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claims | 1. A filter for separating particles in a coolant fluid in a nuclear reactor, the filter comprising:channels for circulation of the coolant fluid through the filter, at least one of the channels extending along a channel centerline, the at least one channel having an inlet upstream section, an outlet downstream section and an intermediate section, the intermediate section extending between the inlet upstream section and the outlet downstream section and being enlarged relative to the inlet upstream section and the outlet downstream section, the inlet upstream section, the intermediate section and the outlet downstream section being arranged along the channel centerline; andat least one separating member defining inside the intermediate section of the at least one channel an annular passage, the annular passage being arranged substantially concentric to the channel centerline in the intermediate section. 2. The filter according to claim 1 further comprising centering elements for maintaining the at least one separating member spaced from an inner surface of the corresponding intermediate section. 3. The filter according to claim 1 wherein the at least one separating member is spherical. 4. The filter according to claim 1 wherein the at least one separating member has an elliptical cross section. 5. The filter according to claim 1 wherein the at least one separating member has a double cone shape. 6. The filter according to claim 1 wherein a first and a second separating member are disposed inside a same intermediate section to define an inner and an outer coaxial annular passage, the inner and outer coaxial annular passage having an axis substantially coaxial to the channel centerline in the same intermediate section, the first separating member being tubular and defining the outer coaxial annular passage with the inner surface of the intermediate section, the second separating member being disposed inside the first separating member and defining therein an inner annular passage. 7. The filter according to claim 1 wherein the at least one separating member has at least one fluid flow hole passing through the separating member. 8. The filter according to claim 1 wherein the channels include at least a first set of channels and a second set of channels, the intermediate sections of the first set of channels being offset relative to the intermediate sections of the second set of channels along a filter main flow direction of the coolant fluid through the filter. 9. The filter according to claim 8 wherein the channels are arranged in a pattern such that each channel of the first set of channels is surrounded by channels of the second set of channels. 10. The filter according to claim 1 wherein the channels are defined by a one-piece filtering plate having a plurality of ducts extending therethrough and a plurality of tubular inserts, each duct having a narrow duct section and an enlarged duct section, each enlarged duct section having a respective one of the tubular inserts inserted therein, each of the channels being defined by one of the ducts and the respective tubular insert inserted inside the enlarged duct section of the duct. 11. The filter according to claim 10 wherein the filtering plate comprises at least two stacked parts, each of the channels extending through the stacked parts and each separating member being disposed between two of the stacked parts. 12. The filter according to claim 11 wherein the filtering plate comprises a lower part, an upper part and an intermediate part interposed between the upper part and the lower part, each separating member being disposed between the intermediate part and one of the upper and lower parts. 13. A nuclear fuel assembly lower nozzle defining a filter as recited in claim 1. 14. The nuclear fuel assembly comprising a bundle of fuel rods and an armature for supporting the fuel rods, the armature comprising a lower nozzle and an upper nozzle, the fuel rods extending between the nozzles, wherein the lower nozzle defines a filter as recited in claim 13. 15. A filter for separating particles in a coolant fluid in a nuclear reactor, the filter comprising:channels for circulation of the coolant fluid through the filter, at least one of the channels extending along a channel centerline, the at least one channel having an upstream section, a downstream section and an intermediate section, the intermediate section extending between the upstream section and the downstream section and being enlarged relative to the upstream section and the downstream section; andat least one separating member defining inside the intermediate section of the at least one channel an annular passage, the annular passage having an axis substantially coaxial to the channel centerline in the intermediate section,wherein a first and a second separating member are disposed inside a same intermediate section to define an inner and an outer coaxial annular passage, the inner and outer coaxial annular passage having an axis substantially coaxial to the channel centerline in the same intermediate section, the first separating member being tubular and defining the outer coaxial annular passage with the inner surface of the intermediate section, the second separating member being disposed inside the first separating member and defining therein an inner annular passage. 16. A nuclear fuel assembly lower nozzle defining a filter as recited in claim 15. 17. The nuclear fuel assembly comprising a bundle of fuel rods and an armature for supporting the fuel rods, the armature comprising a lower nozzle and an upper nozzle, the fuel rods extending between the nozzles, wherein the lower nozzle defines a filter as recited in claim 16. 18. A filter for separating particles in a coolant fluid in a nuclear reactor, the filter comprising:channels for circulation of the coolant fluid through the filter, at least one of the channels extending along a channel centerline, the at least one channel having an inlet upstream section, an outlet downstream section and an intermediate section, the intermediate section extending between the inlet upstream section and the outlet downstream section and being enlarged relative to the inlet upstream section and the outlet downstream section, the inlet upstream section, the intermediate section and the outlet downstream section being arranged along the channel centerline; andat least one separating member defining inside the intermediate section of the at least one channel an annular passage, the annular passage having an axis substantially coaxial to the channel centerline in the intermediate section,wherein the channels are defined by a one-piece filtering plate having a plurality of ducts extending therethrough and a plurality of tubular inserts, each duct having a narrow duct section and an enlarged duct section, each enlarged duct section having a respective one of the tubular inserts inserted therein, each of the channels being defined by one of the ducts and the respective tubular insert inserted inside the enlarged duct section of the duct. 19. The filter according to claim 18 wherein the filtering plate comprises at least two stacked parts, each of the channels extending through the stacked parts and each separating member being disposed between two of the stacked parts. 20. The filter according to claim 19 wherein the filtering plate comprises a lower part, an upper part and an intermediate part interposed between the upper part and the lower part, each separating member being disposed between the intermediate part and one of the upper and lower parts. |
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061817732 | description | DETAILED DESCRIPTION OF INVENTION The invention will next be illustrated with reference to the figures wherein similar numbers indicate the same elements in all figures. Such figures are intended to be illustrative rather than limiting and are included herewith to facilitate the explanation of the apparatus of the present invention. FIG. 1 shows a schematic arrangement in which a source of X-ray radiation 10 provides a beam 18 of X-rays. A target 12 (i.e. a patient in the case of medical diagnostic imaging) is placed in the X-ray beam path. The radiation emerging through patient 12 is intensity modulated because of the different degrees of X-ray absorption in various parts of the patient's body. Cassette enclosure 14, containing radiation sensor 16, intercepts the modulated X-ray radiation beam 18'. Radiation detector 16 absorbs X-rays that penetrate the cassette enclosure 14, and produces a digital image in accordance with the above-referenced patent. A radiation anti-scatter device 20, known in the art as a bucky, comprising an anti-scatter grid attached to a holder, is typically placed between target 12 and cassette 14 to focus the modulated X-ray beam to prevent scattered X-rays from impinging the sensor at undesirable angles. Standard bucky grid architecture comprises a set of parallel vanes. The bucky is typically placed so that it moves in a vertical or horizontal plane orthogonal to the length of the vanes. According to this invention the bucky is moved over the detector in a single stroke during a time period that exceeds the radiation exposure duration. This is obtained by imparting to the moving bucky a decelerating velocity profile preferably one that asymptotically approaches zero. The velocity profile, by necessity, includes an accelerating first period. The accelerating first period must be such as to accelerate the bucky to its maximum velocity quickly enough so as not to unreasonably delay the onset of the actual patient exposure, and not to use up an excessive fraction of the available grid displacement. Typical acceleration times are of the order of a few milliseconds, preferably between 0.001 and 0.5 seconds. The exact time is determined by practical limitations related to the physical environment of a specific installation and equipment available. In general, it is desirable that the grid move between 0.1 and 1.5 cm during the accelerating period, and that the decelerating portion of the grid movement lasts for about 2 seconds and translates the grid another 1 to 5 cm. The acceleration velocity profile may be linear or non-linear, as desired. A linear profile has the advantage of requiring only a constant force to accelerate the grid. In FIGS. 2A-C, there are shown graphs of time versus velocity graph 30, and time versus displacement graph 32, of an exemplary moving bucky. Each graph depicts the same motion, wherein the time period shown in 2B is 10.times. that shown in 2A, and 2C is 10.times. the period in 2B. As illustrated the grid is first accelerated to a first, high velocity, preferably prior to initiating the radiation exposure, and then decelerated again preferably during the exposure. For the first time period, velocity profile 30 conforms to the general equation: EQU V=K.sub.1 t for t equal to or less than 0.005 sec. (1) where: V=velocity in cm/second PA1 K=2236 and PA1 t=time in seconds. PA1 V=velocity in cm/second PA1 K.sub.2 =25, and PA1 m=0.5 PA1 t=time in seconds. For a second time period, for t greater than 0.005 sec. and less than 2 seconds the velocity profile 30 conforms to the general equation: EQU V=K.sub.2 (1000 t).sup.-m (2) where: Referring now to FIG. 3, there is shown an exemplary radiation anti-scatter device 40 of the present invention, showing a grid 42 and grid driver mechanism 44 for imparting motion onto the grid. As shown in FIG. 3, grid driver 44 comprises a motor 46, which may be a variable speed DC motor typical of motors well-known in the art, and a variable-pitch screw 48 that is threaded through a "nut" 50 adapted to mesh with the variable pitch of the screw. Thus, as motor 46 turns screw 48 in the direction of arrow A, nut 50, connected by bracket 51 to grid 42, travels in the direction of arrow B and moves the grid along track 45. Although described as having both a variable speed motor 46 and variable pitch screw 48 with respect to FIG. 3, an alternate grid movement system may comprise a fixed speed motor with a variable pitch screw or any mechanical variable drive coupling known in the art, such as for example, lever/cam or wheel/crank systems. Furthermore, the grid movement system may comprise a variable speed motor with a fixed mechanical coupling. A variable drive coupling and variable speed motor are preferred, however, to promote a operator-changeable accelerating or decelerating velocity profile. Usually, the radiation blocking elements 52 in the grid are parallel to each other and the grid is oriented so that the blocking elements are also parallel to the alignment of sensors 56 of the detector 54, in one direction (i.e. row or column). The motion of the grid is, usually, perpendicular to the grid radiation blocking elements (also known as vanes). Because the grid is moving relative to the detector, any Moire patterns created are transient in nature lasting only a few milliseconds, not long enough to be captured by the detector. An alternate arrangement is shown in FIG. 4. Grid 58 again comprises a plurality of vanes 60 and the motion of the bucky is along arrow B, perpendicular to the orientation of the vanes. The underlying direct radiography panel 62 comprises a plurality of sensors 66 aligned along a first direction (here in rows 64 of sensors 66). The angle a between vanes 60 and rows 64 of sensors 66 is approximately 45 degrees, as shown in FIG. 3. Thus, the angle (90-.alpha.) between the motion along arrow B and the orientation of the rows of pixels is also approximately 45 degrees. Although an approximate 45-degree orientation is shown herein, angle a may be any non-parallel or non-orthogonal angle that minimizes Moire pattern artifacts in a radiograph produced by the imaging system of which the bucky is a component. Referring now to FIG. 5, the invention comprises a radiographic diagnostic imaging system 100 which includes a source 110 of penetrative radiation for emitting a radiation beam 118 along a path through a target 112. The radiation source is captured by a detector 162 positioned in the beam path for receiving the radiation; Detector 162 is a direct radiographic detector comprising a plurality of radiation sensors 164 arrayed in rows and columns of the type described in U.S. Pat. No. 5,319,206 issued to Lee et al. on Jun. 7, 1997. According to the present invention, there is placed in front of the detector 162, between the detector and the target 112, an anti-scatter grid 140 having a plurality of radiation absorbing elements, vanes 160. In the illustration the vanes 160 are oriented parallel to the detector's columns of sensors. However this is not critical, and the vanes can be oriented at an angle to the detector rows and columns, as illustrated in FIG. 3. The anti-scatter grid is mounted so as to be moveable relative to the detector and radiation beam through a supporting and moving mechanism represented by block 146. The drive shown is given by way of illustration rather than limiting the way in which the variable speed profile is achieved. A any other mechanical or electromechanical arrangement that will provide the necessary motion to the antiscatter grid, that is will accelerate and decelerate the grid at the required rates, preferably in accordance with the equations given earlier in this description, may be used. The motion imparted by the mechanism is in the direction of the arrow "A" and is preferably in a direction perpendicular to the vanes 160. The system further comprises a controller 170 adapted to synchronize the radiation exposure to the motion of the grid. Controller 170, which may be a computer, is used to begin the radiation emission from source 110 when the grid velocity is at a desired point, preferably right after it has reached its maximum and the deceleration cycle has just begun. The invention also comprises a method whereby grid generated artifacts are reduced by moving the anti-scatter grid unidirectionally during the full radiation exposure using a continuously decreasing rate of movement of the grid. This is done by imparting a single stroke motion to the grid whereby the grid is first accelerated to a first maximum velocity and then decelerated with a decelerating velocity profile, preferably one which approaches zero asymptotically. For example, the decelerating velocity profile may comprise V=K.sub.2 t.sup.-m. The accelerating speed profile is not important so long as it can produce the desired velocity within a short time, of the order of a few milliseconds. The accelerating profile may be a linear function such as V=K.sub.1 t The variables are as described above, and more preferably V(cm/sec)=2,236 t(sec) for t less than or equal to 0.005 seconds and V=25*(t*1,000).sup.-0.5 for t greater than 0.005 seconds and less than or equal to 2 seconds where V is in cm/sec and t is in seconds. The method steps include moving the grid in a direction perpendicular to its vanes with the grid oriented so that it traverses the detector in a direction perpendicular to the detector rows or columns of sensors when the grid vanes are aligned with either the rows or columns of the detector. Alternatively, the grid may be moved in a direction that is at an acute angle to its vanes. In still an alternate embodiment the motion of the grid may be perpendicular to its vanes but with the grid vanes forming an acute angle with the rows or columns of the detector. This angle is preferably selected to be 45.degree.. The advantage of the last two alternatives is that the dead spaces between detector columns (or rows) never align with the grid vanes therefore further reducing the Moire pattern formation as the grid travels over the detector. The disadvantage is that it is more complicated to implement this type of oblique translation of the grid in existing equipment, and may require a larger grid. In practicing the present method, the beginning of the x-ray exposure is timed to assure that the grid is moving at a sufficient velocity during the exposure. Such timing may comprise an initial delay to allow the grid to reach a predetermined speed, it may comprise a chosen start time to produce a desired average velocity, or it may preferably comprise a chosen start time so that the x-ray generator radiation emission pulses begin at maximum velocity (point 34 on FIG. 2) just as the grid begins decelerating. The method of controlling the grid may comprise starting the radiation exposure at any position in the grid motion optimized for a particular grid, radiation source, or examination procedure. Those skilled in the art having the benefit of the teachings of the present invention as hereinabove set forth, can effect numerous modifications thereto. These modifications are to be construed as being encompassed within the scope of the present invention as set forth in the appended claims wherein |
claims | 1. A collimator for shaping a beam of energy emitted from a focal spot of a beam source, the collimator comprising: two elongated parallel plates arranged side by side to define an elongated collimating slit between the plates, wherein at least one of the plates is movably relative to the other plate for varying a width of the collimating slit; and a movable cam operatively arranged with respect to the at least one movable plate such that movement of the cam in a first direction causes the width of the collimating slit to increase, and movement of the cam in a second direction causes the width of the collimating slit to decrease, wherein the at least one movable plate includes a follower contacting a cam surface of the cam, and wherein the cam surface includes steps. 2. A collimator according to claim 1 , wherein the cam is rotatably movable. claim 1 3. A collimator according to claim 2 , further including a motor having a rotatable shaft coupled to the cam. claim 2 4. A collimator according to claim 1 , further including a motor for moving the cam. claim 1 5. A collimator according to claim 4 , wherein the motor comprises a stepping motor, and the collimator also includes a motor controller having a counter for counting steps of the steping motor and a memory for saving the count. claim 4 6. A computed tomography scanner including a collimator according to claim 1 , and further including: claim 1 an annular gantry rotatable about a rotation axis; a beam source mounted within the gantry and having a focal spot for emitting an x-ray beam through the rotation axis; and an array of x-ray detectors for receiving the x-ray beam from the focal spot; wherein the collimator is mounted within the gantry between the focal spot and the detectors for collimating the x-ray beam. 7. A scanner according to claim 6 , wherein the collimator is located between the rotation axis of the gantry and the detectors. claim 6 8. A scanner according to claim 7 , wherein the plates of the collimator each have curved side profiles sharing a common axis of curvature intersecting the focal spot of the beam source. claim 7 9. A collimator according to claim 1 , wherein both plates are movable. claim 1 10. A collimator according to claim 1 , further including at least one spring biasing the movable plate towards the cam. claim 1 11. A collimator according to claim 10 , wherein the spring biases the plates together and the cam is positioned between the plates. claim 10 12. A collimator according to claim 10 , wherein the spring comprises a compression band of resilient material stretched between the plates. claim 10 13. A collimator according to claim 1 , wherein the plates each have curved side profiles sharing a common axis of curvature. claim 1 14. A collimator according to claim 1 , wherein the collimating slit has a predetermined minimum width. claim 1 15. A collimator for shaping a beam of energy emitted from a focal spot of a beam source, the collimator comprising: two elongated parallel plates arranged side by side to define an elongated collimating slit between the plates, wherein at least one of the plates is movably relative to the other plate for varying a width of the collimating slit; and a movable cam operatively arranged with respect to the at least one movable plate such that movement of the cam in a first direction causes the width of the collimating slit to increase, and movement of the cam in a second direction causes the width of the collimating slit to decrease, wherein the at least one movable plate includes a follower contacting a cam surface of the cam, and wherein the follower is rotatable. 16. A collimator according to claim 15 , wherein the cam surface is substantially smooth. claim 15 17. A collimator according to claim 15 , wherein both plates are movable. claim 15 18. A computed tomography scanner including a collimator according to claim 15 , and further including: claim 15 an annular gantry rotatable about a rotation axis; a beam source mounted within the gantry and having a focal spot for emitting an x-ray beam through the rotation axis; and an array of x-ray detectors for receiving the x-ray beam from the focal spot; wherein the collimator is mounted within the gantry between the focal spot and the detectors for collimating the x-ray beam. 19. A scanner according to claim 18 , wherein the collimator is located between the rotation axis of the gantry and the detectors. claim 18 20. A scanner according to claim 19 , wherein the plates of the collimator each have curved side profiles sharing a common axis of curvature intersecting the focal spot of the beam source. claim 19 21. A collimator according to claim 15 , wherein the cam is rotatably movable. claim 15 22. A collimator according to claim 21 , further including a motor having a rotatable shaft coupled to the cam. claim 21 23. A collimator according to claim 15 , further including a motor for moving the cam. claim 15 24. A collimator according to claim 23 , wherein the motor comprises a stepping motor, and the collimator also includes a motor controller having a counter for counting steps of the stepping motor and a memory for saving the count. claim 23 25. A collimator according to claim 15 , further including at least one spring biasing the movable plate towards the cam. together and the cam is positioned between the plates. claim 15 26. A collimator according to claim 25 , wherein the spring biases the plates together and the cam is positioned between the plates. claim 25 27. A collimator according to claim 25 , wherein the spring comprises a compression band of resilient material stretched between the plates. claim 25 28. A collimator according to claim 15 , wherein the plates each have curved side profiles sharing a common axis of curvature. claim 15 29. A collimator according to claim 15 ,wherein the collimating slit has a predetermined minimum width. claim 15 30. A collimator for shaping a beam of energy emitted from a focal spot of a beam source, the collimator comprising: two elongated parallel plates arranged side by side to define an elongated collimating slit between the plates, wherein at least one of the plates is movably relative to the other plate for varying a width of the collimating slit; a movable cam operatively arranged with respect to the at least one movable plate such that movement of the cam in a first direction causes the width of the collimating slit to increase, and movement of the cam in a second direction causes the width of the collimating slit to decrease; a rotatable shaft extending generally normal to the collimating slit, wherein the cam is mounted on the shaft for rotation therewith and the at least one movable plate is slidingly received on the shaft; and a follower extends from the movable plate and contacts a cam surface of the cam; whereby rotation of the shaft and the cam causes the movable plate to slide on the shaft. 31. A computed tomography scanner including a collimator according to claim 14 , and further including: claim 14 an annular gantry rotatable about a rotation axis; a beam source mounted within the gantry and having a focal spot for emitting an x-ray beam through the rotation axis; and an array of x-ray detectors for receiving the x-ray beam from the focal spot; wherein the collimator is mounted within the gantry between the focal spot and the detectors for collimating the x-ray beam. 32. A scanner according to claim 31 , wherein the collimator is located between the rotation axis of the gantry and the detectors. claim 31 33. A scanner according to claim 32 , wherein the plates of the collimator each have curved side profiles sharing a common axis of curvature intersecting the focal spot of the beam source. claim 32 34. A collimator according to claim 30 , further including a motor for rotating the shaft. claim 30 35. A collimator according to claim 34 , wherein the motor comprises a stepping motor, and the collimator also includes a motor controller having a counter for counting steps of the stepping motor and a memory for saving the count. claim 34 36. A collimator according to claim 30 , further including at least one spring biasing the movable plate towards the cam. claim 30 37. A collimator according to claim 36 , wherein the spring biases the plates together and the cam is positioned between the plates. claim 36 38. A collimator according to claim 36 , wherein the spring comprises a compression band of resilient material stretched between the plates. claim 36 39. A collimator according to claim 30 , wherein the plates each have curved side profiles sharing a common axis of curvature. claim 30 40. A collimator according to claim 30 , wherein the collimating slit has a predetermined minimum width. claim 30 41. A collimator according to claim 30 , wherein both plates are movable. claim 30 42. A collimator according to claim 30 , further comprising a linear-rotary bearing supporting the plate on the shaft. claim 30 |
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summary | ||
056132417 | summary | FIELD OF THE INVENTION The present invention relates to the treatment of halogen-containing wastes and other waste materials. More particularly, the invention relates to a process for treating halogenated waste materials by introducing the waste materials into a bath of molten glass containing a sacrificial metal oxide. BACKGROUND OF THE INVENTION Many industrial waste materials, such as radioactive wastes, chemically hazardous wastes, or toxic elemental wastes, present difficult disposal problems due to the hazard of the material and the intensely regulated nature of these materials. When halogens are contained within such waste material, the halogens create unique problems, which are not adequately addressed by conventional treatment methods. Sources of halogen-containing waste include, not only processes that create halide salts or halogens-containing compounds, but also wastes contaminated by radioactive or other hazardous materials. A chloride salt purification process used to process plutonium in a weapons complex is one example of a process generating halogen-containing waste. That process generates large quantities of chloride-containing radioactive wastes. Halogen-containing plastics, used as components of radioactive or chemically hazardous wastes streams or equipment are another example of halogen-containing wastes. These also require treatment for such disposal. Halogen-containing compounds are also used to create chemically resistant or fireproof manufactured goods which also present unique disposal problems. Examples include many plastics, such as Hypalon.RTM. plastic, a product of Dupont, Wilmington, Del., which are used to insulate electrical wire. Hypalon.RTM. plastic also contains valuable lead oxide that should be recovered both for recycling and to avoid long-term problems with the toxicity of the lead. In other cases, plastic within a waste material must be destroyed to recover copper and other valuable materials. Thus, a problem with many halogen-containing waste materials is that they were originally designed with great care not to burn or be damaged by chemicals. Though very desirable in a manufactured product, these properties cause difficulties in the treatment and disposal of such materials. With their water solubility, another significant problem with many halogen-containing wastes and waste forms (e.g. salts) is leaching and groundwater contamination. If a high quality waste form is to be made, the halogens should be removed from the waste before or during conversion to a waste form used for disposal. SUMMARY OF THE INVENTION The present invention addresses problems existing with the treatment of halogen-containing waste materials as well as other waste materials. Importantly, the present invention separates halogens from a waste material being treated, such as radioactive and hazardous wastes, and, thus, allows production of a non-halogen, high durability waste form. High concentrations of halogens in wastes imply low performance by the final waste disposal form. A second important feature of the invention is its ability to recycle reagents from the waste material treatment process. This is important to both minimize costs and avoid creating other hazardous wastes. Third, the present invention may be used to treat and/or destroy wastes made from materials designed to be difficult to destroy. Accordingly, in one embodiment, the present invention provides a process for treating a halogen-containing waste material. This process comprises the steps of: providing a bath of molten glass containing a sacrificial metal oxide capable of reacting with a halogen in the waste material, where the sacrificial metal oxide is present in at least a stoichiometric amount with respect to the halogen in the waste material; introducing the waste material into the bath of molten glass to cause a reaction between the halogen in the waste material and the sacrificial metal oxide to yield a metal halide which is a gas at the temperature of the molten glass; separating the gaseous metal halide from the molten glass; contacting the gaseous metal halide with an aqueous scrubber solution of an alkali metal hydroxide to yield a metal-containing precipitate and a soluble alkali metal halide; separating the precipitate from the aqueous scrubber solution; and removing the molten glass containing the treated waste material as a waste glass. In another embodiment, the present invention provides a process for the conversion of waste material containing halogen-containing compounds, elemental metal or carbon-containing compounds. This process comprises the steps of: providing a bath of molten glass containing at least one sacrificial metal oxide, where at least one sacrificial metal oxide is capable of reacting with a halogen in the waste material, or of oxidizing elemental metal or carbon-containing compounds within the waste material, and the sacrificial metal oxide is present in at least a stoichiometric amount for the reaction or the oxidation; introducing the waste material into the bath of molten glass to cause a reaction between the waste material and the sacrificial metal oxide, wherein at least one sacrificial metal oxide reacts with a halogen in the waste material to yield a metal halide which is a gas at the temperature of the molten glass and at least one sacrificial metal oxide oxidizes elemental metal to form an oxide of the elemental metal which is soluble in the molten glass or oxidizes the carbon-containing compound thereby reducing the sacrificial metal oxide to its elemental metal; separating the gaseous metal halide from the molten glass; contacting the gaseous metal halide with an aqueous scrubber solution of an alkali metal hydroxide to yield a metal-containing precipitate and a soluble alkali metal halide; separating the precipitate from the aqueous scrubber solution; removing the molten glass containing the treated waste material as a waste glass; and recovering the elemental metal derived from the sacrificial metal oxide from the bath. |
