of features with a required level of performance. For example, a capability rating for a graphics component may require the component to have a graphics chip capable of transforming 1,000,000 generic polygons per second and require a certain level of overall graphics performance for the component. A graphics component may not qualify for the capability rating if it has a graphics chip capable of transforming the required number of generic polygons per second but is not able to transform that many polygons with a specific set of features turned on. Finally, at 230, a capability rating is assigned to the system or individual component based on the inventory and the evaluating of the features and performance. B. Capability Tools Capability tools are tools for assigning capability rating levels to computer products. For example, in some embodiments, a computer user can run a capability tool on a PC to determine a capability rating level for the system. In other embodiments, an OEM, IHV or testing lab can use a capability tool to determine capability rating levels. Capability tools for OEMs, IHVs and testing labs include a more extensive set of testing tools, while capability tools for most consumers include fewer testing tools and run more quickly than capability tools for OEMs, IHVs and testing labs. However, capability tools for consumers wishing to perform more robust testing of computer systems can include a full set of testing tools. Capability tools analyze computer hardware and use results of the analysis in combination with capability rating level data to determine a capability rating level (e.g., an integer number) for the hardware. Capability rating level data can be accessed from a database (e.g., an online database). Or, capability rating level data can be provided with a computer system or component (e.g., embedded in a device driver signature). A capability tool can be accessed or downloaded over a network, accessed locally (e.g., as a feature of an operating system on a computer system to be rated), or provided to a user of the tool by other means (e.g., by distribution on a CD-ROM or other computer-readable medium). FIG. 3 shows an exemplary arrangement 300 in which one or more computer users can access a capability tool 310 at a remote computer 315 via a network 320. In addition to the capability tool 310, the remote computer 315 also may have capability rating level data 325 (e.g., in a database), which the capability tool can use to determine capability ratings for a computer product. Alternatively, capability rating level data is at another location (e.g., within the capability tool itself, or in a device driver signature of a hardware component). A computer user can use capability tool 310 by connecting to remote computer 315 and running the tool from the remote computer (e.g., via a web interface). Or, a computer user can download capability tool 310 and run the tool locally. In some embodiments, a capability tool is available at a website of a computer hardware or software manufacturer for use by users or retailers of the manufacturers' products. In particular, an online capability tool can be useful for retailers providing PC upgrade services. In other embodiments, a capability tool is available on another website for use by a manufacturer. A computer user can access the capability tool 310 from a variety of computing devices connected to network 320 via a wired connection (e.g., via an analog telephone line connection, DSL connection, cable modem connection, or some other wired connection) or a wireless connection, and/or via some other network. For example, a computer user can access capability tool 310 via a desktop computer connected to a local network, such as desktop computers 330-332 connected to local network 335. Or, a computer user can access capability tool 310 via a desktop computer that is not connected to a local network, such as desktop computer 340. A computer user can also access capability tool 310 from computing devices such as portable laptop computer 350, hand-held computer 360, or some other computing device, such as a computing device with cellular telephone functionality or other functionality. Capability tools can provide information to a user of the tool regarding the system or component being rated. For example, in some embodiments, in addition to providing a capability rating, a capability tool for determining a capability rating for PC can display (e.g., via a graphical user interface) capability level requirements for one or more capability levels, capability ratings for components of the PC, and/or a comparison of the PCs features and performance with those needed to reach one or more capability levels (e.g., showing how PC specifications exceed or fall short of requirements for a given capability level). Capability tools for consumers can differ from capability tools for vendors in terms of the amount and nature of information they display. Capability tools for vendors, for example, may contain additional technical details regarding component and feature inventories and performance results. For example, a capability tool for use by a component manufacturer may include technical performance data and details on how to improve a component's capability rating. In some embodiments, a capability tool provides upgrade information to a user of the tool. For example, a capability tool that has determined the capability rating of a computer system to be “3” can provide information on how a user of the computer system could upgrade the system to reach a capability rating of “4” (e.g., by upgrading a “weakest link” component with a rating of “3” to a rating of “4”). Capability tools can use tools such as Microsoft's DirectX Diagnostics (DXDiag), DirectX Caps Viewer, or Display Conformance Test (DCT) or benchmarking tools to perform or assist in performing certain functions, such as taking inventory of components or component features, or testing performance of systems or components. C. Capability Rating Level Requirements: Decision-making In some embodiments, criteria for assigning capability ratings are determined by a board comprising voting members. The board proposes capability rating level requirements and receives feedback on the proposed requirements from a review group. After receiving feedback, the board decides on capability rating level requirements, which can then be used to assign capability ratings of various levels to computer hardware or software. FIG. 4 is a flow chart illustrating a technique 400 for deciding capability rating level requirements. At 410, level requirements are proposed by a decision-making entity. At 420, the decision-making entity receives feedback (e.g., comments on the proposed requirements) and analyzes the feedback. At 430, the decision-making entity determines whether to finalize the level requirement (e.g., whether to end the feedback process and issue actual level requirements to be put into effect for a rating system). If the decision-making entity chooses not to finalize the requirements, the entity can propose additional or revised level requirements and receive feedback on those proposed requirements. If the decision-making entity chooses to finalize the requirements, the entity provides (at 440) the level requirements to test organizations, which can assign appropriate capability ratings to the hardware or software based on the level requirements. Alternatively, a decision-making entity can employ different techniques for deciding on rating level requirements. For example, the decision-making entity may not receive feedback on requirements before deciding on actual requirements to be provided to testing organizations. Or, the decision-making entity can employ additional procedures, omit procedures, or perform procedures in a different order. FIG. 5 shows an exemplary arrangement 500 for deciding capability rating level requirements. A board 510 constructs proposed requirements 520 for one or more capability rating levels. The board 510 sends the proposed requirements 520 to a review group 530, which considers the proposed requirements 520 and provides feedback 540 to the board 510 regarding the proposed requirements 520. The board considers the feedback 540 and determines whether to decide on final requirements 550 for the one or more capability rating levels. The board can also revise the proposed requirements 520 and present new proposals to the review group 530 for further consideration. When the board 520 decides on final requirements for the one or more capability rating levels, the board presents the requirements to testing organizations 560. Alternatively, requirements can be decided upon by different entities, by a different arrangement of entities, or using different procedures. For example, the board 510 may not, in some cases, receive or consider feedback from the review group 530, or the board may provide requirements to manufacturers rather than to testing organizations. Testing organizations can use capability rating level requirements to test computer hardware or software and assign appropriate capability ratings to the hardware or software. For example, FIG. 6 shows an exemplary arrangement for rating of computer products by testing organizations. Product vendors 610 provide products 620 to testing organizations 630. The testing organizations test the products and use the capability rating level requirements to provide capability ratings 640 for the products 620. The product vendors 610 can then sell rated products 650 to purchasers 660. In some embodiments, a board comprises a set of OEM and ISV representatives, who can propose and/or decide on rating level requirements. OEM representatives can include representatives of manufacturers of major PC brands (e.g., Dell, Gateway, Sony, Hewlett-Packard, etc.). ISV representatives can include representatives of software manufacturers involved in producing software that makes significant demands on PC resources (e.g., entertainment software manufacturers such as Electronic Arts, Atari, Microsoft, Activision, etc.). The board also can receive information and comment from IHVs regarding the rating level requirements and relevant capabilities of planned IHV products. In some cases, IHVs can participate in board decisions or act as board members; in other cases, confidentiality or conflict of interest issues may prevent participation on the board by IHV representatives or other representatives. The board can decide on the requirements of new rating levels at regular intervals (e.g., one new rating level per year). For example, at annual meetings, the board could convene for a week of half-day meetings, with remaining time spent by members conferring with interested parties regarding issues relevant to setting rating level requirements. Following the meetings, the board can approve a draft specification for a rating level, and submit the draft specification to a review group for comment. Received comments can be distributed to board members for consideration and discussion, and can be incorporated into future proposed requirements or a final set of requirements. Final requirements can be issued in a variety of ways (e.g., by publication on a publicly available website, or by providing them directly to manufacturers or testing organizations). Board decisions can be made by a majority vote of board members. However, the board can make decisions in other ways. For example, a board also can comprise a leadership group of voting members to work with OEM and ISV representatives. The leadership group can facilitate decision-making by, for example, reserving the right to make a final decision on contentious issues on behalf of the board. The board also can be described as a metrics board, capability ratings board, etc. As explained above, the review group can examine proposed requirements and provide feedback to the board for consideration. The review process can be iterative, proceeding until closed by the board or until some event occurs (e.g., a time limit is reached). In some embodiments, the review group comprises a broader set of OEM and ISV representatives, as well as IHV representatives. OEM and ISV representatives can include representatives of the companies described above with reference to the board, or can come from other companies. IHV representatives can include representatives from companies such as Intel, AMD, nVidia, ATI, Matrox, S3, Via, Trident, Silicon Integrated Systems (SiS), 3DLabs, Creative Labs, etc. Diverse representation in the review group encourages broad industry participation and ensures the effectiveness and acceptability of the capability rating system and its requirements. Once the requirements of a rating level are determined, products can be tested by one or more testing organizations, such as Microsoft's Windows Hardware Quality Labs (WHQL), to determine capability levels for the products. Testing organizations can use a capability tool to determine capability ratings. Typically, testing organizations will use a capability tool with a full set of testing tools, rather than a simpler, consumer-oriented capability tool. The capability level for the product and its technical information and descriptions can be added to a database used by other capability tools. For example, technical information for a graphics accelerator card and its associated capability level can be added to a database accessed by capability tools used by end-users to determine capability ratings for their PCs. In some embodiments the capability level for the product is associated with a vendor ID for the product. D. Standardized Capability Rating Presentation To associate hardware with capability ratings that are assigned to them, a standardized rating presentation can be placed on computer hardware. In some embodiments, a rating presentation indicating the capability rating (e.g., an integer number) for a product is used by hardware manufacturers under a licensing agreement (e.g., an agreement with an entity that has set capability rating level requirements and implemented the presentation of rating levels in a standardized format). The licensing agreement can require, for example, that the component or system must have been tested by an approved testing organization to verify the rating level. The standardized rating presentation allows consumers to use capability ratings when determining which computer system or component to purchase, based on their desire to run applications that require or recommend certain capability levels. Similarly, ISVs can use a standardized rating presentation for software for their products. In some embodiments, the standardized rating presentation for software is virtually identical to the hardware presentation to help consumers recognize and match capability ratings for hardware and software. The standardized rating presentation for software may also be used under a license agreement. Such an agreement may further require, for example, that software products using the rating presentation conform to other standards. For example, the license agreement may require the software to fulfill requirements for running the software on a particular operating system (e.g., Microsoft's Windows XP application logo requirements). The software presentation can include several rating levels. For example, the software presentation can include a minimum rating level, a recommended rating level, and/or a “best experience” rating level. Having several rating levels in the presentation allows consumers with different expectations of performance to select software that works best for them. Furthermore, having rating levels such as a “best experience” rating level can encourage consumers to purchase system upgrades. III. Implementation as Operating System Features Some of the tools and techniques described above can be implemented as operating system features. For example, in some embodiments, an operating system includes a capability tool for determining a capability rating for a PC running the operating system, or for a component of the PC. The capability tool also can update capability ratings for the PC or PC components when the PC is upgraded, or when new capability rating level requirements are determined. Having described and illustrated the principles of our invention with reference to described embodiments, it will be recognized that the described embodiments can be modified in arrangement and detail without departing from such principles. It should be understood that the programs, processes, or methods described herein are not related or limited to any particular type of computing environment, unless indicated otherwise. Various types of general purpose or specialized computing environments may be used with or perform operations in accordance with the teachings described herein. Elements of the described embodiments shown in software may be implemented in hardware and vice versa. In view of the many possible embodiments to which the principles of our invention may be applied, we claim as our invention all such embodiments as may come within the scope and spirit of the following claims and equivalents thereto.
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