062087045 | claims | 1. A method of producing a product isotope by isotopic conversion reaction comprising: providing a target; directing an electron beam having intensity of at least 50 microamps/cm.sup.2 onto a converter to generate a photon beam having photons of energy of at least 8 MeV; and directing the photon beam onto the target to isotopically convert at least a portion of the target to the product isotope. a) the thickness of the target material is about 7.5 centimeters, or less, and b) the photon beam is generated by an electron beam impinging a heavy metal convertor, wherein the electron beam power density within the convertor is about 35,000 watts/cm.sup.3. providing a target; directing an electron beam having an intensity of at least 50 microamps/cm.sup.2 onto a converter to generate a photon beam; and directing the photon beam onto the target to isotopically convert at least a portion of the target to the product isotope. 2. A method of claim 1 wherein: 3. A method of claim 1 wherein the intensity of the electron beam is at least 500 microamps/cm.sup.2. 4. A method of claim 1 wherein the photon beam has a peak energy level of at least 30 MeV. 5. A method of claim 1 wherein the photon beam has a peak energy level of at least 35 MeV. 6. A method of claim 1 wherein the convertor includes at least two separate convertor plates having different thicknesses. 7. A method of claim 6 further including the step of cooling the convertor. 8. The method of producing a product isotope by isotopic conversion reaction comprising: 9. A method of claim 8 wherein the photon beam has a peak energy level of at least 30 MeV. 10. A method of claim 8 wherein the photon beam has a peak energy level of at least 35 MeV. |
052672824 | description | Only the elements essential for understanding the invention are shown. For example, the entire primary and secondary sections of the reactor installation are not shown. The directions of flow of the working media are indicated by arrows. DESCRIPTION OF THE PREFERRED EMBODIMENTS Referring now to the drawings, wherein like reference numerals designate identical or corresponding parts throughout the several views, the containment vessel of the reactor is marked 1 in FIG. 1. A pressure relief line 2, the first part of which is called raw gas line below, leads from the containment vessel to the filter unit 3. A control valve 25 and a bursting disk 26 for defined pressure relief are located in two parallel planes in the raw gas line. A throughput measurement, which is not shown, likewise takes place in the raw gas line. In the case of the example, the filter unit operates according to the wet-filter principle; the type of filter is irrelevant to the use of the invention. In Venturi internals 27, the water is atomized and thus purifies the gases. The purified gases then flow through a water separator 28. They then enter the second part of the pressure relief line 2, called pure gas line 4 below, which leads to the stack 5. In the pure gas line 4, a sampling point 6 is provided, from which a gas sample is continuously taken and introduced into a sampling line 7. The sampling could also be carried out from an exit air duct or exit air stack downstream of the junction of the pure gas line with such a duct or stack. As an example, it may be mentioned that about 10 m.sup.3 /h are branched off from a total exit gas rate of 20,000 m.sup.3 /h. This gas sample is heated by means of a heater 24, operated electrically or by heat exchange, over preferably the entire length of the sampling line 7, in order to avoid condensation. From the sampling line, a part stream is passed to the dilution unit 8. This unit designed as a plurality of stages operates with a defined volumetric flow of particle-free compressed air. This is made available by a compressor 11, upstream of which an iodine filter and an aerosol filter 12 are provided. In principle, only the first two stages of the plurality of dilution stages are heated, since the temperature can no longer fall below the dewpoint after the second stage, even in the case of pure steam in the pure gas line 4. A dilution stage shown in FIG. 2 functions as follows: the compressed air provided flows through an annular gap 29 around the suction nozzle 30 for the gas mixture which is to be diluted. Due to the resulting reduced pressure, the aerosol is drawn in at a certain volumetric flow and mixed homogeneously with the pure air in the mixing chamber 31. If the pure air volumetric flow is increased, the flow velocity in the annular gap increases to the same extent. As a result, the reduced pressure at the suction nozzle increases, whereby the volumetric flow of the gas mixture also increases. Both volumetric flows are thus coupled by the reduced pressure and their ratio remains constant even for different upstream pressures. In the dilution unit 8, a dilution of, for example 1:10.sup.4 is desired. It is of advantage here to accomplish the dilution in a plurality of cascades, which reduces the air requirement for clean dilution air. Only a part of the diluted sample, taken from the sampling line 7, is taken from the mixing chamber 31 and fed to the next stage. This part stream sampling takes place via the extraction nozzle 32. Care must be taken here that this sampling takes place under isokinetic conditions. These apply if the flow velocity in the nozzle 32 is equal to that in the flow channel at the point of extraction. Via various nozzle diameters, different volumetric extraction flows can be adapted to different total volumetric flows. This is of importance for the last stage. As an example, it may be assumed that only 0.3 m.sup.3 /h from the total extraction rate of 10 m.sup.3 /h indicated above are utilized for the measurement. After the dilution, however, a total of about 3 m.sup.3 /h is fed to the measuring apparatus. The residual air of 7.8 m.sup.3 /h for all dilution stages, remaining after the isokinetic part stream sampling, flows through the exit air branch 33 outwards into the return line 9 (FIG. 1). In this return line, the remainder of the sample air and the air from the exit air branches 33 are delivered by a pump 10 back into the pure gas line 4. With this return, care must be taken that no back-pressure, which might affect the dilution ratio, is generated in the mixing chamber of the dilution stage. As a result of the dilution, the activity of the precipitated substances is reduced to a level which is also usual in normal operation. The handling and evaluation of the measuring apparatus can thus take place in the usual manner even in the event of an accident. The measurement line 13 (FIG. 1) which leads to the actual measurement section 14 is connected to the extraction nozzle 32 of the last stage. This measurement section 14, provided also for normal operation of the installation and shown greatly simplified, consists, on the one hand, of a combination 15 of aerosol- and iodine-balancing filters for discontinuous measurement. For quasi-continuous monitoring, an aerosol monitor 16, an iodine monitor 36 and a rare gas monitor 37 are provided. These three monitors are each fitted with a radiation detector 38. The relative activity of the aerosols is detected by this mere indication. Furthermore, it defines the change intervals of the balancing filters. In the normal operation, discontinuous measurement via the elements 15 is normally carried out once per week. In the event of an accident, however, it is envisaged that the measurement is carried out every 4 hours. For this purpose, the balancing filters are dismantled, transferred into a separate room and evaluated there by means of a spectrometer for specific nuclides. Prior to the measurement, the measurement section 14 is flushed, so that the filter activity caused by the rare gas components is reduced to an irrelevant level and does not falsify the actual measurement. For this purpose, the measurement line 13 is isolated by means of a shut-off device 34, and the flushing air line 17 is opened by means of a shut-off device 35. Atmospheric air is drawn in via the transfer pump 19 and passes via iodine- and aerosol-filters 18 into the measurement section. The flushing air is expelled into the return line 9. Of course, it can also be discharged directly into the stack 5. During the actual measurement, the flushing air line 17 is isolated by the shut-off device 35, and the measurement line 13 is opened by the shut-off device 34. The mixture to be measured is drawn in by the same transfer pump 19. Since this pump is designed for the higher flushing air rate, atmospheric air is also drawn in for control purposes in the case of measurement. For this purpose, there is a control valve 20 with an upstream aerosol filter 21 in a branch line upstream of the pump. Upstream of the junction of the part measurement section having the aerosol monitor with that having the balancing filters, a flow rate meter 22 is located in the latter. The flow rate via the aerosol- and iodine-filters is measured therein and integrated over the dust introduction time. In this way, the activity concentration values are determined. To derive the activity relief rate, the concentration is correlated with the throughput measurement. Even though, as mentioned above, the throughput is measured in the raw gas line, a correlation with this measurement would lead to erroneous results. This is because, during a pressure relief, the throughputs in the raw gas line and in the pure gas line can be very different particularly in the initial phase. This can, for example, be due to the steam content condensing out in the still cold water receiver of the filter unit 3. Consequently, the flow is determined once more at 23 in the pure gas line 4. This can be a Venturi measurement or a determination via a pressure/temperature measurement. The result is linked to the concentration values in order to determine the activity relief rates. Obviously, numerous modifications and variations of the present invention are possible in the light of the above teachings. It is therefore to be understood that, within the scope of the appended claims, the invention may be practiced otherwise than as specifically described herein. |
abstract | A reactivity control rod adapted to be used in a reactor core of a fast reactor and disposed at a substantially central portion of the reactor core for controlling a reactivity therein. The reactivity control rod includes a wrapper tube surrounded by a plurality of fuel rods in a reactor core, and a plurality of neutron absorber rods arranged in the wrapper tube. At least one of the plurality of neutron absorber rods includes a cladding tube and a mixture filled in the cladding tube. The mixture is composed of a neutron absorber that absorbs a neutron and a neutron moderator that moderates the neutron. |
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summary | ||
042723213 | claims | 1. For use in a nuclear reactor installation having a vertically reciprocable crane above a vertically oriented reactor vessel, the vessel containing a core, a plurality of control rods in the core and an upper guide structure in which are disposed a plurality of control rod drive shafts connected to the control rods, an apparatus for removing and reinserting the upper guide structure and control rods from the opened reactor vessel, comprising: a rigid, vertical frame for placement on the upper guide structure having means at its lower end for selective attachment to the upper guide structure and having interference means at its upper end; and a control rod support member vertically reciprocable within said frame, said support member including a horizontal platform having gripper means on its underside for simultaneously engaging each drive shaft, said platform including link means connectable to the crane for reciprocating said support member whereby the control rods may be withdrawn from the reactor core into the upper guide structure, said member further having selected surfaces adapted to abut said interference means when said platform is in its maximum vertical position relative to said frame whereby continued upward force by the crane is transferred through the link means to the frame so that the upper guide structure and control rods may be removed simultaneously from the reactor vessel. 2. The apparatus of claim 1 further comprising stop means between said frame and said support member for locking and unlocking said platform in the maximum vertical position relative to said frame. 3. The apparatus of claim 2 wherein said stop means is actuated in response to the position of said frame relative to the vessel. 4. The apparatus of claim 2 wherein said frame comprises a plurality of spaced apart, parallel columns connected at their tops by rigid horizontal beams, said beams forming said interference means. 5. The apparatus of claim 4 wherein said link means includes a plurality of link blocks located on the upper surface of said platform such that said blocks contact said beams when said platform is in the maximum vertical position relative to said frame. 6. The apparatus of claim 5 wherein said beams are detachable from said frame. 7. The apparatus of claim 5 wherein the lower ends of said columns include said means for attaching the upper guide structure. 8. The apparatus of claim 7 wherein said stop means is actuated in response to the position of said frame relative to the vessel. 9. The apparatus of claim 8 wherein said stop means includes a vertically movable actuator rod; and latch means pivotally connected to said frame and to said actuating rod, said latch means having a stop position for preventing said plate from moving below said latch and a neutral position for permitting unobstructed vertical motion of said platform, said latch means being responsive to the vertical movement of said actuating rod. 10. The apparatus of claim 9 wherein said latch means is in the stop position except when said actuating rod is in mechanical contact with said vessel. 11. The apparatus of claim 10 wherein said stop means further includes a collar fixedly attached around said column, said collar carrying a stop support to which said latch means is pivotally connected. 12. The apparatus of claim 11 further comprising a dead bolt carried by the platform and horizontally slidable relative thereto; and wherein said latch support has an upper porting forming a support bar onto which said dead bolt can be placed. |
claims | 1. A particle beam emitter comprising:a hollow particle beam tube having a first end portion, a second end portion, and a longitudinal axis; andan electromagnetic system comprising a voltage supply electrically coupled to the hollow particle beam tube and configured to generate a primary electrical current flowing axially in the hollow particle beam tube from the first end portion towards the second end portion, whereby a primary magnetic field associated with the primary electrical current is operable to induce a secondary electrical current in a plasma located within the hollow particle beam tube, the secondary electrical current flowing generally axially within the plasma and causing the plasma to contract inwardly towards the longitudinal axis. 2. The particle beam emitter of claim 1, further comprising a plurality of electromagnetic coils aligned axially with and surrounding at least a portion of the hollow particle beam tube for generating an axial magnetic field within the hollow particle beam tube. 3. The particle beam emitter of claim 2, further comprising an insulation member positioned between the plurality of electromagnetic coils and the hollow particle beam tube. 4. The particle beam emitter of claim 1, further comprising a pair of electromagnets positioned exterior to the hollow particle beam tube and rotatable about the longitudinal axis. 5. The particle beam emitter of claim 1, wherein the voltage supply is configured to generate the primary electrical current by creating a potential difference between the first end portion and the second end portion. 6. The particle beam emitter of claim 1, wherein the first end portion is configured to receive plasma from a fuel injector. 7. The particle beam emitter of claim 6, further comprising a pressure valve positioned between the fuel injector and the first end portion of the hollow particle beam tube. 8. The particle beam emitter of claim 1, wherein the hollow particle beam tube is made from at least one of tungsten and graphite. 9. The particle beam emitter of claim 1, wherein the hollow particle beam tube comprises a hollow graphene cylinder having an inner surface coated with tantalum hafnium carbide. |
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051184478 | claims | 1. A method of denitrification of nitrates and nitrites in an aqueous stream, comprising the steps of: (a) preparing a mixture by causing formate to be present in the aqueous stream, (b) heating the mixture of step (a) to a predetermined reaction temperature under sufficient pressure to maintain the aqueous stream in an aqueous liquid or supercritical phase, and (c) holding the mixture at the conditions in step (b) for a residence time sufficient to convert the nitrogen in the nitrates and nitrites to nitrogen gas and forming an aqueous product. mixing carbon dioxide with said aqueous product to form a carbonate. (a) heating the aqueous stream to a predetermined reaction temperature under sufficient pressure to maintain the aqueous stream in an aqueous liquid or supercritical phase, (b) preparing a mixture by causing formate to be present in the aqueous stream, and (c) holding the mixture of step (b) at the conditions of step (a) for a residence time sufficient to convert the nitrogen in the nitrates and nitrites to nitrogen gas and form an aqueous product. mixing carbon dioxide with said aqueous product to form a carbonate. (a) preparing a mixture by causing formate to be present in the aqueous stream, (b) heating the mixture of step (a) to a predetermined reaction temperature under sufficient pressure to maintain the aqueous stream in an aqueous liquid or supercritical phase to begin forming reaction products, and (c) holding both the mixture and reaction products at the conditions of step (b) for a residence time sufficient to convert the nitrogen in the nitrates and nitrites to nitrogen gas and form an aqueous product. mixing carbon dioxide with said aqueous product to form a carbonate. (a) heating the aqueous stream to a predetermined reaction temperature under sufficient pressure to maintain the aqueous stream in an aqueous liquid or supercritical phase, (b) causing formate to be present in the aqueous stream, and (c) holding both the mixture of step (b) and accumulating reaction products at the conditions of step (a) for a residence time sufficient to convert the nitrogen in the nitrates and nitrites to nitrogen gas and form an aqueous product. mixing carbon dioxide with said aqueous product to form a carbonate. 2. The method as recited in claim 1, further comprising the step of: 3. The method as recited in claim 1, wherein formate is caused to be present in the aqueous stream. 4. The method as recited in claim 1, wherein formate is caused to be present in the aqueous stream by synthesizing formate in the aqueous stream. 5. The method as recited in claim 1, wherein the aqueous stream has a pH between about 0 to 7. 6. The method as recited in claim 1, wherein the aqueous stream has a pH between about 7 and 14. 7. The method as recited in claim 1, wherein the residence time is from about one minute to about two hours. 8. The method as recited in claim 1, wherein an amount of formate is substantially in stoichiometric ratio to the nitrates and nitrites, resulting in a final concentration of nitrites and nitrates below 44 mg/l. 9. The method as recited in claim 1, wherein the mixture is heated to a temperature of from about 200.degree. C. to about 600.degree. C. 10. The method as recited in claim 9, wherein the mixture is heated from about 250.degree. C. to about 350.degree. C. 11. The method as recited in claim 1, wherein the pressure is from about at least 240 psi to about at least 3200 psi. 12. The method as recited in claim 11, wherein the pressure is from about at least 580 psi to about at least 2900 psi. 13. The method as recited in claim 1, wherein the residence time is determined by the predetermined reaction temperature. 14. A method of denitrification of nitrates and nitrites in an aqueous stream, comprising the steps of: 15. The method as recited in claim 14, further comprising the step of: 16. The method as recited in claim 14, wherein sodium formate is caused to be present in the aqueous stream by adding sodium formate to the aqueous stream. 17. The method as recited in claim 14, wherein formate is caused to be present in the aqueous stream by synthesizing formate in the aqueous stream. 18. The method as recited in claim 14, wherein the aqueous stream has a pH between about 0 and 7. 19. The method as recited in claim 14, wherein the aqueous stream has a pH between about 7 and 14. 20. The method as recited in claim 14, wherein the residence time is from about one minute to about two hours. 21. The method as recited in claim 1, wherein an amount of formate is substantially in stoichiometric ratio with the nitrates and nitrites, resulting in a final concentration of nitrites and nitrates below 44 mg/l. 22. The method as recited in claim 14, wherein the mixture is heated to a temperature of from about 200.degree. C. to about 600.degree. C. 23. The method as recited in claim 22, wherein the mixture is heated from about 250.degree. C. to about 350.degree. C. 24. The method as recited in claim 14, wherein the pressure is from about at least 240 psi to about at least 3200 psi. 25. The method as recited in claim 24, wherein the pressure is from about at least 580 psi to about at least 2900 psi. 26. The method as recited in claim 14, wherein the residence time is determined by the predetermined reaction temperature. 27. A method of denitrification of nitrates and nitrites in an aqueous stream, comprising the steps of: 28. The method as recited in claim 27, further comprising the step of: 29. The method as recited in claim 27, wherein formate is caused to be present in the aqueous stream by adding formate to the aqueous stream. 30. The method as recited in claim 27, wherein formate is caused to be present in the aqueous stream by synthesizing formate in the aqueous stream. 31. The method as recited in claim 27, wherein the aqueous stream has a pH between about 0 and 7. 32. The method as recited in claim 27, wherein the aqueous stream has a pH between about 7 and 14. 33. The method as recited in claim 27, wherein the residence time is from about one minute to about two hours. 34. The method as recited in claim 27, wherein an amount of formate is substantially in stoichiometric ratio to the nitrates and nitrites, resulting in a final concentration of nitrites and nitrates below 44 mg/l. 35. The method as recited in claim 27, wherein the mixture is heated to a temperature of from about 200.degree. C. to about 600.degree. C. 36. The method as recited in claim 35, wherein the mixture is heated from about 250.degree. C. to about 350.degree. C. 37. The method as recited in claim 27, wherein the pressure is from about at least 240 psi to about at least 3200 psi. 38. The method as recited in claim 37, wherein the pressure is from about at least 580 psi to about at least 2900 psi. 39. The method as recited in claim 27, wherein the residence time is determined by the predetermined reaction temperature. 40. A method of denitrification of nitrates and nitrites in an aqueous stream, comprising the steps of: 41. The method as recited in claim 40, further comprising the step of: 42. The method as recited in claim 40, wherein formate is caused to be present in the aqueous stream by adding formate to the aqueous stream. 43. The method as recited in claim 40, wherein formate is caused to be present in the aqueous stream by synthesizing formate in the aqueous stream. 44. The method as recited in claim 40, wherein the aqueous stream has a pH between about 0 and 7. 45. The method as recited in claim 40, wherein the aqueous stream has a pH between about 7 and 14. 46. The method as recited in claim 40, wherein the residence time is from about one minute to about two hours. 47. The method as recited claim 40, wherein an amount of formate is substantially in stoichiometric ratio to the nitrates and nitrites, resulting in a final concentration of nitrates and nitrites below 44 mg/l. 48. The method as recited in claim 40, wherein the mixture is heated to a temperature of from about 200.degree. C. to about 600.degree. C. 49. The method as recited in claim 48, wherein the mixture is heated from about 250.degree. C. to about 350.degree. C. 50. The method as recited in claim 40, wherein the pressure is from about at least 240 psi to about at least 3200 psi. 51. The method as recited in claim 50, wherein the pressure is from about at least 580 psi to about at least 2900 psi. 52. The method as recited in claim 40, wherein the residence time is determined by the predetermined reaction temperature. |
summary | ||
description | 1. Field of the Invention The present invention relates to a concrete cask suitable for the transportation or long-term storage of radioactive material such as spent nuclear fuels. 2. Description of the Related Art Concrete casks described in Japanese Patent Applications Laid-open No. 2001-141891 and Japanese Patent No. 3342994 are known as the conventional concrete casks. Japanese Patent Application Laid-open No. 2001-141891 describes a representative conventional concrete cask provided in the top part thereof with a gas outlet opening and in the lower part thereof with a gas inlet opening. In this structure, convection is generated in a gap between the concrete cask and a canister so as to introduce outside air through the inlet opening and release it through the outlet opening. As a result, heat is removed from the canister (sealed container containing the spent fuel) that is stored inside the concrete cask. Japanese Patent No. 3342994 described a metal cask structure in which a neutron shielding material is provided between an outer shell and an inner shell. In order to enhance the heat transfer between the outer and inner shells, both ends of heat transfer fins made from a metal material with good thermal conduction, such as copper, are connected in their entirety to the inner shell and outer shell. The heat transfer fins are provided radially along the radial direction. In the structure of Japanese Patent Application Laid-open No. 2001-141891, heat is removed by providing gas inlet and outlet openings and introducing outside air. In this case, corrosion-inducing substances such as sea salt particles contained in the outside air are unavoidably introduced into the concrete cask and adhere to the canister surface. As a result the canister surface is corroded and sometimes stress corrosion cracking can occur under the combined effect with the residual stresses present in the vicinity of welds in the canister. Such cracking means that the containment of canister is disrupted and radioactive material can be emitted to the outside. Furthermore, because the above-mentioned openings serving as the inlet and outlet were the portions that were not covered with a shielding body (portions that lack shielding), radiation streaming from those openings could not be avoided. In the configuration described in Japanese Patent 3342994, the inner shell and outer shell were connected by both ends of the heat transfer fins in their entirety. Therefore, the problem was that no shielding body was present in the heat transfer fin portions and radiation penetrated through the heat transfer fins and streamed in the radial direction. Furthermore, because of the structure in which the heat transfer fins were in contact with the inner and outer shells, the neutron shielding material such as a concrete had to be placed in the spaces bounded by the inner and outer shells and heat transfer fins one by one, or structural blocks had to be assembled. In this case the manufacture was a time-consuming operation. It is an object of the present invention to provide a concrete cask that is effective in suppressing the radiation streaming and is easy to manufacture. Problems addressed by the present invention are described hereinabove. In order to solve the above mentioned problems according to the present invention, a concrete cask in which a shielding body composed of concrete and heat transfer fins made from metal are provided between an inner shell and an outer shell made from metal and which comprises an accommodation portion formed inside the inner shell for accommodating a radioactive substance, a containment structure is employed to shield the accommodation portion from the outside of the cask, and in the heat transfer fins, the portions thereof at the inner shell-side are provided in contact with the inner shell and the portions thereof at the outer shell-side are cut so as to form a separation portion with respect to the outer shell, or the portions thereof at the outer shell-side are provided in contact with the outer shell and the portions thereof at the inner shell-side are cut so as to form a separation portion with respect to the inner shell. These and other objects, features, and advantages of the present invention will become more apparent upon reading the following detailed description along with the accompanying drawings. The basic structure of a concrete cask and the structure of heat transfer fins in the concrete cask will be described below. FIG. 1 is a perspective view with a partial cut-out illustrating the storage state of the concrete cask of the first embodiment of the present invention. FIG. 2A is a longitudinal sectional view of the concrete cask of the first embodiment, FIG. 2B is a lateral sectional view. The concrete cask A of the first embodiment shown in FIG. 1 and FIG. 2 is composed of a tubular container body 1 open at both ends. A canister (a) is provided inside the concrete cask A. The container body 1 has a structure in which a concrete container 3 is covered with an outer shell 4 made from carbon steel, a bottom cover 5 made from carbon steel, a thick flange made from carbon steel, and an inner shell 7 made from carbon steel. An accommodation portion for accommodating the canister (a) is constructed inside the inner shell 7 (inside the container body 1). A lid 2 has a structure in which a concrete lid member 8 is covered with a thick upper lid 9 made from carbon steel and a lower cover 10 made from carbon steel. Multiple heat transfer fins 11 made from copper, carbon steel, or aluminum alloy are embedded and installed in the container 3 so as to be connected to the inner wall of the outer shell 4, as shown in FIG. 1 or FIG. 2B. The heat transfer fins are not required to be provided along the entire length in the axial direction of the container and may be provided only in the zones necessary for heat emission. For example, it is not particularly necessary to provide the heat transfer fins in the portion below the canister. Disposing the lid 2 on the container body 1 seals the space (accommodation portion) inside the inner shell 7 and shields the space from the outside of the concrete cask A. A seal monitoring device 12 is installed in the lid 2 to check the sealing state (see FIG. 1). The canister (a) is a sealed container composed of a container body 13 and a lid 14. The inside thereof is filled with a radioactive substance (x) such as spent nuclear fuel. As shown in FIG. 2B, multiple heat transfer fins 11 are provided equidistantly between the inner shell 7 and outer shell 4 in the tangential direction for enhancing the dissipation of heat emitted from the radioactive substance (x) to the outside of the concrete cask A. Respective heat transfer fins 11 are formed to have a flat shape (I-like shape in a lateral sectional view) and are disposed radially along the radial direction of the container 3. The end portions of the respective heat transfer fins 11 at the side of the outer shell 4 are connected to the inner wall of the outer shell 4, whereas the end portions thereof at the side of the inner shell 7 are provided with separation portions with respect to the outer wall of the inner shell 7. Thus, the ends at the inner side of heat transfer fins 11 are cut out and the end portions are located at an appropriate distance from the inner shell 7. As for the cut portions, the cutting is conducted along the entire axial direction of the container, and the heat transfer fins 11 and the inner shell 7 are completely separated. In the structure of the first embodiment, even if the radiation penetrates through the heat transfer fins 11 in the radial direction, because a separation portion is present between the inner shell 7 and the heat transfer fins 11, the radiation has to pass through the concrete 3 of the separation portion. It means that even when the radiation leaks in the radial direction, it has to pass through the concrete 3 serving as a shielding body, and the structure of the concrete cask A with excellent radiation shielding capacity can be provided. Another advantage of this structure is that the container 1 body can be manufactured easily. Thus, when the container 1 is manufactured, the inner and outer shells 7 and 4 are formed and then fresh concrete 3 is placed between the inner and outer shells 7 and 4. With respect to this issue, when the conventional configuration (configuration shown in FIG. 11), such as described in Japanese Patent No. 3342994, is manufactured, a fresh concrete 3 has to be placed in all the cells one by one (that is, in all the spaces separated by respective heat transfer fins 30 shown in FIG. 11). However, in the configuration of the present embodiment, the individual cells are linked together by the separation portion, and even when the fresh concrete 3 is poured in only one place, the fresh concrete can spread to all the cells. Therefore, the number of production process is reduced. Furthermore, the fact that the heat transfer fins 11 and the inner shell 7 are completely separated means that the inner and outer shells 7 and 4 are not connected by the heat transfer fins 11. Therefore, a manufacturing process can be employed by which the inner shell 7 and the outer shell 4 are produced separately in advance and then assembled. As a result, in this sense, too, the structure of the first embodiment can be advantageous in terms of reducing the number of production process. The above-described effects are also demonstrated in the second to eighth embodiments described hereinbelow. All those embodiments will be explained below. FIGS. 3 through 9 are the lateral sectional views of the second to eighth embodiments. In the second embodiment illustrated by a lateral sectional view in FIG. 3, the end portions of the heat transfer fins 11 at the side of the inner shell 7 are connected to the outer wall of the inner shell 7, whereas the end portions at the side of the outer shell 4 are disposed via a separation portion with respect to the inner wall of the outer shell 4. Thus, the heat transfer fins 11′ are disposed at a certain distance from the outer shell 4, that is the structure is inversed with respect to that of the first embodiment (FIG. 2B). In the third embodiment illustrated by a lateral sectional view in FIG. 4, the end portions of the heat transfer fins 18 at the side of the outer shell 4 are connected to the inner wall of the outer shell 4, whereas the end portions at the side of the inner shell 7 (ends that form a separation portion with respect to the inner shell 7) are bent at an almost right angle along the appropriate width to obtain an L-like shape. As a result, the portions that were bent (bent portions) form opposite surfaces that face the outer wall of the inner surface 7 at an appropriate distance therefrom (separation portion). In the fourth embodiment illustrated by a lateral sectional view in FIG. 5, the end portions of the heat transfer fins 18′ at the side of the inner shell 7 are connected to the outer wall of the inner shell 7, whereas the end portions at the side of the outer shell 4 (ends that form a separation portion with respect to the outer shell 4) are bent at an almost right angle along the appropriate width to obtain an L-like shape. As a result, the portions that were bent (bent portions) form opposite surfaces that face the inner wall of the outer shell 4 at an appropriate distance therefrom (separation portion). In the above-described third and fourth embodiments, the heat transfer fins 18, 18′ have such bent portions. Therefore, a large surface area of the surfaces (opposite surfaces) of the heat transfer fins 18, 18′ that face the inner shell 7 or outer shell 4 can be ensured. As a result, heat transfer can be enhanced and a concrete cask A with excellent cooling capacity can be obtained. In the configuration of the fifth embodiment illustrated by a lateral sectional view in FIG. 6, first heat transfer fins 21 and second heat transfer fins 22 are disposed alternately and equidistantly in the tangential direction of the container 3. The first heat transfer fins 21 are cut so that the end portions thereof at the side of the outer shell 4 are connected to the inner wall of the outer shell 4, whereas the end portions thereof at the side of the inner shell 7 form a separation portion with respect to the outer wall of the inner shell 7. The second heat transfer fins 22 are cut so that the end portions thereof at the side of the inner shell 7 are connected to the outer wall of the inner shell 7, whereas the end portions thereof at the side of the outer shell 4 form a separation portion with respect to the inner wall of the outer shell 4. Heat transfer fins of one type (21 or 22) are disposed so as to be inserted between the adjacent fins (22 or 21) of the other type. As a result, the first heat transfer fins 21 and second heat transfer fins 22 have overlap portions in the radial direction of the container 3. In the structure of the fifth embodiment, the first heat transfer fins 21 and second heat transfer fins 22 have overlapping portions. Therefore, the advantage of this structure is that heat transfer between the heat transfer fins 21 and 22 is enhanced and excellent cooling effect is attained. Another merit of this structure is that because the heat transfer fins 21, 22 are formed to have a flat shape without bent portions, as in the first and second embodiments (the so-called I-like shape), bending of the heat transfer fins 21, 22 is not required and the number of processing operations can be reduced. In the sixth embodiment illustrated by the lateral sectional view in FIG. 7, the heat transfer fins 11 of the first embodiment are inclined at a prescribed angle from the radial direction of the container 3 (reference symbol 11b). A structure can be also considered in which the heat transfer fins 11′ of the second embodiment are similarly inclined at a prescribed angle from the radial direction (this structure is not shown in the figures). In the seventh embodiment illustrated by the lateral sectional view in FIG. 8, the portions of the heat transfer fins 18 of the third embodiment, which follow the radial direction of the container 3 (portions other than the aforesaid bend portions), are inclined at a prescribed angle from the radial direction of the container 3 (reference symbol 18b). A structure can be also considered in which the heat transfer fins 18′ of the fourth embodiment are similarly inclined at a prescribed angle from the radial direction (this structure is not shown in the figures). In the eighth embodiment illustrated by the lateral sectional view in FIG. 9, the first heat transfer fins 21 and second heat transfer fins 22 of the fifth embodiment are similarly inclined at a prescribed angle from the radial direction (reference symbols 21b, 22b). In those sixth to eighth embodiments, the heat transfer fins (11b, 18b, 21b, 22b) are disposed in an inclined state so as to decline from the radiation direction (radial direction of the container 3). The effect of such an arrangement is that streaming of radiation in the radial direction can be suppressed even more reliably. Further, the heat transfer capacity (heat removal capability) of the concrete cask will be explained hereinbelow with reference to the case in which heat transfer fins 21, 22 are installed alternately in a zigzag manner, as in the fifth embodiment. FIG. 10 is a partially expanded lateral sectional view of the container of the fifth embodiment, and FIG. 11 is a partially expanded lateral sectional view of the container with the configuration of the comparative reference example (conventional technology). It is well known that the equation relating to heat conduction can be represented by the following equation [A]:Q=λ×S×ΔT/L [A]where:λ: thermal conductivity of a thermally conductive substance (W/m·K);S: surface area of the heat transfer path of the thermally conductive substance (heat transfer surface area perpendicular to the direction of heat flux) (m2);ΔT: difference in temperature between the inner shell and outer shell (K);L: length of the heat transfer path (m). In the above-described fifth embodiment of the present invention in which a discontinuous portion is present in the heat transfer fins 21, 22, the following designations can be used: λc: thermal conductivity of the concrete shielding body 3 (W/m·K); Sc: surface area of the heat transfer path of the concrete shielding body 3 in the region where the heat transfer fins 21, 22 overlap (referred to hereinbelow as “overlap portion”) (m2); Tif: temperature of the heat transfer fins 22 at the side of the inner shell 7 in the overlap portion (K); Tof: temperature of the heat transfer fins 22 at the side of the outer shell 4 in the overlap portion (K); (a) distance between the heat transfer fins 21, 22 in the overlap portion (m), and λ=λc, S=Sc, ΔT=Tif−Tof, L=a can be substituted into the aforesaid equation [A]. As a result, the heat transfer quantity QI between the heat transfer fins of two types can be obtained in the following form:QI=λc×Sc(Tif−Tof)/a [C] Further, as a comparative reference example corresponding to the above-described configuration, a structure will be considered in which the inner and outer shells 7, 4 are directly connected by heat transfer fins 30 (structure shown in FIG. 11 disclosed in Japanese Patent Application Laid-open No. 2001-3342994). In this case, the following designations can be used: λf: thermal conductivity of the heat transfer fins 30 (W/m·K); Sf: surface area of the heat transfer fins 30 (m2); Tis: temperature of the inner shell 7 (K); Tos: temperature of the outer shell 4 (K); Lc: thickness of the concrete shielding body 3 (m), and λ=λf, S=Sf, ΔT=Tis−Tos, L=Lc can be substitute into the aforesaid equation [A]. The heat transfer quantity QP between the inner and outer shells in this structure can be obtained in the following form:QP=λf×Sf(Tis−Tos)/Lc [B] Here, the heat transfer capacity (QI) of the concrete area in the structure of the fifth embodiment is inevitably somewhat inferior to the heat transfer capacity (QP) in the structure in which the inner and outer shells 7, 4 were directly connected by the heat transfer fins 30. However, if the number of the heat transfer fins 21, 22 is increased to compensate for this deficiency, then the heat transfer capacity (heat removal capability) necessary for the concrete cask A can be ensured. However, because the arrangement space of heat transfer fins 21, 22 is also limited, limitations are also placed on the possibility of such compensation. Therefore, the heat transfer quantity QI of the concrete area of this embodiment can be assumed to be limited to ½ of the heat transfer quantity QP obtained in the case in which the inner and outer shells 7, 4 are directly connected to the heat transfer fins 30. Accordingly, if the conditionQP×0.5≦QI [D]is satisfied, it will apparently be possible to obtain a concrete cask 4 in which the required heat transfer capacity can be actually attained, while effectively avoiding the radiation streaming as described hereinabove. Based on those results, the following formula(λf×Sf×(Tis−Tos)/Lc)×0.5≦λc×Sc×(Tif−Tof)/a [E]can be obtained by substituting formulas [B] and [C] into formula [D]. Here, when the heat transfer fins 30 are installed uniformly in the axial direction of the container 3, as in the comparative reference example shown in FIG. 11, the following equation is valid:Sf=t×M [F] Here, M stands for a length of the heat transfer fins 30 in the axial direction of the container 3. Further, in the fifth embodiment, when the heat transfer fins 21, 22 uniformly overlap in the axial direction of the container 3 (the case in which the lateral section of FIG. 10 appears uniform regardless of the position in the axial direction in which the container was cut), the following equation is valid:Sc=w×M [G] Here, w stands for a length of the overlap region of the first and second heat transfer fins 21, 22. Furthermore, when the heat conductivity of the heat transfer fins (21, 22, 30) is sufficiently large by comparison with that of the concrete shielding body 3, the following approximation is possible:Tis−Tos≈Tif−Tof [H] Therefore, substituting formulas [F]–[H] makes it possible to simplify the formula [E] as the formula [I] presented below:(λf×t)/Lc×0.5≦(λc×w)/a [I] The formula of claim 3 can be obtained from the formula [I]. The aforesaid formula [I] demonstrates that the heat transfer capacity (QI) in the concrete heat transfer region of the overlap portion in the fifth embodiment may be not less than the heat transfer capacity (QP) of the configuration of the comparative reference example, that is, the configuration in which the inner and outer shells 7, 4 were directly connected by the heat transfer fins 30, multiplied by 0.5 (QP×0.5≦QI). However, from the standpoint of the production cost and the number of operations, it is better to avoid the increase in the number of installed heat transfer fins 21, 22 even in the fifth embodiment. Furthermore, it is even more preferred that the heat transfer capacity QI be equal to or more than the heat transfer capacity QP obtained when the inner and outer shells 7, 4 are connected by the heat transfer fins 30 (QP≦QI). If the above-described formulas [F]–[H] are substituted into this formula, then formula [J] given below can be derived.(λf×t)/Lc≦(λc×w)/a [J]By equating the lefthand side and the right-hand side of the above mathematical expression [J], the relation of “w” (overlapping amount of heat transfer fins in radial direction) and “a” (separation amount at the overlapping portion) can be obtained in the desired case where the heat transfer capacity Qi and the heat transfer capacity Qp become equal to each other.Hereinafter, example values as practical example to be substituted into the mathematical expression are:λf=392W/(m·K) (In case of Cupper Fin)λc=1.37W/(m·K) (In case of Concrete Material)Lc=0.855mt=0.006m Plug all the above values into the mathematical expression, then we get the following relation between “w” and “a”.w=2.0a (J-1)From the obtained relation in the above [J-1], it can be observed that the overlapping amount “w” needs to be set twice as much as the separation distance “a” in order to have a heat transfer capacity QI substantially the same as the heat transfer capacity Qp. Accordingly, from the following list, it is desirable to pick one or several value combination such that the flow of raw concrete during the filling of the space between the inner shell and the outer shell with concrete is not blocked. w (mm)a (mm) 2010 4020 6030 80401005012060141701618018190201100 Note that the above values such as Lc and t are merely for the examples and the suitable values are to be determined for an individual situation. The heat transfer capacity (heat removal capacity) of the concrete cask A obtained when the L-shaped heat transfer fins 18 were mounted as in the third embodiment will be described below. FIG. 12 is a partially expanded lateral sectional view of the container of the third embodiment. Similarly to the approach followed with respect to formula [D] above, the heat transfer capacity (QI1) obtained when the heat transfer fins 18 are disposed on the side of the outer shell 4, as in the third embodiment, because of the formula QP×0.5≦QI1, the following condition should be satisfied:(λf×Sf(Tis−Tos)/Lc×0.5≦λc×Sc×(Tis−Tof)/a [K]Here,Sc: surface area of the heat transfer path of the concrete in the region between the bent portion at the distal end of the heat transfer fin 18 and the inner shell 7 (m2);Tof: temperature of the region (the aforesaid bent portion) of the heat transfer fin 18 that faces the inner shell 7 (K);a: distance between the region (the aforesaid bent portion) of the heat transfer fin 18 that faces the inner shell 7 and the inner shell 7 (m). The definitions of other parameters are absolutely identical to those of the parameters in the formulas of the above-described fifth embodiment and comparative reference example. When the thermal conductivity of the heat transfer fins (18, 30) is sufficiently larger than that of the concrete shielding body, the following formula is valid:Tis−Tos≈Tis−Tof [L] Furthermore, when the heat transfer fins 18 in the third embodiment are disposed uniformly in the axial direction, the equationSc=w×M [M]is valid. Here, w stands for a length of the bent portion (portion facing the outer wall of the inner shell 7) of the heat transfer fin 18. Thus, w means the widthwise length of the opposite surface. Therefore, the aforesaid formula [K] can be simplified as follows:((λf×t)/Lc)×0.5≦(λc×w)/a [N] The formula of claim 5 can be obtained from this formula [N]. Similarly to the approach followed with respect to formula [J] above, based on the formula QP≦QI1, it is preferred that the following formula be satisfied, which will allow the number of heat transfer fins 18 to be decreased:(λf×t)/Lc≦(λc×w)/a [O] The heat transfer capacity (heat removal capacity) of the concrete cask obtained when the L-shaped heat transfer fins 18′ were mounted on the side of the inner shell 7, as in the fourth embodiment, will be described below. FIG. 13 is a partially expanded lateral sectional view of the container of the fourth embodiment. Similarly to the approach followed with respect to formula [D] above, the heat transfer capacity (QI2) obtained when the heat transfer fins 18 are disposed on the side of the inner shell 7, as in the fourth embodiment (FIG. 13), because of the formula QP×0.5≦QI2, the following condition should be satisfied:(λf×Sf(Tis−Tos)/Lc×0.5≦λc×Sc×(Tif−Tos)/a [P]Here,Sc: surface area of the heat transfer path of the concrete in the region between the bent portion at the distal end of the heat transfer fin 18′ and the outer shell 4 (m2);Tif: temperature of the region (the aforesaid bent portion) of the heat transfer fin 18′ that faces the outer shell 4 (K);a: distance between the region (the aforesaid bent portion) of the heat transfer fin 18′ that faces the outer shell 4 and the outer shell 4 (m). The definitions of other parameters are absolutely identical to those of the parameters in the formulas of the above-described fifth embodiment and comparative reference example. When the thermal conductivity of the heat transfer fins (18′, 30) is sufficiently larger than that of the concrete shielding body, the following formula is valid:Tis−Tos≈Tif−Tos [Q] Furthermore, when the heat transfer fins 18′ in the fourth embodiment are disposed uniformly in the axial direction, the equationSc=w×M [R]is valid. Here, w stands for a length of the bent portion (portion facing the inner wall of the outer shell 4) of the heat transfer fin 18′. Thus, w means the widthwise length of the opposite surface. Therefore, the aforesaid formula [K] can be simplified as follows:((λf×t)/Lc)×0.5≦(λc×w)/a [S] The formula [S] is identical to the formula [N] and can be also used to obtain the formula of claim 5. Similarly to the approach followed with respect to formula [J] above, based on the formula QP≦QI2, it is preferred that the following formula be satisfied, which will allow the number of heat transfer fins 18′ to be decreased:(λf×t)/Lc≦(λc×w)/a [T] The heat transfer capacity (heat removal capacity) of the concrete cask having no heat transfer fins will be explained below. FIG. 14 is a partially expanded lateral sectional view of the container with a configuration containing no heat transfer fins. An assumption will be made that in the structure shown in FIG. 14 heat transfer fins 31 are present in the radial direction between the inner and outer shells 7, 4, and the width of the region of the concrete shielding body 3 of one-pitch spacing sandwiching the heat transfer fin 31 will be denoted by w. Further, the following designations will be used: Lc: thickness of the concrete shielding body 3 (m); a: length of the virtual heat transfer fin 31 in the radial direction (m); λc: thermal conductivity of the concrete shielding body 3 (W/m·K); λf: thermal conductivity of the virtual heat transfer fin 31 (W/m·K); t: thickness of the virtual heat transfer fin 31 (m); w: width of the region of the concrete shielding body 3 of one-pitch spacing sandwiching the heat transfer fin 31 (m). In this case, as a singular example of the above-described formulas [N] and [S], the following equation is valid:Lc=a [U]Therefore, the following formula is valid:λf×t≦λc×w [V] This formula [V] means that if a concrete is used that has thermal conductivity satisfying the relation described by the aforesaid formulas, then a concrete cask with a sufficient heat removal capacity can be designed (even if the heat transfer fins that have been considered indispensable in the past are absent). The thermal conductivity of a concrete shielding material enabling the heat removal design without heat transfer fins will be found hereinbelow by assuming a specific design structure of the concrete cask. The size, caloric value, and temperature difference between the inner and outer shells in the cask for which the heat removal capacity is to be established are substituted into the aforesaid formula [A] (Q=λ×s×ΔT/L). Those values were obtained by preliminary testing. More specifically, those values are: Internal caloric value: Q=14 kW. Difference in temperature between the inner shell 7 and the outer shell 4: ΔT=50K. Thickness of the shielding body: L=Lc=0.35 m. Inner diameter of the inner shell 7: D=1.6 m. Length of the heat-generating region in the axial direction: M=3.7 m. As for the heat transfer path surface area S, the virtual cylinder obtained by dividing the shielding body 3 into two equal sections in the radial direction is considered and the surface area of the circumference thereof is considered as a mean heat transfer path surface area. Furthermore, to simplify the calculations, the thickness of the inner and outer shells 7, 4 is ignored, and the diameter of the virtual cylinder is considered to be D+Lc. Therefore, the following equation is validS=π(D+Lc)×M=π×(1.6+0.35)×3.7=23 (m2). If those numerical values are substituted into the equation (A), then λ=14000/23/50×0.35=4.3 (W/m·K). Thus, this calculation example shows that if a concrete shielding body with a thermal conductivity of at least about 4 W/m·k is prepared, then the heat removal capacity identical to that of the concrete cask of the conventional type having heat transfer fins can be demonstrated even without the heat transfer fins. A concrete material with the above-described excellent thermal conduction characteristic can be obtained by admixing copper or copper alloys having excellent thermal conduction characteristic in the form of a powder, fibers, lumps, and the like. Furthermore, from the standpoint of increasing the density (effective for gamma radiation shielding), in addition to improving the thermal conduction characteristic of this concrete material, the addition of a metal material or compounds comprising iron, copper, tungsten, and the like is also effective. Using copper or copper alloys for the above-described heat transfer fins (11, 11′, 18, 18′, 21, 22) is most preferred because of their excellent thermal conduction capacity and high corrosion resistance in the alkali environment of concrete. However, when the caloric value of the radioactive substance, x, introduced into the canister (a) is comparatively small, it is not necessary to use copper or copper alloys, and ferrous materials may be used. Examples of materials with an excellent heat transfer capacity also include aluminum and aluminum alloys, but because they are dissolved in alkali environment, they can hardly be used by mixing with concrete. However, if the surface thereof is plated or subjected to anodization, they still can be used as heat transfer fins for the concrete cask. Because the concrete cask A with the present structure does not allow for the ventilation of the canister (a) (the structure such as disclosed in Japanese Patent Application laid-open No. 2001-141891), it is highly probable that the concrete material will be exposed to a high temperature of 100° C. or higher. In such an atmosphere, the free water contained in the concrete material will be released. As a result, the content ratio of hydrogen (effective for neutron shielding) can be decreased and the neutron shielding capacity can be degraded. To prevent those effects, the necessary hydrogen content in the concrete material used for this concrete cask A can be maintained by admixing hydroxides retaining water (hydrogen) in the form of crystals, rather than retaining hydrogen in form of free water. In this case, even if the concrete temperature exceeds 100° C., the content of hydrogen necessary for neutron shielding will be present and the neutron shielding capacity of the concrete will be maintained as long as the decomposition temperature (temperature at which the dissociation pressure becomes 1 atm) and melting temperature of the hydroxides are not reached. It is preferred that the hydroxides be contained at a ratio of 15 mass % or more, based on the concrete material. Examples of hydroxides with a melting point and decomposition temperature higher than 100° C., that is, hydroxides in which water is not decomposed at a temperature of 100° C., include hydroxides of alkaline earth metals such as Ca, Sr, Ba, Ra and hydroxides of metals analogous thereto, e.g. Mg. Such hydroxides hold water (hydrogen) as water of crystallization when mixed with the cured product and have excellent neutron shielding capacity. For example, because the decomposition temperature of calcium hydroxide is 580° C. and the melting point of barium hydroxide is 325° C. and the decomposition temperature thereof is 998° C., those compounds retain water (hydrogen) up to a high-temperature range. Examples of other hydroxides that can be mixed with the composition or cured product include lithium hydroxide, sodium hydroxide, potassium hydroxide, lanthanum hydroxide, chromium hydroxide, manganese hydroxide, iron hydroxide, cobalt hydroxide, nickel hydroxide, copper hydroxide, zinc hydroxide, aluminum hydroxide, lead hydroxide, gold hydroxide, platinum hydroxide, and ammonium hydroxide. Furthermore, it is preferred that the hydroxide be insoluble or poorly soluble in water. Adding such hydroxides makes it possible to introduce reliably the hydroxides that do not release water by decomposing at a temperature of more than 100° C. in the cured product after hydration reaction with cement. The hydroxides for mixing with the concrete composition have a dissolution quantity in 100 g of pure water at 20° C. of 15 g or less, more preferably of 5 g or less, most preferably 1 g or less. In terms of solubility, too, the above-mentioned hydroxides of alkaline earth metal or Mg which is a metal analogous thereto are preferred. For example, the aforesaid dissolution quantity of hydroxides of calcium, strontium and magnesium is 1 g or less, and the dissolution quantity of barium hydroxide is 5 g or less. Among those hydroxides, the hydroxides of calcium and magnesium are especially effective for increasing the neutron shielding capacity because the ratio of hydrogen contained in these hydroxides is high due to a low atomic weight of Ca and Mg. Furthermore, because calcium contained in calcium hydroxide is the main component of Portland cement and because calcium hydroxide is a substance formed by a hydration reaction in usual cements, the calcium hydroxide is most preferred among the above-mentioned hydroxides. As described hereinabove, hydroxides are introduced into the present concrete material, thereby ensuring the necessary content of hydrogen. However, because hydroxides are sometimes decomposed by reacting with carbon dioxide present in the atmosphere and release water, they have to be shielded from the atmosphere. For example, in the case of calcium hydroxide, if it reacts with carbon dioxide present in the atmosphere, it eventually becomes calcium carbonate and water (hydrogen) can be released from the crystals, causing long-term degradation of neutron shielding capacity. This reaction is represented by the following chemical formula:Ca(OH)2+CO2→CaCO3+H2O To prevent this effect, in the present embodiment, the concrete material is provided in a space shielded by the inner shell 7, outer shell 4, flanges, and a bottom plate composed from a carbon steel, stainless steel and the like, as a concrete cask structure. The term “containment” as mentioned hereinabove means that outside air comprising carbon dioxide has no contact with the concrete cured body (the aforesaid concrete shielding body 3), and the “containment” in the aforesaid sense is not lost even if a safety relief valve is provided, for example in the outer shell 4, this valve serving to release gases generated during use of the concrete cask A to the outside. Moreover, the “containment” in the aforesaid sense may be substantially attained with a structure in which contact of the concrete cured body with carbon dioxide is prevented by adsorbing carbon dioxide with an adsorbent or the like. Degassing of concrete during the manufacture of the concrete cask A will be explained below. Thus, there is a high probability that the air will penetrate into the concrete and pores will be formed therein when the concrete is mixed and placed. When the container 3 is composed of such a concrete, the pores present therein become the loss areas of the shielding body, which is undesirable from the standpoint of preventing the streaming of radiation. Therefore, a method for vacuum degassing during mixing or placing may be used. FIG. 15 illustrates an example of the configuration for vacuum degassing during concrete mixing, and FIG. 16 illustrates an example of the configuration for vacuum degassing during concrete placing. Vacuum degassing during mixing can be conducted by employing a containment (sealed) structure of the mixing chamber of a mixing machine such as a pot mixer, a screw mixer, or a puddle mixer, and disposing a vacuum pump therein. An example of the configuration for vacuum degassing during concrete mixing is shown in FIG. 15. In FIG. 15, the reference numeral 61 stands for a pot-type concrete mixer with a mixing chamber constructed inside the pot. The pot is equipped with a disk-like vacuum flange 62 detachably provided in the opening 61a of the pot. The vacuum flange 62 has an appropriate containment structure and can air-tightly cover the opening 61a. As a result, the inside of the pot is sealed. An air suction opening (not shown in the figures) is formed on one side surface of the vacuum flange 62, and when the vacuum flange 62 is mounted on the concrete mixer 61, this air suction opening is connected to the space inside the pot. A boss portion is provided in a protruding condition in the center of the surface on the other side of the vacuum flange 62, and a linking hole 63 is formed in the boss portion. The linking hole 63 is connected to the aforesaid air suction opening via an appropriate path formed in the space inside the vacuum flange 62. One end of the flexible hose 65 is attached to the linking hole 63. In order to prevent the flexible hose from twisting, a rotary joint 64 is introduced into a place of connection to the linking hole 63. The other end of the flexible hose 65 is connected to the suction side of the vacuum pump 66. In the above-described structure, air bubbles are introduced into the concrete by mixing inside the pot, but the air bubbles can be sucked out and removed via the flexible hose 65 and the concrete can be degassed by degassing the inside of the mixing chamber by driving the vacuum pump 66 in parallel with the mixing operation. FIG. 16 illustrates a structure for vacuum degassing during concrete placing. In the structure shown in FIG. 16, a sealable lid 68 is disposed above the inner and outer shells 7, 4. In the lid, concrete placing holes 69 are provided in several zones and a suction opening 70 is formed. The suction opening 70 is connected via an appropriate hose 71 to a vacuum pump 72. A pipe denoted by the reference numeral 73 serves for feeding the concrete. When concrete is placed in this structure, fresh concrete is poured from the placing holes 69 into the space between the inner and outer shells 7, 4, and the vacuum pump 72 is driven to degas the space between the inner and outer shells 7, 4. As a result, the concrete is degassed. In the structure of the embodiments of the present invention, because the inner and outer shells 7, 4 are not entirely partitioned by the heat transfer fins (11, etc.), the fresh concrete can flow from one cell to another. As a result, the number of zones for disposing the concrete placing holes 69 can be reduced, as shown in FIG. 16. Further, the above-described easiness of concrete placement can be similarly improved even in the structure in which the heat transfer fins 180, each is formed with a cutout portion 180C, that is, cut only partially as shown in FIG. 18A, rather than completely, in the axial direction of the container 3 in the separation space like the one 181A shown in FIG. 18B. Needless to say the cutout similar to the one 180C can also be formed on the inner-side end of the heat transfer fin 180. Moreover, if through holes (openings 181C) are provided in addition to the aforesaid separation portion 181A in the heat transfer fins 181 as shown in FIG. 18B, then the concrete can be also caused to flow through those through holes 181C, thereby also increasing the easiness of placing. The shape, number and location of the openings may be appropriately set in balance with the above-described heat transfer capacity. For example, in the case of the zigzag arrangement of heat transfer fins 21, 22 as in the fifth embodiment as shown in FIG. 6, it is preferred that the openings, 182C1, 182C2, be provided in the regions aside of the overlap portions of the both the heat transfer fins 182A, 182B, in order to minimize the decrease in heat transfer capacity. Yet moreover, it may be possible to provide a heat transfer fin 183, as shown in FIG. 18D, having both of radial ends fixed to the outer shell 4 and the inner shell 7, respectively, and on which it is formed with a plurality of openings 183C1, 183C2 (not limited to the plural opening configuration but a single opening can be used). As described for the embodiments shown in FIGS. 18A, 18B, and 18C, the shape, number and location of the openings may be appropriately set in balance with the above-mentioned heat transfer capacity. Furthermore, any feasible combination of the openings shown in FIGS. 18A to 18D, can be made without departing the essential concept of the present invention. The verification test of heat transfer performance of the concrete cask will be described below. FIG. 17A is a longitudinal sectional view of a sample in the heat transfer capacity verification test of the concrete cask of the fifth embodiment. FIG. 17B is a lateral sectional view. A heat transfer sample C used in the verification test is shown in FIG. 17. The heat transfer sample C is equivalent to the structure in which a tubular portion of the container body 1 of the concrete cask of the fifth embodiment is cut out and comprises the aforesaid inner and outer shells 7, 4 and the concrete shielding body 3. As shown in FIG. 17A, both end surfaces in the axial direction of the heat transfer sample C are covered with thermally insulating materials 80, 80. A thermally insulating material 81 is also disposed inside the inner shell 7. A cylindrical gap of an appropriate thickness if formed between the thermally insulating material 81 and the inner shell 7, and a heater 82 for heating is disposed in this gap portion. The thermally insulating material 81 and heater 82 are not shown in FIG. 17B. In the structure shown in FIG. 17, a heat transfer test was carried out with a heater output of 2.1 kW. The heat transfer analysis was also conducted under the identical conditions and the analysis results were compared with the results of the heat transfer test. Here, (w) was 90 mm and (a) was 38 mm. The mixing composition of the concrete material used for the heat transfer test is shown in Table 1. The materials used for the sample are shown in Table 2. TABLE 1mixing composition of the concrete material used for the heat transfer testUnit Amount (Kg/m3)Chemical Admixturehighlow heatMetalperformancePortlandsilicacalciumironironAE waterdeformingcementfumehydroxidepowderfiberreducing agentagentwater287321131640157940.9281 TABLE 2Materials Used for the TestHeatConductivityParts NameMaterialThickness (mm)(W/m · K)Inner shellcarbon steel1652Outer shellcarbon steel1652Heat Transcupper2398FinShieldingconcrete2502.0body Calculating (λf×t)/Lc and (λc×w)/a from those dimensions and physical property values, yields the following:(λf×t)/Lc=3.1(W/m·K)(λc×w)/a=3.3(W/m·K). It is clear, that the aforesaid formula [T], that is,(λf×t)/Lc≦(λc×w)/a, is satisfied. The results of the heat transfer test and heat transfer analysis are shown in Table 3. TABLE 3results of the heat transfer test and heat transferanalysis (Unit: degree in Celsius)Temp of InnerTemp of OutershellshellTest results8868Result by Heat8867Transfer Analysis The results matched well and the difference in temperature between the inner shell and outer shell was about 20° C. in both the heat transfer test and the heat transfer analysis. On the other hand, the difference in temperature between the inner shell and outer shell that was calculated for the conventional structure in which the inner and outer shells were connected by heat transfer fins by using the present test model was about 20° C. and was confirmed to be equal to that of the heat transfer test results and heat analysis results obtained for the concrete cask of the present invention. The above results proved that the concrete cask in accordance with the present invention has sufficient heat transfer capacity (heat removal capacity). Eight embodiments of the present invention are described above, but the present invention is not limited to the configurations of the above-described embodiments, and a variety of modifications can be made without departing from the essence of the present invention. For example, in the first embodiment, the explanation was conducted with respect to a concrete cask for accommodating a radioactive substance contained in a canister in an accommodation unit. However, the present invention is also applicable to a concrete cask accommodating a radioactive substance contained in a basket. Furthermore, in the above-described embodiments, the heat transfer fins (11, etc.) were installed radially along the axial direction of the container 3. However, a configuration may be also employed in which the heat transfer fins are formed to have a fan-like shape perpendicular to the axial direction of the container and are mounted with equal spacing in the axial direction, alternately on the inner and outer shells 7, 4, while ensuring the overlap region necessary for thermal conduction (modification example of the aforesaid fifth embodiment). Further, when a structure is used with the heat transfer fins having a fan-like shape, if air bubbles are introduced into the concrete during placing, the problem is that they hang on the heat transfer fins and are difficult to remove. In order to resolve this problem associated with degassing, the heat transfer fins may be inclined so that the edge portions thereof be higher than the mounting position or the heat transfer fins may be inclined spirally. The present invention has the above-described configuration and therefore produces the following effects. In summary, the present invention relates to a concrete cask, in which a shielding body composed of concrete and heat transfer fins made from metal are provided between an inner shell and an outer shell made from metal and which comprises an accommodation portion formed inside the inner shell for accommodating a radioactive substance, a containment structure is employed to shield the accommodation portion from the outside of the cask, and said heat transfer fins each has an inner shell-side and an outer shell-side and is configured such that said inner shell-side is in contact with the inner shell and the outer shell-side is formed with at least a portion that is not in contact with the outer shell or such that said outer shell-side is in contact with the outer shell and the inner shell-side is formed with at least a portion that is not in contact with the inner shell. Therefore, in the conventional structure in which the heat transfer fins were connected to both the inner shell and the outer shell, it was necessary to place the concrete in each cell individually, whereas in accordance with the present invention such a configuration is not necessary and the manufacture is facilitated. Furthermore, in the conventional structure, because the heat transfer fins could create a region in which the shielding body was not present over the entire range in the radial direction, there was a problem associated with radiation streaming. However, in accordance with the present invention, even if the radiation passes through the heat transfer fins, it also has to pass through the shielding body before it can reach the outer shell. Therefore, the radiation streaming can be suppressed. In the above described cask, the concrete cask may comprise at least first heat transfer fins provided in contact with the outer shell-side and second heat transfer fins provided in contact with the inner shell-side, the first heat transfer fins and second heat transfer fins being provided so as to overlap each other and so that there is a distance between both the heat transfer fins in the overlap portion. The advantage of this configuration is that, in addition to the effect identical to that of claim 1, because the overlap portion is present, thermally conductive properties can be sufficiently ensured by the discontinuous region of heat transfer fins. Furthermore, if the length of the overlap portion of the both the heat transfer fins is denoted by w1 and the distance between the both the heat transfer fins in the overlap portion is denoted by a1, the following relationship is preferably satisfied: a1≦(2·λc·w1·Lc)/(λfλt). Therefore, heat transfer capacity equal to or better than that obtained when the heat transfer fins connect the outer and inner shells, as in the conventional configuration, can be obtained. Moreover, the side of the heat transfer fins that forms the separation portion can be formed to have an almost L-like shape so as to be provided with an opposite surface facing the inner shell or the outer shell. Therefore, heat transfer to the side opposite to that where the heat transfer fins are mounted can be enhanced. Furthermore, because the heat transfer fins are secured only to one shell of the inner shell and outer shell, the mounting time is shortened. Furthermore, if the separation distance of the separation portion is denoted by a2, the following relationship is satisfied: a2≦(2·λc·w2·Lc)/(λf·t). Therefore, heat transfer capacity equal to or better than that obtained when the heat transfer fins connect the outer and inner shells, as in the conventional configuration, can be obtained. As an alternate example, the heat transfer fins can be formed to have an almost I-like shape. Therefore, the manufacture of the heat transfer fins is facilitated and the production cost and the number of operations can be reduced. In one example, the separation portion can be composed so as to separate completely the heat transfer fins and the inner shell or outer shell. Therefore, because the heat transfer fins are mounted only on the outer shell or inner shell, the time required for mounting the heat transfer fins can be saved. Furthermore, because the inner shell and outer shell are not connected, the inner shell and outer shell can be manufactured independently. Therefore, the manufacturing process can be shortened. In another example, the heat transfer fins are disposed at an angle to the radial direction of the shielding body. Therefore, the radiation streaming can be avoided more reliably. Furthermore, openings can be formed in the heat transfer fins. Therefore, concrete can easily flow through the opening and concrete placing is facilitated. In another form of the embodiment of the present invention, a concrete cask comprising a shielding body composed of concrete and provided between the inner shell and the outer shell made from metal and an accommodation portion for accommodating a radioactive substance inside the inner shell, wherein a containment structure is employed to shield the accommodation portion from the outside of the cask, and the shielding body is composed of concrete that has good thermal conductivity comprising a metal material. Therefore, introducing a metal material increases thermal conduction capacity and makes it possible to provide a cut portion between the heat exchange fins and the inner shell or outer shell, thereby suppressing radiation streaming. Furthermore, the concrete density is increased and gamma radiation shielding capacity is increased. In the aforementioned embodiments, the thermal conductivity of the shielding body is preferably 4 (W/m·K) or more. Therefore sufficient thermal conduction capacity can be obtained. In particular, because a sufficient heat removal capacity can be attained even when no heat transfer fins are present, the heat transfer fins can be omitted and the structure of the concrete cask can be simplified. In the aforementioned embodiments, the shielding body comprises a metal material in at least one shape of grains, particles, or fibers. Therefore, thermal conduction capacity can be improved. Moreover, the shielding body preferably contains 15 mass % or more of hydroxide retaining water as crystals with a melting point and decomposition temperature higher than 100° C. Therefore, the shielding body has excellent neutron shielding capacity, in particular in a high-temperature environment with a temperature of 100° C. and higher. Yet moreover, the hydroxide shows poor solubility or insolubility in water. Therefore, the hydroxide that neither decomposes nor releases water at a temperature of 100° C. and higher can be reliably introduced into the cured body obtained after hydration with the cement. Furthermore, the shielding body is preferably sealed so as to be shielded from outside air. Therefore, the concrete material is prevented from reacting with carbon dioxide present in the atmosphere and releasing hydrogen present therein and the degradation of neutron shielding capacity can be prevented. The invention also related to a method for manufacturing the concrete cask, the method comprises a mixing step for mixing a shielding body material that forms the shielding body and a placing step for placing the mixed shielding body materials, wherein the shielding body material is vacuum degassed in at least one of those processes. Therefore, pores present in the concrete shielding body can be eliminated and a concrete cask with excellent shielding capacity can be obtained. In the mixing step, the shielding body material is vacuum degassed by mixing the shielding body material in a mixing chamber of a mixing machine and degassing the inside of the mixing chamber with a vacuum pump. Therefore, the introduction of air during mixing is prevented. As a result, pores present in the concrete shielding body can be eliminated and a concrete cask with excellent shielding capacity can be obtained. In the placing step, the shielding body material is vacuum degassed by placing the shielding body material mixed in the mixing step into a space formed between the inner shell and the outer shell and degassing the space with a vacuum pump. Therefore, the introduction of air during placing is prevented. As a result, pores present in the concrete shielding body can be eliminated and a concrete cask with excellent shielding capacity can be obtained. This application is based on Japanese patent application serial no. 2003-24208, filed in Japan Patent Office on Jan. 31, 2003, the contents of which are hereby incorporated by reference. Although the present invention has been fully described by way of example with reference to the accompanying drawings, it is to be understood that various changes and modifications will be apparent to those skilled in the art. Therefore, unless otherwise such changes and modifications depart from the scope of the present invention hereinafter defined, they should be construed as being included therein. |
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abstract | A method and system is disclosed for acquiring image data of a subject. The image data can be collected with an imaging system using various selection techniques. The selection techniques may be used to assist in generating selected images for viewing. Selection techniques may include moving a filter to filter a selected portion of an imaging beam. |
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claims | 1. A computer system for testing derating performance of a component of an electronic device, the computer system comprising:a storage operable to store component information of the component;at least one processor; andone or more programs stored in the storage and being executable by the at least one processor, the one or more programs comprising:a data obtaining module operable to obtain a component list, a pin list, and a standard derating list of the electronic device from the storage;a data classification module operable to classify components in the component list into different categories;a parameter recording module operable to receive parameters of each component as measured by a user in each category, the parameters of each component comprising voltages of two pins of the component and a working temperature of the component;a derating test module operable to obtain component information of a component from the storage, calculate a working voltage of the component according to the voltages of two pins of the component, calculate a derating ratio of the component according to the calculated working voltage and a rated voltage of the component in the component information, compare the derating ratio with a standard derating ratio range of the component in the standard derating list to get a first analysis result of the derating ratio, compare the working temperature of the component with a standard temperature range of the component in the standard derating list to get a second analysis result of the working temperature; anda report generation module operable to generate a test report comprising the derating ratio, the working temperature, the first and second analysis results of each component in the component list, and store the test report in the storage. 2. The computer system according to claim 1, wherein the working voltage of the component is an absolute value of a difference between the voltages of the two pins of the component. 3. The computer system according to claim 1, wherein the derating ratio of the component is a ratio of the calculated working voltage of the component and the rated voltage of the component. 4. The computer system according to claim 1, wherein the derating test module determines that the derating ratio is valid if the derating ratio is within the standard derating ratio range in the standard derating list, or determines that the derating ratio is invalid if the derating ratio is not within the standard derating ratio range in the standard derating list. 5. The computer system according to claim 1, wherein the derating test module determines that the working temperature is valid if the working temperature is within the standard working temperature range in the standard derating list, or determines that the working temperature is invalid if the working temperature is not within the standard working temperature range in the standard derating list. 6. A computer-implemented method being executed by a processor of a computer for testing derating performance of a component of an electronic device, the method comprising:obtaining a component list, a pin list, and a standard derating list of the electronic device from a storage;classifying components in the component list into different categories;receiving parameters of each component as measured by a user in each category, the parameters of each component comprising voltages of two pins of the component and a working temperature of the component;obtaining component information of a component from the storage;calculating a working voltage of the component according to the voltages of two pins of the component;calculating a derating ratio of the component on the processor according to the calculated working voltage and a rated voltage of the component in the component information;comparing the derating ratio with a standard derating ratio range of the component in the standard derating list to get a first analysis result of the derating ratio;comparing the working temperature of the component with a standard temperature range of the component in the standard derating list to get a second analysis result of the working temperature; andgenerating a test report comprising the derating ratio, the working temperature, the first and second analysis results of each component in the component list, and storing the test report in the storage. 7. The method according to claim 6, wherein the working voltage of the component is an absolute value of a difference between the voltages of the two pins of the component. 8. The method according to claim 6, wherein the derating ratio of the component is a ratio of the calculated working voltage of the component and the rated voltage of the component. 9. The method according to claim 6, wherein the derating ratio is valid if the derating ratio is within the standard derating ratio range in the standard derating list, or the derating ratio is invalid if the derating ratio is not within the standard derating ratio range in the standard derating list. 10. The method according to claim 6, wherein the working temperature is valid if the working temperature is within the standard working temperature range in the standard derating list, or the working temperature is invalid if the working temperature is not within the standard working temperature range in the standard derating list. 11. A non-transitory computer readable medium containing computer instructions stored therein for causing a computer processor to perform a method for testing derating performance of a component of an electronic device, the method comprising:obtaining a component list, a pin list, and a standard derating list of the electronic device from a storage;classifying components in the component list into different categories;receiving parameters of each component as measured by a user in each category, the parameters of each component comprising voltages of two pins of the component and a working temperature of the component;obtaining component information of a component from the storage;calculating a working voltage of the component according to the voltages of two pins of the component;calculating a derating ratio of the component according to the calculated working voltage and a rated voltage of the component in the component information;comparing the derating ratio with a standard derating ratio range of the component in the standard derating list to get a first analysis result of the derating ratio;comparing the working temperature of the component with a standard temperature range of the component in the standard derating list to get a second analysis result of the working temperature; andgenerating a test report comprising the derating ratio, the working temperature, the first and second analysis results of each component in the component list, and storing the test report in the storage. 12. The non-transitory computer readable medium according to claim 11, wherein the working voltage of the component is an absolute value of a difference between the voltages of the two pins of the component. 13. The non-transitory computer readable medium according to claim 11, wherein the derating ratio of the component is a ratio of the calculated working voltage of the component and the rated voltage of the component. 14. The non-transitory computer readable medium according to claim 11, wherein the derating ratio is valid if the derating ratio is within the standard derating ratio range in the standard derating list, or the derating ratio is invalid if the derating ratio is not within the standard derating ratio range in the standard derating list. 15. The non-transitory computer readable medium according to claim 11, wherein the working temperature is valid if the working temperature is within the standard working temperature range in the standard derating list, or the working temperature is invalid if the working temperature is not within the standard working temperature range in the standard derating list. |
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058754067 | abstract | A method for reducing radioactive waste, particularly oils and solvents in nuclear power stations and military research centers, and a device therefor. The method comprises feeding the waste into a tank (1) in which it is continuously stirred, preheating the waste, performing a chemical precipitation treatment, feeding the mixture into a centrifuge (2), performing an electrostatic or conventional filtration step, and testing the level of radioactivity. The treatment is continuously repeated until the desired level of decontamination is reached. |
description | This application claims the benefit of U.S. Provisional Application No. 60/704,411, filed Aug. 1, 2005, which is hereby incorporated herein by reference in its entirety. This invention was made with support from the United States Department of Defense Breast Cancer Research Program under Grant Nos. DAMD 17-02-1-0517 and DAMD 17-02-1-0518. Accordingly, the United States government may have certain rights in the invention. This application also contains subject matter which is related to the subject matter of the following applications, each of which is hereby incorporated herein by reference in its entirety: “An Optical Device For Directing X-Rays Having a Plurality of Optical Crystals,” by Zewu Chen, U.S. Ser. No. 11/048,146, filed Feb. 1, 2005, which application is a continuation of PCT Application PCT/US2003/023412, filed Jul. 25, 2003, and published under the PCT Articles in English as WO 2004/013867 A2 on Feb. 12, 2004, which PCT application claimed priority to U.S. Provisional Application No. 60/400,809, filed Aug. 2, 2002. This invention relates generally to devices and methods for focusing high-energy electromagnetic radiation. Specifically, the present invention provides improved imaging systems for directing and three-dimensional focusing x-rays to allow for low dose, high definition imaging of an object, such as a biological object. X-ray analysis techniques have been some of the most significant developments in science and technology over the previous century. The use of x-ray diffraction, x-ray spectroscopy, x-ray imaging, and other x-ray analysis techniques has lead to a profound increase in knowledge in virtually all scientific fields. Today, x-ray imaging is used in a variety of applications, including medical, scientific and industrial applications. Various ones of these applications can be extremely challenging. For example, screening mammography using x-ray imaging is a critical and challenging application, where dose, contrast, resolution and costs are all important Patient dose is reduced and image quality is increased using monochromatic beams by the removal of low energy photons that are otherwise heavily absorbed in the patient without contributing to the image, and the removal of high energy photons that give relatively low subject contrast and cause Compton scattering, which degrades image quality. One problem, however, is that synchrotrons are expensive, and not generally clinically available. Monochromatic beams can also be achieved by using single crystal diffraction from a conventional source, but such implementations do not give the desired intensity since only the small fraction of the incident beam at the right energy and the right angle is diffracted. Thus, there exists a need in the art for x-ray imaging system enhancements to, for example, more beneficially balance dose, contrast, resolution and costs considerations than currently available x-ray imaging systems. The shortcomings of the prior art are overcome and additional advantages are provided through the provision of an x-ray imaging system which includes an x-ray source, an optical device, and a detector. The optical device, which directs x-rays from the x-ray source, includes at least one point-focusing, curved monochromating optic for directing x-rays from the x-ray source towards a focal point. The at least one point-focusing, curved monochromating optic directs a focused monochromatic x-ray beam towards the focal point, and the detector is aligned with the focused monochromatic x-ray beam directed from the optical device. The optical device facilitates x-ray imaging of an object using the detector when the object is placed between the optical device and the detector, within the focused monochromatic x-ray beam directed from the optical device. In enhanced implementations, each point-focusing, curved monochromatic optic has a doubly-curved optical surface, and the at least one point-focusing, curved monochromating optic comprises a plurality of doubly-curved optical crystals or a plurality of doubly-curved multilayer optics. The optical device facilitates passive image demagnification of an object when the object is placed before the focal point, and the detector is located closer to the focal point than the object is to the focal point. The optical device facilitates passive image magnification of an object when the detector is located further from the focal point than the object is to the focal point. Depending upon the imaging application, the object can be placed either before or after the focal point, as can the detector. In a further aspect, an imaging system is provided which includes an x-ray source, a first optical device, a second optical device, and a detector. The first optical device includes at least one first point-focusing, curved monochromating optic for directing x-rays from the x-ray source towards a first focal point in the form of a first focused monochromatic x-ray beam. The second optical device is aligned with the first focused monochromatic x-ray beam, and includes at least one second point-focusing, curved monochromating optic for directing x-rays of the first focused monochromatic x-ray beam towards a second focal point in the form of a second focused monochromatic x-ray beam. The detector is aligned with the second focused monochromatic x-ray beam. The first and second optical devices facilitate imaging of an object using the detector when the object is placed between the first optical device and the second optical device, within the first focused monochromatic x-ray beam. Further, additional features and advantages are realized through the techniques of the present invention. Other embodiments and aspects of the invention are described in detail herein and are considered a part of the claimed invention. In the following discussion and the appended claims, various aspects of the present invention are described in terms of their application to the modification of the path of x-ray radiation, but it should be understood that the present invention is applicable to the manipulation and use of other types of radiation, for example, gamma rays, electron beams and neutrons. In areas of x-ray spectroscopy, high x-ray beam intensity is an essential requirement to reduce sample exposure times and, consequently, to improve the signal-to-noise ratio of x-ray analysis measurements. In the past, expensive and powerful x-ray sources, such as high-power sealed tubes, rotating anode x-ray tubes or synchrotrons, were the only options available to produce high-intensity x-ray beams. Recently, the development of x-ray optical devices has made it possible to collect the diverging radiation from an x-ray source by focusing the x-rays. A combination of x-ray focusing optics and small, low-power x-ray sources can produce x-ray beams with intensities comparable to those achieved with more expensive devices. As a result, systems based on a combination of small x-ray sources and collection optics have greatly expanded the capabilities of x-ray analysis equipment in, for example, small laboratories, or in-situ field, clinic, process line or factory applications. One existing x-ray optical technology is based on diffraction of x-rays by crystals, for example, germanium (Ge) or silicon (Si) crystals. Curved crystals can provide deflection of diverging radiation from an x-ray source onto a target, as well as providing monochromatization of photons reaching the target. Two different types of curved crystals exist: singly-curved crystals and doubly-curved crystals (DCC). Using what is known in the art as Rowland circle geometry, singly-curved crystals provide focusing in two dimensions, leaving x-ray radiation unfocused in the third or orthogonal plane. Doubly-curved crystals provide focusing of x-rays from the source to a point target in all three dimensions, for example, as disclosed by Chen and Wittry in the article entitled “Microprobe X-ray Fluorescence with the Use of Point-focusing Diffractors,” which appeared in Applied Physics Letters, 71 (13), 1884 (1997), the disclosure of which is incorporated by reference herein. This three-dimensional focusing is referred to in the art as “point-to-point” focusing. The point-to-point focusing property of certain doubly-curved crystals has many important applications in, for example, material science structural or composition analysis. Curved crystals further divide into Johansson and Johan types. Johansson geometry requires, e.g., crystal planes to have a curvature that is equal to twice the radius of the Rowland circle, but a crystal surface grinded to the radius of the Rowland circle, while Johan geometry configuration requires, e.g., a curvature twice the radius of the Rowland circle. One advantage of providing a high-intensity x-ray beam is that the desired sample exposure can typically be achieved in a shorter measurement time. The potential to provide shorter measurement times can be critical in many applications. For example, in some applications, reduced measurement time increases the signal-to-noise ratio of the measurement. In addition, minimizing analysis time increases the sample throughput in, for example, industrial applications, thus improving productivity. Another important application is x-ray imaging, which is the application to which the present invention is directed. Presented herein are various radiation imaging systems for facilitating three-dimensional focusing of characteristic x-rays by diffraction employing optical devices, such as point focusing, monochromating curved optics. Implementation of high contrast monochromatic imaging utilizing very low power sources is demonstrated using point focusing, monochromating curved optics at, for example, patient imaging energies. The curved optic can comprise various optical devices, including one or more doubly-curved crystal (DCC) optics or one or more doubly-curved multilayer optics. One embodiment of such a doubly-curved optical device is depicted in FIGS. 1 and 1A, and is described in detail in U.S. Pat. No. 6,285,506 B1, issued Sep. 4, 2001, the entirety of which is hereby incorporated herein by reference. In the embodiment of FIG. 1, a doubly-curved optical device includes a flexible layer 110, a thick epoxy layer 112 and a backing plate 114. The structure of the device is shown further in the cross-sectional elevational view in FIG. 1A. In this device, the epoxy layer 112 holds and constrains the flexible layer 110 to a selected geometry having a curvature. Preferably, the thickness of the epoxy layer is greater than 20 μm and the thickness of the flexible layer is greater than 5 μm. Further, the thickness of the epoxy layer is typically thicker than the thickness of the flexible layer. The flexible layer can be one of a large variety of materials, including: mica, Si, Ge, quartz, plastic, glass etc. The epoxy layer 112 can be a paste type with viscosity in the order of 103 to 104 poise and 30 to 60 minutes pot life. The backing plate 114 can be a solid object that bonds well with the epoxy. The surface 118 of the backing plate can be flat (FIG. 1A) or curved, and its exact shape and surface finish are not critical to the shape and surface finish of the flexible layer. In the device of FIGS. 1 & 1A, a specially prepared backing plate is not required. Surrounding the flexible layer may be a thin sheet of protection material 116, such as a thin plastic, which is used around the flexible layer edge (see FIG. 1A). The protection material protects the fabrication mold so that the mold is reusable, and would not be necessary for a mold that is the exact size or smaller than the flexible layer, or for a sacrificial mold. Doubly-curved optical devices, such as doubly-curved crystal (DCC) optics, are now used in material analysis to collect and focus x-rays from a large solid angle and increase the usable flux from an x-ray source. As noted, three-dimensional focusing of characteristic x-rays can be achieved by diffraction from a toroidal crystal used with a small electronic bombardment x-ray source. This point-to-point Johan geometry is illustrated in FIG. 2. The diffracting planes of each crystal optic element 200 can be parallel to the crystal surface. If the focal circle 210 containing a point source and the focal point has radius R0, then the crystal surface has a radius R of curvature of 2R0 in the plane of the focal circle and a radius of curvature of r=2R0 sin2 θBrag in the perpendicular plane, with the radius centered on a line segment drawn between the source and the focal point. X-rays diverging from the source, and incident on the crystal surface at angles within the rocking curve of the crystal will be reflected efficiently to the focal or image point. The monochromatic flux density at the focal point for a DCC-based system is several orders of magnitude greater than that of conventional systems with higher power sources and similar source to object distances. This increase yields a very high sensitivity for use in many different applications, including (as described herein) radiographic imaging. As a further enhancement, FIG. 2 illustrates that the optical device may comprise multiple doubly-curved crystal optic elements 200 arranged in a grid pattern about the Rowland circle. Such a structure may be arranged to optimize the capture and redirection of divergent radiation via Bragg diffraction. In one aspect, a plurality of optic crystals having varying atomic diffraction plane orientations can be used to capture and focus divergent x-rays towards a focal point. In another aspect, a two or three dimensional matrix of crystals can be positioned relative to an x-ray source to capture and focus divergent x-rays in three dimensions. Further details of such a structure are presented in the above-incorporated, co-pending U.S. patent application Ser. No. 11/048,146, entitled “An Optical Device for Directing X-Rays Having a Plurality of Optical Crystals”. Point focusing, monochromating curved optics, such as the doubly-curved crystals (DCC) discussed above, are employed herein to point focus and monochromatically redirect x-rays from a large solid angle x-ray source for x-ray imaging. Monochromatic beams improve contrast and minimize radiation dose, which can be significant where a lower powered source is desired or where the object to be imaged is a patient. Although described in detail herein below with respect to mammography, it should be understood that the x-ray imaging system and techniques described herein are applicable to radiographic imaging in general, and not to any particular application. For example, the systems and techniques described can be employed to image any biological object, or non-biological object, such as an integrated circuit chip. Monochromatic beams are achieved herein by employing point-focusing, monochromating curved optics. One embodiment of an x-ray imaging system employing such an optic is depicted in FIG. 3. As shown, x-rays 305 from an x-ray source 300 are directed by an optical device 320 comprising at least one point-focusing, curved monochromating optic, to controllably converge at a focal point 360. The size of the x-ray source can vary with the x-ray imaging application. Typically the source would be a conventional x-ray source, such as an electron impact source including a conventional fixed anode or rotating anode x-ray tube. For certain applications however low power sources may be used, for example less than 500 Watts. The curved monochromating optic 320 directs x-rays from the source towards the focal point as a focused monochromatic x-ray beam 325. An input focal slit 310 and output focal slit 330 may be employed alone or together to further limit background radiation, limit divergence or shape the output beam. An object 340 to be imaged is placed within the focused monochromatic x-ray beam 325 after optical device 320. Object 340 can be placed either before focal point 360 as shown, or after focal point 360. With object 340 disposed before the focal point, a detector 350 can be located before the focal point or after the focal point. Detector 350 is an imaging detector that provides a two-dimensional map of x-ray intensity. This can be either a direct detector or an indirect detector coupled to a phosphor (which converts x-rays to visible light). The detector could be film, a film/screen cassette, a CCD coupled to a phosphor, an amorphous selenium or amorphous silicon detector, a computed radiography plate, a CdZnTe detector, or any other analog or digital detector. If detector 350 is located before focal point 360, then the image can be demagnified onto the detector when the detector is placed closer to the focal point, for example, at locations 370 or 380. Alternatively, the object can be magnified onto a larger detector located at location 390, i.e., with the detector disposed further from focal point 360 than object 340. Magnification is beneficial if system resolution is detector limited, while demagnification is beneficial if it is desired to use smaller, cheaper detectors. Image blur due to angular divergence can be reduced by placing the detector at location 350, near to object 340, and further reduced by increasing the distance between the object and the focal point, as shown in FIG. 4. In FIG. 4, x-rays 405 from an x-ray source 400 are again directed by an optical device 420, comprising a point-focusing, curved monochromating optic, into a focused monochromatic x-ray beam 425 to controllably converge at a focal point 460. In this embodiment, object 440 to be imaged is located after the focal point in order to be further from the focal point than possible if disposed between optical device 420 and focal point 460. Imaging detector 450 is placed near object 440. Intensity of the point-focused, monochromatic beam is tailored to a particular imaging application. Intensity depends, in part, on the collected solid angle, which can be increased by decreasing the input focal distance between the x-ray source and the optical device, or by increasing the size of the optical device or by using multiple point-focusing, curved monochromating optics arranged in a structure such as depicted in the above-incorporated U.S. patent application Ser. No. 11/048,146, entitled “An Optical Device for Directing X-Rays Having a Plurality of Optical Crystals”. The resolution of the image can be improved by decreasing the angular divergence of the point-focused monochromatic beam, which can be accomplished by increasing the object-to-focal-point distance, or increasing the output focal distance. Intensity and resolution can also be adjusted by employing an optical device with a symmetrical optic having equal input and output focal lengths, or an asymmetrical optic with differing input and output focal lengths. A doubly-curved optical device such as described herein produces a curved fan beam output, such as shown by way of example, in FIG. 5. This fan beam output can be scanned across the object in a manner similar to a conventional slot/scan system. Scanning, which facilitates covering a large area without requiring a very large object-to-optic distance, could be accomplished by (1) moving the object across the beam or (2) by moving the source and optic(s) together so that the beam crosses the object, or (3) by rotating the optic so that the input remains pointed at the source but the output scans across the object. These options are shown respectively with reference to FIGS. 5A-C. In these figures (where like numerals are used to denote like elements), the system of FIG. 3 is used as an example, but those skilled in the art will recognize that these principles can be extended to any of the imaging systems disclosed herein. FIG. 5A depicts gantries 510 and 512 to move the object 340 and large area detector 380 across the beam. The object is moved, for example, in the direction perpendicular to the beam across its narrow dimension. The detector is moved, for example, simultaneously in the same direction, at a speed equal to the magnification factor times the object speed. FIG. 5B depicts a gantry 520 supporting source 300, optic 320, and slits 310 and 330 (if any), collectively scanned to move the beam across the stationary object 340. FIG. 5C depicts a gantry 530 supporting optic 320 and slits 310 and 330 (if any), scanned to move the optic 320 along a circle centered at the source, maintaining the alignment at the Bragg angle. Depending on the constraints of the source geometry, it may be necessary to simultaneously rotate the source housing. Those skilled in the art will recognize that features of these scanning systems may be combined to produce the desired results. The detector could be a large area detector fixed with respect to the object, or a smaller detector, for example the size of the output beam, which is fixed with respect to the beam and records the changes as the object moves with respect to the beam. The beam can be limited to a straight fan with the use of slits, such as depicted in FIG. 3, or employed as single or multiple curved segments. Further, significant reduction in scatter production when imaging large objects is obtained using a scanned beam. The output focal spot from a doubly-curved optical device can be used as the source for a second optical device (as shown in FIG. 6) to effect refractive contrast (diffraction enhanced) imaging. In this embodiment, a doubly-curved optical device 620 collects radiation 605 from an x-ray source 600 and directs the radiation in a point-focused monochromatic x-ray beam 625 to converge at a first focal point 630, which as noted, is the x-ray source for a second optical device 650. Device 650, which comprises a second doubly-curved optical device, directs focused monochromatic x-ray beam 625 into a second point-focused monochromatic x-ray beam 655 to converge at a second focal point (not shown). In this embodiment, an object 640 to be imaged is disposed between the first optical device 620 and the second optical device 650 within the point-focused monochromatic x-ray beam. Imaging detector 660 is aligned with an axis of the second focused monochromatic x-ray beam 655 output from the second optical crystal 650. Gradients in refractive index in the object deflect the x-rays, which produce contrast in the manner termed “refractive index imaging” or “diffraction enhanced imaging” when employed with flat crystals. As a variation to the embodiment of FIG. 6, the first optical device 620 may be used with a Bragg angle near 45° so that its output beam is polarized, and the second optical device 650 may be placed orthogonal to the first optical device (and also have a Bragg angle near 45°) so that it functions as an analyzing filter. For example, in this embodiment, if the beam from the source travels in the x direction, the diffracted beam from the first optical device travels in the y direction, and the diffracted beam from the second optical device (which will not exist unless there is depolarization in the object) travels in the z direction. This allows polarized beam imaging which can increase the contrast and reduce background. Thus, doubly-curved optical devices are employed herein to produce a point-focused, monochromatic beam with high intensity and good angular resolution for various imaging applications. The monochromatic x-rays and the narrow output beam reduce the scatter produced in the patient or object, which improves the contrast and lowers the required dose. This is especially advantageous if no anti-scatter grid is employed. Employing pairs of optics as illustrated in FIG. 6 allows for high contrast refractive index imaging using conventional x-ray sources, or even low power x-ray sources (i.e., sources less than 1 kW). Experimental results for various monochromatic radiography implementations are described below with reference to FIGS. 7-12. FIG. 7 is a further schematic diagram of a monochromatic imaging system, in accordance with an aspect of the present invention. This system again employs an optical device comprising a point-focusing, doubly-curved optic, such as a doubly-curved crystal (DCC) optic. The imaging system includes an x-ray source 700 which provides x-rays 705 via slit 710 to optical device 720. Device 720 directs the x-rays in a focused monochromatic x-ray beam 725 through a second slit 730 towards a focal point 760. An object to be imaged 740 is disposed, in this example, after the focal point 760, with the detector 750 placed close by the object. The parameters of the DCC optic for the experiments performed are shown in Table 1. The source was a Microfocus Mo source, from Oxford Instruments, LTC, with a maximum source power of 120 watts, and the detector was a Fuji Computed Radiography Plate. TABLE 1DCCEnergyθBragg,Input focalOutput focalVerticalHorizontaloptics(keV)(°)dist. (mm)dist. (mm)convergence (°)convergence (°)Si (220)17.510.612012090.7 The DCC optic was mounted on two translation stages transverse to the x-ray beam and roughly placed at the designed focal distance from the source spot. Rough source to optic alignment was first performed with a CCD camera. The camera was placed at the 2θBragg angle at 180 mm from the optic to intercept the diffracted image and the optic scanned to obtain the maximum brightness. Then the diffracted intensity was recorded with a Ge detector (which gives total photon flux in the whole beam) and maximized by scanning the DCC optic along the x and y directions at different z positions. The resulting output spectrum is shown in FIG. 8. In FIG. 8, a diffraction spectrum with a DCC optic with source voltage of 25 kV, current of 0.1 mA and a recording time of 20 seconds is shown. Note that the measured spectral width is detector limited. Contrast measurements were performed with two plastic phantoms as depicted in FIG. 9. As shown in FIG. 9, a 5 mm step phantom 1100 of polypropylene was employed attached to a 10 mm polymethyl methacrylate block 1110. The phantom was placed at a 70 mm distance beyond the output focal point as shown in FIG. 7, with a computed radiography plate immediately adjacent. The beam size at the position was approximately 1 mm along the horizontal axis and 10 mm along the vertical axis. The phantom and the plate were mounted on three translation stages and scanned along the vertical direction. The image size of approximately 200 mm2 was obtained in an exposure time in the range 2-5 minutes with a 10 W source. This would correspond to an exposure of less than one second for a 10 kW pulse source. Monochromatic contrast measurements were also performed on the polypropylene phantom with 5 mm step height (see FIG. 9), then on a 12 mm thick polystyrene phantom (not shown) with two holes of 6 mm and 7 mm depths. For the polypropylene step phantom, the image with a monochromatic beam and its intensity profile are shown in FIG. 10. The measured and calculated contrasts are listed in Table 2. The measurements are in good agreement with the calculations. As listed in Table 2, a contrast ratio of 2.5 is obtained compared to the contrast measured with the full spectrum radiation before the optics. TABLE 2StepContrastContrastheightMono beamratio(mm)DataCalc.Data50.45 ± 0.050.382.5 The image of the polystrene phantom and its intensity profile are shown in FIG. 11. The figure shows two holes of different depths of 6 mm and 7 mm, with the intensity profile being taken through the dashed line. The source voltage was 35 kV at 10 W power. The phantom was translated at 0.1 mm/s with an exposure time of three minutes. The measured and calculated contrasts are shown in Table 3. TABLE 3Hole DepthContrast(mm)DataCalc.60.46 ± 0.070.4070.59 ± 0.100.49 For any imaging system, image resolution is an important parameter. For thick objects or large object-to-detector distances, spatial resolution is approximately proportional to the angular divergence, also called angular resolution. Angular resolution measurements were performed with an Oxford Instrument Microfocus 5011 molybdenum source. This source has a larger focal spot size, approximately 60 μm, somewhat closer to that of a clinical source than the 15 μm spot size of the Oxford Microfocus source. Angular resolution was measured by recording a knife edge shadow with a Fuji restimuble phosphor computed radiography image plate with 50 μm pixels. The knife-edge was placed after the crystal to block half of the output beam. The intensity profile recorded by the detector was differentiated and a Gaussian fit used to obtain the full width at half maximum. The detector angular resolution, σD, is 50 μm divided by the knife to detector distance. The results listed in Tables 4 and 5 have the detector resolution subtracted in quadrature; the measured angular resolution is assumed to be √{square root over (σ2+σD2)} where σ is the actual angular resolution of the x-ray beam. First, the resolutions were measured with the knife-edge 70 mm beyond the output focal point, which is 190 mm from the optics so that the beam is diverging rather than converging. The image plate was placed at 50, 200, 300, 450 mm distances to the knife-edge. The resolutions are different in the horizontal and vertical directions because the convergence angles are different. For this detector, which had 50 micron pixels, increasing the plate distance beyond 300 mm did not greatly improve the resolution, so the remaining measurements were made with a plate distance of 300 mm. Then measurements were performed for the different knife-edge positions shown in FIG. 12. The measured results are shown in Table 5. The angular resolution changes with distance from the optic as different fractions of the total convergence are sampled. As expected, the resolution is improved as the distance of the phantom from the output focal point is increased. However, the uniformity, which is quite good at 70 mm from the focal spot, as shown in FIGS. 10 & 11, is somewhat poorer at longer focal spot to phantom distances. The resolution might be further improved without a decrease in uniformity by increasing the output focal length in an asymmetric optic design. If the beam is to be scanned, uniformity in the direction along the scan is not critical. TABLE 4knife distanceplate distanceHorizontalVerticalto opticto knifeangular widthangular width(mm)(mm)(mrad)(mrad)190504.83 ± 0.246.05 ± 0.121902001.68 ± 0.152.97 ± 0.261903001.07 ± 0.152.78 ± 0.421904501.06 ± 0.272.73 ± 0.26 As shown in FIG. 8, the flux diffracted by the DCC optic was measured with a Ge detector. The diffraction efficiency η of the DCC optic, which is the ratio of the number of photons diffracting from the optic to incident photons on the surface of the optic, obtained with an aperture measurement, was calculated by: η DCC = C D C W A aper A optic ( d optic d aper ) 2 where CD is the diffracted counts, CW is the counts without the optic, Aaper is the area of the aperture, Aoptic is the effective area of the optic surface, doptic is the optic distance to the source, daper is the aperture distance to the source. Considering only those input photons within a 1 keV window from 17 to 18 keV, the diffraction efficiency η is 1.8%. Estimating the efficiency η as the ratio of the crystal bandwidth σc to the sum in quadrature of the of the source divergence σs and the energy angular bandwidth, σE gives: η = σ c σ S 2 + σ E 2 = 1.9 % where σs is the ratio of the source size to the source distance, about 0.5 mrad, and σE˜0.2 mrad is the angular width, computed from Bragg's law, due to the energy width of the characteristic line. The calculated η is in good agreement with the measurement. TABLE 5knife dist.Dist. fromHorizontalVerticalto opticfocal pointangular widthangular width(mm)(mm)(mrad)(mrad)1113−71.32 + 0.3513.7 + 1.62135151.89 + 0.4112.1 + 2.13190701.09 + 0.15 2.8 + 0.4242451250.87 + 0.15 2.0 + 0.4653001800.62 + 0.15 1.15 + 0.28 Measured diffraction fluxes using the DCC optic at different source voltages with a current of 0.1 mA are shown in Table 6. TABLE 6Source voltageMeasured flux(kV)(photons/s/kW)20(1.0 ± 0.01) * 10723(7.2 ± 0.03) * 10725(1.4 ± 0.03) * 10828(2.6 ± 0.05) * 10830(3.6 ± 0.06) * 108 The intensity at different distances to the optic is shown in Table 7. TABLE 7Detector distanceDetector distanceMeasured intensityto optic (mm)to focal point (mm)(photons/s/cm2)230110 (2.2 ± 0.001) * 1010310190(9.0 ± 0.009) * 109390270(5.1 ± 0.007) * 109500380(2.9 ± 0.005) * 109630510(1.8 ± 0.004) * 109 Those skilled in the art will note from the above discussion that x-ray imaging systems are presented herein which employ one or more point-focusing, curved monochromating optics for directing out a focused monochromatic x-ray beam for use in imaging an object, such as a patient. Monochromatization of x-rays can be readily achieved using doubly-curved crystal optics or doubly-curved multilayer optical devices. In addition, such optics are relatively easy to make in large sizes so that an x-ray imaging system as proposed herein is simpler and easier to produce then conventional approaches. Beam divergence of the x-ray imaging system can be controlled by increasing the optic-to-object distance, or changing the optic design to increase the output focal length. The optics discussed above include curved crystal optics (see e.g., X-Ray Optical, Inc.'s, U.S. Pat. Nos. 6,285,506, 6,317,483, and U.S. Provisional Application Ser. No. 60/400,809, filed Aug. 2, 2002, entitled “An Optical Device for Directing X-Rays Having a Plurality of Crystals”, and perfected as PCT Application No. PCT/US2003/023412, filed Jul. 25, 2003, and published under the PCT Articles in English as WO 2004/013867 A2 on Feb. 12, 2004)—each of which are incorporated by reference herein in their entirety); or similarly functioning multi-layer optics. The optics may provide beam gain, as well as general beam control. Also, as discussed above, monochromating optical elements may be desirable for narrowing the radiation bands depending on system requirements. Many of the optics discussed above, especially curved crystal optics and multi-layer optics, can be employed for this function, as set forth in many of the above-incorporated U.S. patents. Optic/source combinations are also useable such as those disclosed in X-Ray Optical Systems, Inc.'s U.S. Pat. No. 5,570,408, issued Oct. 29, 1996, as well as in U.S. Provisional Application Ser. Nos.: (1) 60/398,968 (filed Jul. 26, 2002, entitled “Method and Device for Cooling and Electrically-Insulating a High-Voltage, Heat-Generating Component,” and perfected as PCT Application PCT/US02/38803); (2) 60/398,965 (filed Jul. 26, 2002, entitled “X-Ray Source Assembly Having Enhanced Output Stability,” and perfected as PCT Application PCT/US02/38493); (3) 60/492,353 (filed Aug. 4, 2003, entitled “X-Ray Source Assembly Having Enhanced Output Stability Using Tube Power Adjustments and Remote Calibration”); and (4) 60/336,584 (filed Dec. 4, 2001, and entitled “X-Ray Tube and Method and Apparatus for Analyzing Fluid Streams Using X-Rays,” perfected as PCT Application PCT/US02/38792-WO03/048745, entitled “X-Ray Tube and Method and Apparatus for Analyzing Fluid Streams Using X-Rays”)—all of which are incorporated by reference herein in their entirety. Although preferred embodiments have been depicted and described in detail herein, it will be apparent to those skilled in the relevant art that various modifications, additions, substitutions and the like can be made without departing from the spirit of the invention and these are therefore considered to be within the scope of the invention as defined in the following claims. |
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claims | 1. An apparatus comprising:a chemical neutron emitter;a neutron shield having an aperture, at least one of the neutron shield and the chemical neutron emitter arranged for linear movement through a position that aligns the chemical neutron emitter with the aperture of the neutron shield; anda control unit that controls the movement of at least one of the chemical neutron emitter and the neutron shield to chop a neutron beam emitted from the chemical neutron emitter with the neutron shield. 2. The apparatus of claim 1, wherein:the chemical neutron emitter is fixed in position;the neutron shield is arranged for linear movement with respect to the chemical neutron emitter; andthe control unit controls linear movement of the neutron shield. 3. The apparatus of claim 2, wherein the neutron shield is arranged to oscillate linearly with respect to the chemical neutron emitter. 4. The apparatus of claim 1, further comprising:wherein the neutron shield is fixed in position;wherein the chemical neutron emitter is arranged for linear movement with respect to the neutron shield; andthe control unit controls linear movement of the chemical neutron emitter. 5. The apparatus of claim 4, further comprising:a platform, wherein the chemical neutron emitter is attached to the platform, the platform structured for linear movement with respect to the neutron shield; andwherein the control unit that controls the linear movement of the chemical neutron emitter moves the platform. 6. The apparatus of claim 5, wherein the platform is structured to oscillate linearly with respect to the neutron shield. 7. The apparatus of claim 4, wherein the control unit includes circuitry structured to regulate motion of the chemical neutron emitter or the neutron shield at a selected frequency. 8. The apparatus of claim 1, wherein the apparatus includes:a processing unit arranged to process signals collected in response to generation of neutrons from the aperture in a borehole; anda communications unit to transmit results from the processing unit. 9. The apparatus of claim 1, wherein the chemical neutron emitter includes one or more radioactive isotopes. 10. A method comprising:generating neutrons from a chemical neutron emitter; andlinearly moving one or more of the chemical neutron emitter or a neutron shield through a position at which an aperture of the neutron shield is aligned with the chemical neutron emitter to pass neutrons output through the aperture. 11. The method of claim 10, wherein linearly moving includes moving the neutron shield with the chemical neutron emitter fixed or linearly moving the chemical neutron emitter with the neutron shield fixed. 12. The method of claim 10, wherein linearly moving includes an oscillatory linear motion. 13. The method of claim 10, wherein linearly moving includes moving with a selected frequency of movement. 14. The method of claim 10, wherein linearly moving includes moving with a selected variation of frequency of movement. 15. The method of claim 10, further comprising:analyzing signals from generating the pulses of neutrons output in a borehole; anddirecting a drilling-based operation in response to analyzing the signals. 16. An apparatus comprising:a neutron shield having an aperture;a chemical neutron emitter arranged to move through a position that aligns the chemical neutron emitter with the aperture of the neutron shield; anda control unit that controls the movement of the chemical neutron emitter to chop a neutron beam emitted from the chemical neutron emitter with the neutron shield. 17. An apparatus of claim 16, wherein the apparatus includes a backstop partially encircling the chemical neutron emitter, the backstop disposed between the chemical neutron emitter and neutron shield. 18. The apparatus of claim 16, wherein the chemical neutron emitter is rotatable such that the chemical neutron emitter is operatively aligned with the aperture at one angular region in a rotation of the chemical neutron emitter. 19. The apparatus of claim 16, wherein the chemical neutron emitter includes one or more radioactive isotopes. 20. The apparatus of claim 16, wherein the chemical neutron emitter is rotatable with the neutron shield surrounding the chemical neutron emitter such that the chemical neutron emitter is operatively aligned with the aperture at one angular region in each rotation of the chemical neutron emitter. 21. The apparatus of claim 20, further comprising:a platform, wherein the chemical neutron emitter is attached to the platform, the platform structured to rotate with respect to the neutron shield; andwherein the control unit that controls movement of the chemical neutron emitter moves the platform. 22. The apparatus of claim 16, wherein the control unit includes circuitry to regulate motion of the chemical neutron emitter at a selected frequency. 23. The apparatus of claim 16, wherein the apparatus includes:a processing unit arranged to process signals collected in response to generation of neutrons from the aperture in a borehole; anda communications unit to transmit results from the processing unit. 24. A method comprising:generating neutrons from a chemical neutron emitter; andmoving the chemical neutron emitter through a position at which an aperture of a neutron shield is aligned with the chemical neutron emitter to pass pulses of neutrons output through the aperture. 25. The method of claim 24, wherein the neutron shield is fixed in position. 26. The method of claim 24, wherein moving the chemical neutron emitter comprises moving the chemical neutron emitter in a rotational motion. 27. The method of claim 24, wherein moving the chemical neutron emitter comprises moving with a selected frequency of movement. 28. The method of claim 24, wherein moving the chemical neutron emitter comprises moving with a selected variation of frequency of movement. 29. The method of claim 24, further comprising:analyzing signals from generating the pulses of neutrons output in a borehole; anddirecting a drilling-based operation in response to analyzing the signals. |
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06326627& | abstract | A device and method for separating ions uses electric and magnetic fields that are specifically configured and oriented in a vacuum chamber. Also, a central electrode that is made of the materials whose ions are to be separated is positioned in the chamber. Magnetic coils mounted on the chamber generate a magnetic field, B, that is oriented parallel to the central electrode and is configured with a disk-shaped magnetic mirror at one end of the chamber, and an annular-shaped magnetic mirror at the other end. A plurality of electrodes generate an electric field, E, that is oriented perpendicular to the central electrode. In operation, neutral atoms in the chamber are ionized by the electric field. The electric field, however, is specifically configured to confine relatively lighter mass ions in the chamber. These ions are then subsequently removed from the chamber through the opening in the annular-shaped magnetic mirror. Simultaneously, the electric field directs the heavier mass ions into contact with the central electrode, to thereby sputter the electrode and generate additional neutral atoms for ionization in a sustained operation. |
claims | 1. A method of producing a localised concentration of energy comprising:creating at least one shockwave;propagating the at least one shockwave through a non-gaseous medium;allowing the at least one shockwave to be incident upon a pocket of gas suspended within the medium, wherein the pocket of gas is spaced from a concave surface; andreflecting said at least one shockwave from the concave surface onto said gas pocket. 2. The method as claimed in claim 1, wherein the concave surface comprises a plurality of discrete portions. 3. The method as claimed in claim 2, wherein the discrete portions are piecewise polynomial. 4. The method as claims in claim 1, wherein the concave surface focuses the reflected at least one shockwave onto the gas pocket. 5. The method as claimed in claim 4, wherein the concave surface focuses the reflected shockwave to a point. 6. The method as claimed in claim 1, wherein the gas pocket is placed no more than three times a maximum radius of curvature of the closest section of the concave surface away from the concave surface. 7. The method as claimed in claim 6, wherein the gas pocket's edge closest to the concave surface is spaced from it by a distance of less than five times the dimension of the gas pocket's widest part. 8. The method as claimed in claim 1, comprising using an external device to apply one or more shockwaves to a static volume of the non-gaseous medium to create the at least one shockwave propagating through the non-gaseous medium. 9. The method as claimed in claim 8, comprising using the external device to create the shockwave with a pressure of between 0.1 GPa and 50 GPa. 10. The method as claimed in claim 1, comprising using a lithotripsy device to create the shockwave with a pressure of between 100 MPa and 1 GPa. 11. The method as claimed in claim 1, wherein the gas pocket is formed with the use of a membrane that defines the boundary between the gas pocket and the non-gaseous medium, and wherein the membrane is frangible and breaks upon impact from the shockwave. 12. The method as claimed in claim 11, wherein the membrane includes a line or region of weakness that breaks upon impact from the shockwave. 13. An apparatus for producing a localised concentration of energy comprising:a non-gaseous medium having therein a pocket of gas, wherein the pocket of gas is spaced from a concave surface; andan external device for creating at least one shockwave propagating through said medium so as to be incident upon said pocket of gas,wherein said concave surface is shaped so as at least partially to reflect said shockwave in such a way as to direct it onto said gas pocket and wherein the gas pocket is placed no more than three times a maximum radius of curvature of the closest section of the concave surface away from the concave surface. 14. The apparatus as claimed in claim 13, wherein the non-gaseous medium comprises a static volume of non-gaseous medium and wherein the external device is arranged to apply one or more shockwaves to the static volume of non-gaseous medium to create the at least one shockwave propagating through the non-gaseous medium. |
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062815083 | claims | 1. A method for assembling a microlens assembly, comprising: providing a first microlens component and a second microlens component; forming a first base opening in said first microlens component; forming a first alignment opening in said second microlens component; and threading a first rigid aligner through the first base opening and the first alignment opening, thereby aligning the first microlens component and the second microlens component. said forming a first base opening comprises forming a first base opening and a second base opening; said forming a first alignment opening comprises forming a first alignment opening and a second alignment opening both in said second micro lens component; and threading a second aligner through the second base opening and the second alignment opening. the threading a first aligner includes threading the first aligner through the first alignment opening such that the first aligner exerts a force on an inside edge of the first alignment opening in a first direction; and the threading a second aligner includes threading the second aligner through the second alignment opening such that the second aligner exerts a force on an inside edge of the second alignment opening in a second direction, said second direction being substantially opposite said first direction. the forming a first base opening comprises etching the first base opening in the first microlens component; and the forming a first alignment opening comprises etching the first alignment opening in the second microlens component. said first alignment opening includes a threading portion and an intersecting locking portion, said locking portion being eccentric with said threading portion and having a diameter smaller than a diameter of the threading portion; said threading includes inserting the first aligner through the threading portion of the first alignment opening, and moving the first aligner into the locking portion of the first alignment opening. a first microlens component including a first base opening; a spacer disposed on a first surface of the first microlens component; a second microlens component defining a first alignment opening, said second microlens component being disposed on a surface of the spacer opposite the first microlens component; a first aligner extending through the first base opening and the first alignment opening. a second base opening in said first microlens component; a second alignment opening in said second microlens component; and a second aligner extending through the second base opening and the second alignment opening. the first aligner exerts a force on an inside edge of the first alignment opening in a first direction; and the second aligner exerts a force on an inside edge of the second alignment opening in a second direction, said second direction being substantially opposite said first direction. the first direction of the force exerted by the first aligner is towards a vertex of the first alignment opening, whereby the first aligner contacts an inside edge of the first alignment opening at each of two sides of the polygon forming the vertex; and the second direction of the force exerted by the second aligner is towards a vertex of the second alignment opening, whereby the second aligner contacts an inside edge of the second alignment opening at each of two sides of the polygon forming the vertex. the spacer defines a first spacer opening; the first aligner additionally extends through the first spacer opening. said first alignment opening includes a threading portion and an intersecting locking portion, said locking portion being eccentric with said threading portion and having a diameter smaller than a diameter of the threading portion; said first aligner extends through the first base opening and the locking portion of the first alignment opening. 2. The method of claim 1, further comprising bonding the first and second microlens components to an insulating layer disposed therebetween. 3. The method of claim 1, wherein: 4. The method of claim 3, wherein: 5. The method of claim 4, wherein said forming a second alignment opening includes forming the first and second alignment openings in the shape of a polygon. 6. The method of claim 5, wherein said forming a second alignment opening includes forming the first and second alignment openings in the shape of a square having sides longer than a diameter of the first and second aligners. 7. The method of claim 1, wherein the first aligner is a length of glass fiber. 8. The method of claim 7, wherein the glass fiber is borosilicate glass. 9. The method of claim 1, wherein the first and second microlens components are comprised of silicon. 10. The method of claim 9, wherein: 11. The method of claim 10, wherein said etching act includes anisotropic dry etching. 12. The method of claim 10, wherein said etching act includes reactive ion etching. 13. The method of claim 1, wherein: 14. The method of claim 13, wherein said threading includes moving the first aligner into the locking portion of the first alignment opening, said locking portion of the first alignment opening having a diameter approximately equal to a diameter of the first aligner. 15. A microlens, comprising: 16. The microlens of claim 15, further comprising: 17. The microlens of claim 16, wherein: 18. The microlens of claim 17, wherein the first and second alignment openings are in the shape of a polygon. 19. The microlens of claim 18, wherein said polygon is a square having sides longer than a diameter of the first and second aligners. 20. The microlens of claim 18, wherein: 21. The microlens of claim 15, wherein the first and second microlens components are bonded to the spacer. 22. The microlens of claim 15, wherein: 23. The microlens of claim 15, wherein the first aligner is comprised of glass fiber. 24. The microlens of claim 15, wherein the first aligner is comprised of borosilicate glass fiber. 25. The microlens of claim 15, wherein the first aligner is round in cross-section. 26. The microlens of claim 15, wherein the first and second microlens components are comprised of silicon. 27. The microlens of claim 15, wherein: 28. The microlens of claim 27, wherein the locking portions of the first alignment opening has a diameter approximately equal to a diameter of the first aligner. |
claims | 1. A method to scrabble an exposed surface of structural materials, comprising: a) coating the exposed surface of structural materials with one or more layers of inert material; b) applying a layer of explosive material to the exposed surface of the one or more layers of inert material, such that said layer of explosive material is sufficiently thick to sustain a detonation front propagating parallel to the exposed surface; and, c) initiating a detonation front in the layer of explosive material so that the detonation front freely propagates throughout the layer of explosive material. 2. The method of claim 1 , wherein the exposed surface of the layers of inert material is flat and smooth. claim 1 3. A method to scabble an exposed surface of structural materials, comprising: a) applying a layer of explosive foam onto the exposed surface, said layer having sufficient thickness to sustain a detonation front propagating parallel to the exposed surface, wherein applying said layer of explosive foam comprises coating the exposed surface with an explosive foam precursor, and transforming the explosive foam precursor into a layer of explosive foam; and, b) initiating a detonation front in the layer of explosive foam so that the detonation front freely propagates throughout the layer of explosive material. 4. A method to scabble an exposed surface of structural materials, comprising: a) applying a layer of explosive foam onto the exposed surface, said explosive foam comprising an aqueous solution of hydroxylammonium nitrate (HAN) and a foaming agent, and said layer having sufficient thickness to sustain a detonation front propagating parallel to the exposed surface; and, b) initiating a detonation front in the layer of explosive foam so that the detonation front freely propagates throughout the layer of explosive material. 5. The method of claim 4 , wherein the liquid explosive composition further comprises between 10% and 60% triethanolammonium nitrate (TEAN) by weight. claim 4 6. The method of claim 5 , wherein the liquid explosive composition consists essentially of 73% HAN by weight, 23%TEAN by weight, and 4% water by weight. claim 5 |
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abstract | An accelerator (10) generates an electron beam (22) of selected energy that is swept (16) up and down. A conveyor (32) moves items (30) through the electron beam for irradiation treatment. An array (40a) of inductive electron beam strength detectors is disposed on a down stream side of the item to detect the energy of the electron beam exiting the item at the plurality of altitudes. The electron beam strength entering and leaving the item are communicated to a processor (54) which determines the absorbed dose of radiation absorbed by the item. The dose information is archived (56) or compared by a parameter adjustment processor (58) with target doses and deviations are used to control one or more of MeV or beam current of the electron beam, the sweep rate, and the conveying speed of the items. Each of the detectors includes a vacuum chamber in which two current transformers (60, 62) disposed on either side of a metal foil layer (64). From the difference in the current induced in the two transformers by a pulsed, collimated electron beam, the energy of the beam is determined. |
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claims | 1. A method for producing a micro-gripper for clamping a micrometer-scale or sub-micrometer-scale object in a receptacle slot for gripping and holding between a first gripping element and a second gripping element, the method comprising the steps:providing a flat substrate;applying a first sacrificial layer over the flat substrate;forming the first gripping element by applying at least one first material layer to the first sacrificial layer;applying a second sacrificial layer to an area of the first material layer at least in an area where the receptacle slot is to be formed;forming the second gripping element by applying a second material layer to the second sacrificial layer in the area where the slot is to be formed and forming a base body of the micro-gripper by applying the second material layer to an area of the first material layer to bond the first material layer to the second material layer;removing the second sacrificial layer to provide the receptacle slot with free ends respectively comprising the first and second material layers with the receptacle slot being provided between the first material layer and the second material layer; andremoving an area of the first sacrificial layer to detach the micro-gripper from the flat substrate. 2. The method according to claim 1, further comprising:applying the first sacrificial layer only to at least one surface area on the flat substrate or, after applying the first sacrificial layer on the flat substrate, structuring the first sacrificial layer. 3. The method according to claim 1, further comprising:applying the first material layer to the first sacrificial layer only to at least one surface area of the first sacrificial layer or, after applying the first material layer to the first sacrificial layer, structuring the first material layer. 4. The method according to claim 1, further comprising:structuring the second sacrificial layer after applying the second sacrificial layer to an area of the first material layer. 5. The method according to claim 1, further comprising:applying the second material layer to at least one area of the second sacrificial layer and to the first material layer or structuring the second material layer after applying the second material layer to the at least one layer of the sacrificial layer and to the first material layer. 6. The method according to claim 1, further comprising:applying at least one intermediate layer on the first substrate before applying the first sacrificial layer over the flat substrate. 7. The method according to claim 2, further comprising:applying at least one intermediate layer on the first substrate before applying the first sacrificial layer over the flat substrate. 8. The method according to claim 3, further comprising:applying at least one intermediate layer on the first substrate before applying the first sacrificial layer over the flat substrate. 9. The method according to claim 4, further comprising:applying at least one intermediate layer on the first substrate before applying the first sacrificial layer over the flat substrate. 10. The method according to claim 5, further comprising:applying at least one intermediate layer on the first substrate before applying the first sacrificial layer over the flat substrate. 11. The method according to claim 2, further comprising:applying at least one stabilization layer to the second material layer. 12. The method according to claim 3, further comprising:applying at least one stabilization layer to the second material layer. 13. The method according to claim 4, further comprising:applying at least one stabilization layer to the second material layer. 14. The method according to claim 5, further comprising:applying at least one stabilization layer to the second material layer. 15. The method according to claim 6, further comprising:applying at least one stabilization layer to the second material layer. 16. The method according to claim 7, further comprising:applying at least one stabilization layer to the second material layer. 17. The method according to claim 8, further comprising:applying at least one stabilization layer to the second material layer. 18. The method according to claim 9, further comprising:applying at least one stabilization layer to the second material layer. 19. The method according to claim 10, further comprising:applying at least one stabilization layer to the second material layer. 20. The method according to claim 1, further comprising:using plastic, ceramic, metal, or silicon oxide as the sacrificial layers. 21. The method according to claim 1, further comprising:using plastic, ceramic, metal, semimetal, or polysilicon as the material layers. 22. The method according to claim 1, further comprising:structuring the sacrificial layers and/or the first and/or second material layer by masking the layers and chemical etching, reactive ion etching (RIE), sputter etching, ion beam etching, or plasma etching masked layers. 23. The method according to claim 1, further comprising:using CVD, LPCVD, PECVD, PVD, or galvanic deposition to provide material deposition. 24. The method according to claim 1, wherein:the first and the second material layers are different materials. 25. The method according to claim 2, further comprising:clamping the micrometer scale or the sub-micrometer scale object between the first material layer and the second material layer. |
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description | The present invention relates to a fault monitoring method for a robot or other work machines in which a servo motor is used as a drive source. Servo motors are a type of control motor, and are characterized as having exceptional rotational speed control and positional control. Servo motors are installed in a variety of work machines, and in particular are widely employed as driving sources for robots. Servo motors are capable of directly driving a load. However, the load is often driven via a reducer or other power transmission mechanism. This is due to the fact that the output torque from the servo motor can be markedly increased when the load is driven via a reducer. In addition, if the reducer is composed of a belt, chain, or drive shaft, a benefit will be presented in that the movement can be transmitted to a location that is distant from the servo motor. The fact that faults occur in servo motors and power transmission mechanisms in a work machine must be taken into account. Faults that occur in work machines range from serious faults, which are critical enough to necessitate an emergency shutdown of the machine; minor faults, which cause no interference even if operation is continued until the time of the next repair; and moderate faults, which are between the serious faults and the minor faults in terms of severity. The presence of faults in the servo motor itself can be determined by detecting anomalies via monitoring electric current values or performing other electrical monitoring. On the other hand, anomalies are difficult to detect when the servo monitor is operating normally but a fault has occurred in the power transmission mechanism. A detecting technique such as disclosed in JP-5-346812 A and JP-11-129186 A has been required. In the anomaly-detecting device disclosed in JP-5-346812 A, a post-feedback command signal is integrated when the servo motor is being controlled by a controlling device via feedback. The time required for the resulting integration value to reach saturation is monitored. For example, the servo motor drives an arm as a load via a belt. However, if the belt breaks, the load suddenly changes, and the integration value changes markedly. Specifically, the load decreases, which causes fluctuations in the load to become less pronounced and the magnitude of the post-feedback command signal to decrease. When this occurs, the time required for the integration value to reach saturation (saturation time) increases. If a configuration is used in which an anomaly is determined when the saturation time extends past a reference time, the breaking of a belt or another anomaly can be detected. However, an anomaly cannot be detected if the change in the load fluctuation is small. In other words, the anomaly detecting device disclosed in JP-5-346812 A is intended for discovering serious faults such as the breaking of a belt, and is unable to address minor faults. In the work robot disclosed in JP-11-129186 A, the work rate W1 on the drive side of a driving shaft is calculated on the basis of a driving current Ii and actual angle θi of a servo motor. A work rate W0 on the load side of the driving shafts is additionally calculated on the basis of the actual angle θi and a motion equation that relates to the particle model of a robot mechanism part. The difference or ratio between the resulting work rate W1 and the work rate W0 is compared with an established criterion. If a gear or the like is worn down due to change over time, the work rate W1 will generally be greater than the work rate W0 during acceleration. The difference or ratio at this time will become progressively more pronounced as the wear on the gear or other components increases. In other words, a detectable difference or ratio will not be generated in the event of small-scale wear. This approach is suitable for serious faults in which a difference occurs between the work rates, but is unsuitable for minor faults in which a difference between the work rates is not likely to occur. A first problem arises in that a monitoring technique that corresponds to minor faults is not established in JP-5-346812 A or JP-11-129186 A. It is also possible that, e.g., a turntable and a robot occupy a single work area. The turntable, in which a servo motor is used as a driving source, would be monitored by the anomaly-detecting device disclosed in JP-5-346812 A, and the robot would be monitored by the anomaly sensor of the work robot disclosed in JP-11-129186 A. Faults will be detected in the turntable and robot using different fault monitoring systems. In other words, when there are multiple types of work machines manufactured by different makers, the fault monitoring systems attached to the servo motors for driving the work machines tend to have various functions. In such instances, diverse fault monitoring systems must be arranged in the work area (the production site or the surrounding area), substantial effort is required to maintain control over the systems, and greater space is needed to accommodate the systems. In other words, the second problem arises from the fact that no technique has been established for allowing central control over a variety of servo motors when a variety of work machines are present. A demand has accordingly arisen for a technique whereby a monitoring technique corresponding to minor faults can be performed and a variety of servo motors can be centrally controlled. According to the present invention, there is provided a fault monitoring method for a robot or other work machine, wherein output from a servo motor is transmitted via a reducer or other power transmission mechanism and a load is driven, the method comprising the steps of: acquiring first torque data generated from the servo motor, in units each of which starts when the motor starts operating and ends when the motor stops operating; selecting a maximum first torque fluctuation range designated by the difference between a maximum torque and a minimum torque for each unit obtained from the acquired first torque data; collecting maximum fluctuation ranges of the first torque for a plurality of cycles and obtaining a first average value; selecting a fluctuation range control value by multiplying the first average value by a factor greater than 1.0; acquiring second torque data generated from the servo motor, in units each of which starts when the motor starts operating and ends when the motor stops operating, after the fluctuation range control value has been selected; selecting a maximum second torque fluctuation range designated by the difference between a maximum torque and a minimum torque among the units obtained from the acquired second torque data; collecting a second maximum fluctuation range for a plurality of cycles and obtaining a second average value; making a comparison to determine whether the second average value exceeds the fluctuation range control value; and determining that a fault has occurred when the second average value exceeds the control value for the fluctuation range in the comparison. The average value of the maximum fluctuation range of the torque is used as a control value and a monitoring value. The difference between the maximum torque and minimum torque per unit is the fluctuation range. When the temperature or another external factor changes and the maximum torque increases, the minimum torque will also increase. In other words, the fluctuation range will remain steady without changing markedly even if the temperature or another external factor changes. The use of a torque fluctuation range is beneficial for this reason. However, there are instances in which a noise signal is mixed into the signal system or another event occurs and the fluctuation range momentarily undergoes a substantial change. The average value is used as a control value in such instances. As a result, the effects of instantaneous fluctuations can be reduced. Instantaneous changes are thus ignored. Therefore, this configuration is favorable in monitoring minor faults that worsen only gradually, and a technique allowing minor faults to be addressed can be provided for a robot or another work machine in which a servo motor is used as a driving source. Preferably, a plurality of servo motors is provided, a fluctuation range control value is set for each of the servo motors, and the servo motors are uniformly managed by a single checking unit. In other words, control is provided using torque, which is a joint index that characterizes different types of servo motors. As a result, the servo motors can be centrally controlled by a single checking unit. If uniform control is provided, the motors will be easier to control, only one fault monitoring system will suffice, and the installation space for the system can be reduced. Desirably, the first torque data and second torque data are acquired from a motor driver or controller for controlling the servo motors. A torque sensor need not be newly prepared and installed. Therefore, the present invention can readily be applied to pre-existing work machines, and the cost of installation can be reduced even in new work machines. Preferably, the acquired torque data is converted to consolidated data by a data converter and then sent to the checking unit. Having the data be consolidated by the data converter will allow a plurality of torque data sets to be efficiently processed and sent to an input unit. In a preferred form, a warning signal is generated from an alerting unit when a fault is determined to be present during a fault-determining step. If a lamp is illuminated, an alarm is sounded, or another action is performed on the basis of the warning signal, a worker can be alerted, and serious accidents can be averted. In FIG. 1, a robot 10 and turntable 30 are shown as specific examples of work machines. The robot 10 used as a work machine is an articulated robot composed of a first member 13 rotatably supported on a base 11 via a first shaft 12; a second member 15 oscillatably supported in a vertical direction on the first member 13 via a second shaft 14; a third member 17 oscillatably supported in a vertical direction on the second member 15 via a third shaft 16; and a fourth member 19 rotatably supported on the third member 17 via a fourth shaft 18. The robot 10 is called a welding robot when the fourth member 19 is a welding gun, and is called a painting robot when the fourth member 19 is a spray gun. In this example, a second reducer 21 is connected to the second shaft 14 as a power transmission member, and a second servo motor 22 is connected to the second reducer 21. The rotational speed and other aspects of the second servo motor 22 are controlled by a second motor driver 23. A third reducer 24 is connected to the third shaft 16 as a power transmission member in a similar manner, and a third servo motor 25 is connected to the third reducer 24. The rotational speed and other aspects of the third servo motor 25 are controlled by a third motor driver 26. The first shaft 12 and fourth shaft 18 are configured in a similar manner; however, the power transmission member, servo motor, and motor driver are omitted. The motor drivers 23, 26 and other motor drivers in the robot are housed together in a robot control board 28 along with a controller 27. In the turntable 30 that is used as a work machine, an oscillating member 33 is mounted to a fixed member 31 via an oscillating shaft 32; a lower plate 34 is fixed to the oscillating member 33; and a rotating plate 36 is rotatably mounted to the lower plate 34 via a rotating shaft 35. A work piece 37 mounted to the rotating plate 36 is inclined about the oscillating shaft 32 and rotates about the rotation shaft 35. As a result, the orientation of the work piece 37 can be freely changed. An oscillating reducer 38 is connected to the oscillating shaft 32 as a power transmission member, and an oscillating servo motor 39 is connected to the oscillating reducer 38. The oscillation speed and other aspects of the oscillating servo motor 39 are controlled by an oscillating motor driver 41. The rotating shaft 35 is configured in a similar manner, but the power transmission member, servo motor, and motor driver are omitted. The motor driver 39 and other motor drivers in the turntable are housed together in a turntable control board 43 along with a controller 42. Preferably, a fault monitoring device 50 is disposed in the vicinity of the robot 10 and turntable 30, and signal wires 51, 52 are connected to the robot control board 28 and the turntable control board 43. The torque data is introduced to the fault monitoring device 50 from the controller 27 and motor driver 41 via these wires. The second motor driver 23, third motor driver 26, and oscillating motor driver 41 shall be described in detail hereunder. The second motor driver 23 has a position controller 54, a speed controller 55, an electric current amplifier 56, a speed converter 57, and three adders 58, 59, 61, as shown in FIG. 2. The rotational speed of the second servo motor 22 is monitored by a rotary encoder 62, fed back to the adder 58 in the form of a position feedback signal Xf, and used for positioning control. When a command Xs is transferred from the upper controller to the second motor driver 23, the position feedback signal Xf is added by the adder 58. The position controller 54 that has received the resulting signal generates a speed command Vs. The rotational speed of the second servo motor 22 is converted to a speed feedback signal Vf by the speed converter 57, and then fed back to the adder 59 and used for positioning control. In other words, the speed feedback signal Vf is added to the speed command Vs by the adder 59. The speed controller 55 that has received the resulting signal generates a torque command Ts. A torque signal Tf is then added to the torque command Ts by the adder 61. The electric current amplifier 56 that has received the resulting signal outputs a drive current and drives the second servo motor 22. The torque command signal Ts of the electric current amplifier 56 and the feedback signal Tf fed back to the torque command signal Ts are both electrical current values. However, in the present example, both indicate the magnitude of the torque of the second servo motor 22, and therefore correspond to “torque.” The second so-configured motor driver 23 can be obtained as a commonly available product, and comprises a speed terminal 63 able to extract a speed signal, and a torque terminal 64 able to extract torque data (“torque signal” hereunder). The third motor driver 26, oscillating motor driver 41, and other motor drivers shown in FIG. 1 have the same configuration as the second motor driver 23, and a description thereof shall accordingly be omitted. The speed signal extracted from the speed terminal 63 of FIG. 2 is described in FIG. 3A. As shown in FIG. 3A, changes in the rotational speed are recorded as a curved line in a graph in which the horizontal axis indicates time, and the vertical axis indicates the speed signal Vf of the rotational speed. In FIG. 3A, movement starts at point P1, a control speed is reached at point P2, deceleration starts at point P3, and the motor stops at point P4. The period that starts with the start of the movement and ends with the end of the movement is called a “unit.” Two units are present in FIG. 3A. The torque extracted from the torque terminal 64 of FIG. 2 is described in FIG. 3B. Points that correspond to the points P1 through P4 shown in FIG. 3A are indicated by points P11 through P14 in FIG. 3B. In FIG. 3B, change in torque is recorded via a curved line in a graph in which the horizontal axis indicates time, and the vertical axis indicates the torque signal Tf. In FIG. 3B, the torque suddenly decreases starting at point P11. When a control speed is reached, the torque gradually decreases and moves toward a minimum at point P12. Reverse torque is generated starting from point P13, and braking is performed. At around point P14, the rotational speed of the motor becomes zero and the torque approaches zero. In general, the torque directly after point P11 is called starting torque, the torque after point P12 is called normal torque, and the torque directly after point P13 is called braking torque. As the starting torque and braking torque increase in magnitude, responsiveness improves, and the position and speed can be controlled very precisely. Meanwhile, the region L1 in which the effects of the starting torque and braking torque are not experienced is regarded as a stable region for normal torque. A range that contains one such region L1 corresponds to the “unit” shown in FIG. 3A. When the control is repeated, the next region L2 will appear. An enlarged view of the region L1 shall next be described. The region L1 is the curved line of a waveform, as shown in FIG. 4. The region L1 of the first torque data contains one maximum torque and one minimum torque. The maximum torque shall be called T1max, and the minimum torque shall be called T1min. Maximum torque T1max−Minimum torque T1min is defined as a maximum fluctuation range ΔT1 in the region L1 of the first torque data. The maximum fluctuation range (ΔT2) is obtained from the region L2 (see FIG. 3) of the first torque in a similar manner. Maximum fluctuation ranges (ΔT3 to ΔTm) of the first torque are obtained from the remaining regions (L3 to Lm) of the first torque data. A first average value is described in FIG. 5. In FIG. 5, the horizontal axis indicates the number of times (n), and the vertical axis indicates the maximum fluctuation range. In the graph, the maximum fluctuation ranges ΔT1 to ΔTm of the first torque data are plotted, and have data points that vary in the vertical direction. The horizontal line is a first average value obtained by averaging the maximum fluctuation ranges ΔT1 to ΔTm of the first torque data. Steps of a basic data acquisition phase according to the present invention shall next be described on the basis of the above descriptions. The basic data acquisition phase is configured as follows. When the work machine is used in production for the first time, the point at which trial running is complete is preferably the basic data acquisition phase. When the power transmission mechanism of the work machine is repaired or a component is replaced, a point at which normal operation becomes possible is preferably the basic data acquisition phase. In other words, basic data is acquired when the work machine, particularly the power transmission mechanism, is not in a state of advanced deterioration over time. As long as deterioration over time has not occurred, the basic data can be acquired at any point, and the basic data acquisition phase therefore does not need to be set firmly. FIG. 6 is a flowchart showing the steps of the basic data-acquisition phase. Step (abbreviated as ST hereunder) 01: 1 is input for n. ST02: T1max and T1min in the region L1 (FIG. 4) are acquired from the first torque data. ST03: The maximum fluctuation range ΔT1 of the first torque in the region L1 (FIG. 4) is calculated. ST04: ΔT1 is saved. ST05: A decision is made as to whether n has reached a predetermined repeat count (maximum value) m. ST06: If not; i.e., if n<m, then 1 is added to n, and the process returns to ST02. ST07: Once the necessary ΔT1 through ΔTm have been acquired, an average value is determined, and the average value is set as the first average value. The first average value is not greatly affected by the ΔT5 shown in FIG. 5. ST08: The first average value is multiplied by K (where K is a value greater than 1.0), whereby a fluctuation range control value is set. K is set with consideration given to the degree to which the work machine is able to continue operating in a stable manner. Preparation for normal operation is thus completed. Fault observation during normal operation (“fault monitoring”) shall accordingly be described next. FIG. 7 is a flowchart showing steps after the fluctuation range control value has been selected. ST11: The fluctuation range control value (FIG. 6, ST08) is retrieved. ST12: 1 is input as n. A waveform diagram of regular torque during normal operation, i.e., the second torque data, shall next be described with reference made to FIG. 8. In FIG. 8, AL1 corresponds to a region during normal operation (corresponding to the region L1 shown in FIG. 4). The region AL1 of the second torque data contains one maximum torque and one minimum torque. The maximum torque shall be called AT1max, and the minimum torque shall be called AT1min. Maximum torque AT1max−Minimum torque AT1min is defined as the maximum fluctuation range ΔT1 in the region AL1 of the second torque data. Returning to FIG. 7: ST13: AT1max and AT1min (FIG. 8) are acquired from the second torque data. ST14: ΔAT1 (FIG. 8), which is the maximum fluctuation range for the second torque, is calculated. ST15: ΔAT1 is saved. ST16: A decision is made as to whether n has reached a predetermined repeat count (maximum value) M. ST17: If not, i.e., if n<M, then 1 is added to n, and the process returns to ST13. ST18: Once the necessary ΔAT1 through ΔATM have been acquired, an average value is determined and set as the second average value. Variation in the second average value shall be described with reference made to FIG. 9. The horizontal axis indicates time, and the vertical axis indicates the second average value. In the graph, the first average value set in the basic data acquisition phase is included as a horizontal line, and a fluctuation range control value, which is twice the value of the first average value (i.e., K=2.0), is included as a horizontal line. One bar of the bar graph indicates the second average value at a given point in time. The bar t1 is the second average value directly after the start of normal operation. This second average value is substantially level with the “first average value” indicated by the horizontal line. The bar t10 represents a second average value after a short time has elapsed since the start of normal operation (e.g., six months), and the value is greater than the bar t1. In other words, the bar t10 indicates that the fluctuation range has increased. The bar t100 represents a second average value after considerable time has passed since the start of normal operation (e.g., twenty-four months), and is markedly greater than the bar t1. The bar t100 indicates that the fluctuation range control value has been exceeded. A fault is determined to have occurred at the point at which the bar t100 appears. Returning to FIG. 7: ST19: A decision is made as to whether the second average value has exceeded the fluctuation range control value. ST20: If the response in ST19 is “NO,” the machine is in the state indicated by the bar t1 or t10 in FIG. 9, and is therefore normal. ST21: If the response in ST19 is “YES,” the machine is in the state indicated by the bar t100 in FIG. 9, and a warning signal is therefore generated. A variety of faults can be expected to occur in work machines. For example, in a gear-type reducer, wear on surfaces of the gear teeth and wear on a roller surface of a bearing that supports a rotating shaft develop over time. This phenomenon is referred to as aging. Aging occurs gradually, as described in FIG. 9. An examination shall be made as to the suitability of a fault monitoring method that is an example of a prior art technique, wherein the maximum torque is constantly monitored and a fault is determined to have occurred when the torque exceeds a threshold value. In a servo motor of a work machine, high torque may momentarily occur due to a variety of causes. If such high torque is taken into account when the threshold value is selected, the fault will be impossible to detect [using the prior art technique] if wear and other parameters are not in an advanced state. Minor faults wherein deterioration occurs gradually cannot be detected with methods for monitoring maximum torque using a threshold value. For this reason, a maximum fluctuation range is obtained in the present invention for each unit during normal operation. This maximum fluctuation range is stored over a set period of time (e.g., several days or one week) and averaged to obtain the second average value. The second average value is compared with the fluctuation range control value. Employing an average value for the value of the object being monitored will minimize the effect caused by instantaneous fluctuation. In other words, instantaneous fluctuations are disregarded, allowing minor faults resulting from gradual deterioration to be efficiently detected. A technique for providing uniform control over multiple servo motors shall be described hereunder. A fault monitoring device 50 comprises a data converter 65, a checking unit 66, a memory 67 for fluctuation range control values, and an alerting unit 68, as shown in FIG. 10. The fault monitoring device 50 collectively monitors, e.g., a robot control board 28A for a robot (comprising, e.g., reducers 21A, 24A and servo motors 22A, 25A) manufactured by company A, a robot control board 28B for a robot (comprising, e.g., reducers 21B, 24B and servo motors 22B, 25B) manufactured by company B, a general-purpose device control board 43C for a general-purpose device (comprising, e.g., a reducer 38C and a servo motor 39C) manufactured by company C; and other control boards. The data converter 65 acquires multiple torque data sets from a torque terminal 64A of the robot control board 28A, a torque terminal 64B of the robot control board 28B, a torque terminal 64C of the general-purpose device control board 43C, and torque terminals of the other control boards; converts the acquired data to consolidated data as necessary; and outputs the data in the stated order to the checking unit 66. A fluctuation range control value for the second servo motor, a fluctuation range control value for the third servo motor, and a fluctuation range control value for the oscillation servo motor are stored in the memory 67 for fluctuation range control values. These fluctuation range control values are input or modified via a keyboard or other input means 69. The sequence of actions shown in FIG. 7 is performed in parallel for a plurality (three in this example) of servo motors 22A, 22B, 39C by a single checking unit 66. When a fault is detected, the alerting unit 68 generates a warning signal, and the lamp 71 illuminates. An alarm may be sounded at the same time. If a lamp is illuminated an alarm is sounded, or another action is performed on the basis of the warning signal, a worker can be alerted, and serious accidents can be averted. The present invention is characterized in that torque data is introduced to the checking unit 66. The torque terminals 64A, 64B, 64C are usually provided to the motor drivers 23, 26, 41 (FIG. 1), the controllers 27, 42 for controlling the motor drivers, or control boards 28A, 28B, 43C (FIG. 10) for accommodating the controllers. Torque can be obtained in joint units even if the model number or characteristics of the servo motors 22A, 22B, 39C are different. The present invention is characterized in that a fluctuation range control value that corresponds to a threshold value is set for each of the servomotors of the work machine in the basic data acquisition phase, and a plurality of torque data sets is introduced to a single checking unit 66. Faults are accordingly monitored. As a result, a plurality of servo motors can be centrally controlled by a single checking unit. When the servo motors are centrally controlled, the motors are more readily controlled, a single fault monitoring system will be sufficient, and space for installing the system can be reduced. There are no limitations regarding the type of work machine in which the present invention is employed, provided that the machine is equipped with a servo motor and a power transmission mechanism for transmitting the output of the servo motor. The power transmission mechanism may be a reducer, belt, gear, link, chain, or drive shaft, with no limitation placed on the type of mechanism. The present invention is useful in monitoring faults in a work machine comprising a servo motor and a reducer or another power transmitting mechanism. |
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056429550 | abstract | A strongback for lowering a tie rod into the downcomer annulus of a boiling water reactor during a shroud repair operation. The tie rod strongback is suspended from a cable via a cable adaptor at its upper end. The lower end of the strongback is coupled to a tie rod adaptor, which in turn couples to the top of the tie rod. The strongback is a welded assembly of square tubes, channels for reinforcing the joints of the welded tubes, and upper and lower couplings. In particular, the strongback has mutually parallel first and second rigid linear members which are disposed vertically when the strongback is suspended from a plumb cable. The second rigid linear member is connected to the first rigid linear member by a relatively obliquely disposed third rigid linear member. The first and second rigid linear members lie in a vertical plane which is offset from the axis of a plumb cable to allow the strongback assembly to circumvent the core spray downcomer piping when the tie rod/lower spring assembly is in its final position in the annulus. The first rigid linear member is further offset from the second rigid linear member cable axis to allow the strongback assembly to circumvent the feedwater sparger and the core spray header. This facilitates proper positioning of the bottom of the tie rod/lower spring assembly relative to the gusset plate to which the assembly will be anchored. |
abstract | The ultrasonic probe of the ultrasonic inspection apparatus, which is pushed onto the outer surface of the reactor pressure vessel, transmits and receives an ultrasonic wave to and from a penetration having a welded portion while changing an incident angle of the ultrasonic wave. Based on a result of reception of an echo obtained by the reflection of the ultrasonic wave on the inner surface of the penetration, an inclination angle of the penetration relative to a wall surface of the reactor pressure vessel is measured. A circumferential direction position of the penetration, which corresponds to the inclination angle, is calculated based on the relationship of an inclination angle and a circumferential direction position, which have been calculated in advance. Then, the circumferential direction position can be obtained as information on the inspection position. |
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046817298 | claims | 1. A vessel provided internally with a device for enabling external detection of temperature changes within the vessel, said device comprising a permanent magnet, means mounting the magnet within the vessel for movement towards and away from a boundary wall of the vessel, means having sufficient magnetic coupling with the magnet to restrain the magnet against movement away from a predetermined position with espect to said boundary wall of the vessel, and temperature-responsive means for urging the magnet away from said predetermined position such that, in the event of a substantial rise in temperature within the vessel, the latter means becomes effective to move the magnet relative to the boundary wall to vary the magnetic field strength detectable by detector means located on the external side of said boundary wall. 2. A vessel as claimed in claim 1 in which said vessel boundary wall is composed of a magnetically soft material and constitutes said restraining means, and in which the urging means develops sufficient power to move the magnet away from said boundary wall in the event of a substantial rise in internal temperature. 3. A vessel as claimed in claim 1 in which a magnetically soft element determines said predetermined position, said element being fixed against movement with respect to said boundary wall and constituting said restraining means, and in which the urging means develops sufficient power to move the magnet away from said element in the event of a substantial rise in internal temperature. 4. A vessel as claimed in claim 1 in which said urging means includes a thermal link which, while intact, restrains the urging means from exerting its urging force on the magnet but breaks in response to substantial temperature rise and thereby renders the uring means effective to move the magnet away from said predetermined position. 5. A vessel as claimed in claim 1 in which said urging means includes a bimetallic component. 6. A vessel as claimed in claim 1 in which said boundary wall is constituted by a top wall portion of the vessel. 7. A handling and monitoring facility for spent nuclear fuel material comprising at least one vessel as claimed in claim 1, and detector means external to the vessel and responsive to changes in magnetic flux caused by movement of said permanent magnet away from said predetermined position. 8. A facility as claimed in claim 7 in which said detector means comprises a Hall effect transducer. 9. A facility as claimed in claim 7 in which said detector means is operable to induce a time-varying magnetic flux in said vessel wall for superimposition on the steady state magnetic flux of the permanent magnet in said boundary wall or said element and to measure changes resulting from movement of the magnet away from said wall. 10. A facility as claimed in claim 7 in which a plurality of said vessels are provided, said vessels each incorporating a device as aforesaid and being arranged in an array such that said vessel boundary walls are all accessible at one face of the array, and in which means is provided for moving the detector means from one vessel to another whereby all of the vessels can be periodically monitored. 11. A facility as claimed in claim 10 in which said moving means is operable to bring the detector means into close proximity with said boundary wall of each vessel and scan the detector means over said boundary wall. |
044029039 | description | DESCRIPTION OF THE PREFERRED EMBODIMENT The control system of the present invention can be best understood by reference to the exemplary embodiments seen in the drawings. In FIG. 1, the control system 10 is associated with a nuclear power plant steam supply system 12 represented by the dotted line block. A plurality of sensors 14a and 14b are disposed within the nuclear power plant steam supply system 12. The sensors 14a and 14b are redundant sensors for a specific reactor parameter, such as radiation level, coolant temperature, steam generator pressure, with the appropriate sensor for the parameter to be measured. Thus, sensors 14a and 14b would be radiation detectors for neutron or gamma radiation measurement. An electrical instrumentation cable 16a and 16b is connected between respective sensors and the control system 10. The control system 10 in the FIG. 1 embodiment includes two identical redundant channel means 18a and 18b, which each comprise a signal processing means 20a, 20b connected via cable 16a, 16b to the respective sensor 14a or 14b. The channel means 18a and 18b each includes an optical logic unit interface 22a and 22b and an optical logic unit 24a, 24b. The optical logic units 24a and 24b are connected by a fiber optic cable link 26. The control system 10 includes a light source 28 which is inputed by fiber optic cable 30 to optical logic unit 24a. A light detector means 32 is connected to the output of optic logic unit 24b via fiber optic cable 34. The light source 28 and detector means 32 are connected to permit signal synchronization and provision of a reference signal for the detected light. The optical logic unit 24a, 24b, as seen in greater detail in FIG. 2, comprise electro-optic modulatable member 36 which is illustrated as an elongated member with its optical axis aligned with the input cable 30, cable link 26, and output cable 34. The electro-optic modulatable member 36 is formed of a selected light transmissive material which has the property that an applied electric potential orthogonal to the optical transmissive axis is effective to modulate the light beam as it passes through the member 26. For a member 36 which modulates light intensity, an input lens 38 and polarizer 40 are aligned along the optic axis between cable 30 and member 36, with output polarizer 42 and lens 44 aligned along this same axis between member 36 and cable 26. A potential V.sub.A is applied across member 36 establishing an electric field orthogonal to the light transmission axis as is well known to modulate the intensity of light passing through member 36. The input polarizer 40 has its polarization axis at an angle of 45 degrees with respect to the slow transmission axis of the member 36. The fast transmissive axis of member 36 is aligned with the optical axis through lens 38 and 42 as well as polarizer 40 and 42. The slow transmissive axis is orthogonal to the first transmissive axis. Light from input polarizer 40 is linearly polarizer. The polarization axis of the output polarizer 42 is offset 90.degree. from that of the input polarizer 40. The modulating potential V.sub.A applied across electro-optic modulatable member 36 causes a phase differential for the linearly polarized light passing through member 36 varying from 0 to .pi. radians. This causes the polarization of the light to change from linear at 0 radians, to circular at .pi./2 radians, to linear normal relative to the input polarizer at .pi. radians. Thus, the intensity of light passing through the output polarizer, which has its polarization axis 90.degree. offset from that of the input polarizer, varies from 0 to 100% of the input intensity as the phase varies from 0 to .pi. radians. The relationship of input and output light intensity for the system is defined by the equation: EQU (I.sub.o /I.sub.i =1/2(1- cos .phi.) where I.sub.o is the output intensity, I.sub.i is input intensity, and .phi. is the differential phase shift between the fast and slow transmission axis for the system. The ratio of I.sub.o to I.sub.i is thus 0 at 0 radians for .phi., is 0.5 at .pi./2 radians, and is 1 for .pi. radians. The second optical logic unit 24b is essentially identical to unit 24a with the fiber optic cable link 26 serially connecting them along a common optical axis. This fiber optic cable link 26 serves to couple the redundant logic channels and maintain physical and electrical isolation between the units. The logic table below helps to explain operation of the system. LOGIC TABLE ______________________________________ V.sub.A V.sub.B Intensity ______________________________________ 0 0 1/4 0 1 1/2 1 0 1/2 1 1 1 ______________________________________ The light intensity which the detector senses is the product of the light output intensities of the optical logic units 24a, 24b. The modulating potentials V.sub.A and V.sub.B are applied respectively to units 24a, 24b, and are a function of the respective sensor 14a, 14b condition. The logic table shows that when V.sub.A is 0, and V.sub.B is 0, the light intensity sensed by detector 32 is 1/4 of the light input intensity from light source 28. This results from the fact that when V.sub.A is 0, the output intensity from unit 24a is 1/2, and since this is the light input for unit 24b, when V.sub.B is also 0, the output intensity from unit 24b is 1/2x1/2=1/4. When either V.sub.A or V.sub.B is 0, and the other is 1, the intensity at the detector will be 1/2. Only when the values of V.sub.A and V.sub.B are each 1 will the intensity be 1. These values can be used to denote power plant conditions. Thus, when sensor 14a, 14b are reading normal operating conditions the potentials V.sub.A and V.sub.B could both be 1, and the intensity value 1 will indicate normal operating condition. When either of the sensors 14a or 14b provides a signal which is translated to a value of V.sub.A, V.sub.B which is different from that of the other sensor this implies a fault in the system. The fault is a warning to check the sensor or compare other parameters to determine if the power plant is in fact operating normally or whether an operating parameter is outside of normal operating tolerances as to require operator correction or intervention. The intensity level of 1/4 which is had when the sensor 14a, 14b both indicate the same parameter condition, and with potentials V.sub.A and V.sub.B of 0 applied to the optical logic units would indicate a definite out of tolerance system parameter and result in an automatic trip or shutdown control signal being generated. When these redundant channels both indicate the same condition the control system is actuated to provide continued safe operator control. The system described is a two channel system but can easily be expanded to the more commonly used three channel system used in many nuclear power plant control systems. |
summary | ||
abstract | An image processing apparatus includes an acquiring unit and an image processing unit. The acquiring unit acquires a radiographic image acquired by irradiating a radiation on an object and a scattered-ray component contained in the radiographic image, wherein the scattered-ray component originates from a scattered ray which is a radiation scattered in the object. The image processing unit performs an image process on the radiographic image in accordance with an instruction, wherein the image processing unit performs an image process based on a radiographic image acquired by the acquiring unit and a scattered-ray component acquired by the acquiring unit in a case where an instruction to perform the image process is received again. |
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047864613 | claims | 1. A nuclear reactor comprising: a reactor pressure vessel; a lower-internals assembly disposed within the reactor pressure vessel including a core barrel having a flange with a plurality of first annularly-spaced coolant passages; an upper-internals assembly disposed within the reactor pressure vessel including an upper-internals barrel having a flange, axially disposed above said core-barrel flange, with a plurality of second annularly-spaced coolant passages; each of said plurality of second passages being positioned to be in coolant-passage communication with a corresponding one of said plurality of said first passages; a plurality of reactor-internals hold-down spring assemblies annularly-spaced about said core-barrel flange, each of said spring assemblies being disposed between said core-barrel flange and said upper-internals barrel flange and being associated with one of said plurality of second passages and said corresponding one of said plurality of first passages; each said spring assembly comprising: a resilient biasing device; a retainer means for carrying said resilient biasing device and having therein a central bore; and means, disposed in said central bore, for defining a connecting flow passage for coolant between a said second and said corresponding first coolant passages with which said each spring assembly is associated. a reactor pressure vessel, a lower-internals assembly disposed within the reactor pressure vessel including a core barrel having a flange with a plurality of first annularly-spaced coolant passages; an upper-internals assembly disposed within the reactor pressure vessel including an upper-internals barrel having a flange, axially disposed above said core-barrel flange, with a plurality of second annularly-spaced coolant passages; each of said plurality of second passages being positioned to be in coolant-passage communication with a corresponding one of said plurality of said first passages; a plurality of reactor-internals hold-down spring assemblies annularly-spaced about said core-barrel flange, each of said spring assemblies being disposed between said core-barrel flange and said upper-internals barrel flange and being associated with one of said plurality of second passages and said corresponding one of said plurality of first passages; each said spring assembly comprising: a resilient biasing device; a retainer means for carrying said resilient biasing device and having therein a central bore; and means disposed in said central bore, for defining a connecting flow passage for coolant between a said second and said corresponding first coolant passages with which said each spring assembly is associated; said defining means comprising: a bellows flange, fixed at one end of said central bore, and having a central opening therein; a spring bellows carried by said bellows flange and disposed within said central bore; a movable plunger carried by said spring bellows and having a central opening therein; said plunger being biased by said spring bellows against the one of said plurality of first coolant passages with which it is associated. a reactor pressure vessel; a lower-internals assembly disposed within the reactor pressure vessel including a core barrel having a flange with a plurality of first annularly-spaced coolant passages; an upper-internals assembly disposed within the reactor pressure vessel including an upper-internals barrel having a flange, axially disposed above said core-barrel flange, with a plurality of second annularly-spaced coolant passages; each of said plurality of second passages being positioned to be in coolant-passage communication with a corresponding one of said plurality of said first passages; a plurality of reactor-internals hold-down spring assemblies annularly-spaced about said core-barrel flange, each of said spring assemblies being disposed between said core-barrel flange and said upper-internals barrel flange and being associated with one of said plurality of second passages and said corresponding one of said plurality of first passages; each said spring assembly comprising: a resilient biasing device; a retainer means for carrying said resilient biasing device and having therein a central bore; and means disposed in said central bore, for defining a connecting flow passage for coolant between a said second and said corresponding first coolant passages with which said each spring assembly is associated; said defining means comprising: a first movable plunger, movably retained within one end of said central bore and having a central opening therein; a second movable plunger movably disposed within another end of said central bore, and having a central opening therein; and a spring bellows disposed within said central bore and between said first and second plungers for biasing said first and second plungers against the second coolant passageway associated with said defining means and the corresponding one of the first coolant passageways in communication with said second coolant passages. a reactor pressure vessel; a lower-internals assembly disposed within said reactor pressure vessel, said assembly including a core barrel having a flange with a plurality of first annularly-spaced coolant passages; an upper-internals assembly disposed within said reactor pressure vessel, said upper-internals assembly including an upper-internals barrel having a flange, axially disposed above said core-barrel flange, with a plurality of second annularly-spaced coolant passages; each of said second passages being positioned to be in coolant passage communication with a corresponding one of said first passages; a head secured to said vessel defining an upper-head region between it and said upper-internals assembly with said second passages in communication with said upper-head region; a reactor hold-down spring assembly disposed between said core-barrel flange and said upper-internals barrel flange at each associated first and corresponding second passages between said flanges; each said hold-down spring assembly having a bore in coolant communication between each associated first passage and the corresponding second passage; an inlet nozzle in said pressure vessel; and means, connected to said inlet nozzle, for conducting the coolant flowing into said inlet nozzle to the upper-head region through said first and second passages and through said bore. 2. The nuclear reactor according to claim 1, wherein said resilient biasing device comprises a stack of Belleville springs carried on said retainer means. 3. The nuclear reactor according to claim 1, wherein the core barrel and pressure vessel form an inlet coolant flow annulus in fluid communication with the plurality of first annularly-spaced coolant passages, and wherein the upper-internals barrel and pressure vessel form an upper-head region in fluid communication with the plurality of second annularly-spaced coolant passages whereby coolant from said inlet coolant flow annulus flows through said plurality of first passages, the connecting flow passages, said plurality of second passages, and into the upper-head region. 4. A nuclear reactor comprising: 5. The nuclear reactor according to claim 4 wherein the biasing device comprises a stack of Belleville springs and, wherein the retainer means has an upper flange seated against the upper-internals flange, said upper flange being dimensioned to retain a first side of said stack of Belleville springs, and a second side of said stack being seated against said core-barrel flange to resiliently support said upper-internals barrel. 6. The nuclear reactor according to claim 4, wherein the plunger has a generally spherical end and the one of the first plurality of coolant passages with which the defining means of which the plunger forms a part is associated has a generally cone shaped seating surface, said spherical end being biased against said seating surface by the spring bellows. 7. The nuclear reactor according to claim 4, wherein said bellows flange has a central tube disposed within said spring bellows extending towards said plunger. 8. The nuclear reactor according to claim 5 further including a locking nut connected to and cooperative with the retainer to preload the stack of springs. 9. The nuclear reactor according to claim 5, wherein a plurality of annularly-spaced counter bores are formed in the core-barrel flange about each of first coolant passages, each of said counter bores having a shoulder for supporting said stack of Belleville springs, said counter bore being dimensioned so that a gap is formed between the core-barrel flange and the upper-internals flange when the retainer flange is seated against said upper-internals flange, said gap defining the maximum allowable deflection of said Belleville spring stack. 10. A nuclear reactor comprising: 11. The nuclear reactor according to claim 10 wherein the biasing device comprises a stack of Belleville springs and, wherein the retainer has an upper flange to be seated against the upper-internals flange, said upper flange being dimensioned to retain a first side of said stack of Belleville springs, a second side of said stack being seated against said core-barrel flange to resiliently support said support internals barrel. 12. The nuclear reactor according to claim 10, wherein the first and second plungers have first and second generally spherical ends respectively and the associated second coolant passage and the first passage in communication therewith have generally cone shaped seating surfaces, said first and second spherical ends being biased against said second and first seating surfaces respectively by the spring bellows. 13. The nuclear reactor according to claim 11, wherein the first plunger has a central tube disposed within the spring bellows, said tube extending toward the second plunger. 14. The nuclear reactor according to claim 11 further including a locking nut connected to and cooperative with the retainer to preload the stack of springs. 15. A nuclear reactor comprising: |
063273241 | abstract | A fuel assembly for a boiling water reactor comprising a plurality of fuel units (3a, 3b, 3c, 3d), stacked on top of each other, each one comprising a top tie plate (5), a bottom tie plate (6) and a plurality of fuel rods (4a, 4b, 4c) arranged between the top tie plate and the bottom tie plate. The fuel units are surrounded by a fuel channel (9) with a substantially square cross section. At least some of the fuel units comprise fuel rods with different diameters and different fuel quantities. The fuel rods are adapted such that fuel quantity and lattice space are optimized laterally and axially in the fuel assembly. |
description | This application claims the benefit of Korean Patent Application No. 10-2010-0138442, filed on Dec. 30, 2010, in the Korean Intellectual Property Office, the disclosure of which is incorporated herein in its entirety by reference. 1. Field of the Invention The present invention relates to a system of controlling a level of main feed-water in a steam generator of a nuclear power plant during a main feed-water control valve transfer to prevent an excessive level of the main feed-water in the steam generator. 2. Description of the Related Art A nuclear power plant generally includes 100 or more systems respectively having individual functions. The systems are greatly classified into a nuclear steam supply system (NSSS) which is based on a nuclear reactor, a turbine/generator system which is supplied with steam to operate a generator so as to produce electricity, and other subordinate facilities. A pressurized water reactor (PWR), which is currently a main part of the nuclear power plant, includes a primary system which is based on a reactor, a secondary system which includes a steam generator, a turbine, an electric generator, and a condenser, an engineered safeguard system which is prepared for accidents, a power transmission and supply system, an instrumentation and control system, other subsidiary systems. Hot water generated in a reactor circulates through a heat transfer pipe of a steam generator, which is connected to the reactor through a coolant pipe, to transfer heat to feed-water flowing into the steam generator through another pipe and then returns to the reactor. For the easy performance of this function of the steam generator, a level of the feed-water in the steam generator should be appropriately maintained. A feed-water control system of a nuclear power plant uniformly controls the level of the feed-water in the steam generator. The present invention provides a system of controlling main feed-water by which a gain and an integral time constant of a proportional-integral (PI) controller are transformed according to a timer lapse time to relieve an excessive level of feed-water in a steam generator caused by sudden changes in open degrees of a downcomer feed-water valve and an economizer feed-water valve when reactor power transfers between low and high power sections. According to an aspect of the present invention, there is provided a system for controlling a main feed-water pump and a main feed-water control valve including a downcomer feed-water valve and an economizer feed-water valve to control a steam generator level of a nuclear power plant, including: a proportional-integral (PI) controller which performs a PI operation on a compensated steam generator error signal generated according to a deviation between the steam generator level and level setpoints of the steam generator and generates a flow demand signal according to the result of the PI operation; a power determiner which determines a power section to which reactor power belongs, among low and high power sections and determines a transfer time when the determined power section transfers; a main feed-water controller which generates a main feed-water control signal for controlling the main feed-water pump and the main feed-water control valve according to the flow demand signal and the determined power section; an information provider which provides PI information including a gain and an integral time constant set according to the reactor power; and an information transformer which provides the PI information to the PI controller and provides transformed PI information comprising a transformed gain and a transformed integral time constant to the PI controller only for a predetermined timer time at the transfer time. The information transformer may include: a transformation information provider which provides the transformed gain and the transformed integral time constant to the PI controller; and a first timer which provides the PI information to the PI controller and provides the transformed PI information to the PI controller only for a predetermined first timer time at a first transfer time when the transfer time transfers from the low power section to the high power section. The transformed gain may be set to a value greater than a gain set according to the reactor power, and the transformed integral time constant may be set to a value smaller than an integral time constant set according to the reactor power. The transformed gain and the transformed integral time constant may vary with the timer time. The present invention will now be described more fully with reference to the accompanying drawings, in which exemplary embodiments of the invention are shown. Like reference numerals in the drawings denote like elements, and thus their description will be omitted. The present invention provides a system and a method of controlling an excessive level of feed-water in a steam generator of a nuclear power plant during a main feed-water control valve transfer. FIG. 1 is a view illustrating a system 100 for controlling a level of main feed-water in a steam generator of a nuclear power plant according to an embodiment of the present invention. Referring to FIG. 1, the system 100 includes a proportional-integral (PI) controller 110, an information provider 120, an information transformer 130, a power determiner 140, and a main feed-water controller 150. The PI controller 110 performs a PI control with respect to a compensated steam generator level error signal of a steam generator and generates a flow demand signal according to the result of the PI control. As will be described later, a proportional gain and an integral time constant used by the PI controller 110 are provided from the information provider 120 and the information transformer 130. The system 100 of the nuclear power plant may further include an error signal provider 111 which is installed in front of the PI controller 110 and provides the compensated steam generator level error signal. The error signal provider 111 receives a difference between a level signal corresponding to a steam generator level and a level setpoint signal corresponding to a steam generator level setpoint and provides the compensated steam generator level error signal, which is the difference between the level signal and the level setpoint signal, to the PI controller 110. The flow demand signal generated by the PI controller 110 is provided to the main feed-water controller 150. The main feed-water controller 150 generates a main feed-water control signal for controlling a main feed-water pump and a main feed-water control valve based on the flow demand signal. The main feed-water control signal includes at least one of a pump speed demand signal, a downcomer valve position demand signal, and an economizer valve position demand signal. In the present invention, the main feed-water control valve includes a downcomer feed-water valve and an economizer feed-water valve. The downcomer feed-water valve is a valve which controls a position and an open degree of the downcomer feed-water valve. The economizer feed-water valve is a valve which controls a position and an open degree of the economizer feed-water valve. Therefore, the main feed-water controller 150 generates the pump speed demand signal, the downcomer valve position demand signal, and the economizer valve position demand signal based on the flow demand signal. For this purpose, the main feed-water controller 150 includes a pump speed controller 160 which controls the speed of the main feed-water pump, a downcomer valve controller 170 which controls the position of the downcomer feed-water valve, and an economizer valve controller 180 which controls the position of the economizer feed-water valve. The pump speed controller 160 generates the pump speed demand signal for controlling the speed of the man feed-water pump. The downcomer valve controller 170 generates the downcomer valve position demand signal for controlling the open degree of the downcomer feed-water valve. The economizer valve controller 180 generates the economizer valve position demand signal for controlling the open degree of the economizer feed-water valve. The speed of the main feed-water pump, the position of the downcomer feed-water valve, and the position of the economizer feed-water valve are respectively controlled by the pump speed demand signal, the downcomer valve position demand signal, and the economizer valve position demand signal, thereby controlling the level of the feed-water in the steam generator. The main feed-water controller 150 changes its control method according to a power section of reactor power and a transfer time of the power section. A reference point of the power section and the transfer time of the reactor power is a point at which the reactor power is about 20%. If the reactor power belongs to a low power section of about 20% or less, the main feed-water controller 150 controls the main feed-water control valve in a low power control mode. If the reactor power belongs to a high power section of about 20% or more, the main feed-water controller 150 controls the main feed-water control valve in a high power control mode. The power determiner 140 receives a reactor power signal corresponding to the reactor power and determines a power section to which the reactor power belongs and the transfer time when the power section of the reactor power transfers. A power section signal of the power section of the reactor power determined by the power determiner 140 is provided to the main feed-water controller 150. A transfer signal generated at the transfer time when the power section of the reactor power transfers is provided to the information transformer 130. As described above, in the present invention, a control method of the main feed-water controller 150 is changed according to the reactor power. In other words, a flow and a distribution ratio of feed-water supplied through a downcomer and an economizer are controlled according to the reactor power. Since the main feed-water pump operates at the lowest speed in the low power section in which the reactor power is about 20% or less, an actual feed-water flow control function in the low power section is performed by the downcomer feed-water valve. In other words, in the low power section in which the reactor power is about 20% or less from a start operation, all amount of feed-water is supplied to the steam generator only through a downcomer feed-water nozzle. In the high power section in which the reactor power is about 20% or more, the position of the downcomer feed-water valve is fixed to a position in which a feed-water flow corresponding to about 10% of a total feed-water flow (a feed-water flow if the reactor power is 100%) can pass. Also, the position of the economizer feed-water valve is controlled to supply a remaining feed-water flow to the steam generator through the economizer. Accordingly, the downcomer feed-water valve and the economizer feed-water valve are controlled to be automatically transferred at a point at which the reactor power is about 20%. If the reactor power reaches a point at which the reactor power decreases and is about 20%, the economizer feed-water valve is controlled to be completely closed and to supply feed-water to the steam generator only through the downcomer. Also, if the reactor power reaches a point at which the reactor power increases and is about 20%, the position of the downcomer feed-water valve is rapidly changed to a valve position corresponding to about 10% of the feed-water flow required during a 100% power operation. Simultaneously with this, the economizer feed-water valve is controlled to supply the remaining feed-water flow to the steam generator. A reference point for dividing the power section of the reactor power is not limited to the above-described reference point but may be determined using various methods. The reference point for dividing the power section of the reactor power may be set to 20% as described above. However, as in the following example, a reference point when the reactor power increases and a reference point when the reactor power decreases may be differently set. For example, if the reactor power increases, a reference point for dividing the power section of the reactor power into low and high power sections may be set to a point when the reactor power is about 20%. If the reactor power decreases, the reference point may be set to a point when the reactor power is about 18%. In this case, if the reactor power increases, the point when the reactor power is about 20% is a transfer time when the low power section transfers to the high power section. If the reactor power decreases, the point when the reactor power is about 18% is a transfer time when the high power section transfers to the low power section. FIG. 2 is a view illustrating the power determiner 140 in detail. Referring to FIG. 2, the power determiner 140 includes a power comparator 141 and a transfer time determiner 143. The power comparator 141 receives a reactor power signal corresponding to the reactor power, determines a power section of current reactor power based on the reactor power signal, and generates a power section signal of the determined power section. The power section of the reactor power means one of low and high power sections to which the current reactor power belongs. If the reactor power is lower than about 20% based on a point of about 20%, the power comparator 141 determines the power section of the reactor power as the low power section, generates a low power section signal TL, and provides the low power section signal TL to the economizer valve controller 180. If the reactor power is higher than or equal to 20% based on the point of about 20%, the power comparator 141 determines the power section of the reactor power as the high power section, generates a high power section signal TH, and provides the high power section signal TH to the downcomer valve controller 170. The transfer time determiner 143 generates a first transfer signal TLH at a first transfer time when the reactor power transfers from the low power section to the high power section and provides the first transfer signal TLH to the information transformer 130. The transfer time determiner 143 also generates a second transfer signal THL at a second transfer time when the reactor power transfers from the high power section to the low power section and provides the second transfer signal THL to the information transformer 130. A reference point for dividing the power section of the reactor power into the low and high power sections is a point when the reactor power is about 20%. However, as described above, the reference point may be differently set when the reactor power increases and when the reactor power decreases. The description will return to FIG. 1. The pump speed controller 160 receives the flow demand signal and generates the pump speed demand signal for controlling the speed of the main feed-water pump based on the flow demand signal. The pump speed demand signal is provided to the main feed-water pump to be used to control a feed-water speed of the main feed-water pump. The downcomer valve controller 170 receives the flow demand signal and the high power section signal TH and generates the downcomer valve position demand signal for controlling the position of the downcomer feed-water valve based on the flow demand signal and the high power section signal TH. The downcomer feed-water valve controls the position thereof according to the downcomer valve position demand signal to control the open degree thereof. The downcomer valve controller 170 includes a downcomer distributer 171, a first transfer logic 172, and a downcomer fixer 173. The downcomer distributer 171 calculates a feed-water flow, which is supplied through the downcomer, according to the flow demand signal. If the reactor power is in the high power section, the downcomer fixer 173 provides a bias signal to the first transfer logic 172. The bias signal refers to a signal which is to fix the position of the downcomer feed-water valve to supply the feed-water flow corresponding to about 10% of the total feed-water flow (the feed-water flow if the reactor power is 100%) through the downcomer. If the first transfer logic 172 receives the high power section signal TH from the power determiner 140, the first transfer logic 172 generates the downcomer valve position demand signal which is to fix the position of the downcomer feed-water valve to supply the feed-water flow corresponding to about 10% of the feed-water flow during the 100% power operation according to the bias signal of the downcomer fixer 173. If the first transfer logic 172 does not receive the high power section signal TH, the first transfer logic 172 generates the downcomer valve position demand signal to supply feed-water through the downcomer according to the flow demand signal. Accordingly, in the low power section, the downcomer valve controller 170 controls the total feed-water flow to be supplied through the downcomer according to the flow demand signal. In the high power section, the downcomer valve controller 170 controls a feed-water flow corresponding to about 10% of the total feed-water flow during the 100% power operation to be supplied through the downcomer. The economizer valve controller 180 includes an economizer distributer 181, a second transfer logic 182, and a closed fixer 183. The economizer distributer 181 calculates the feed-water flow supplied through the economizer according to the flow demand signal. If the reactor power is in the low power section, the closed fixer 183 provides a closed signal to the second transfer logic 182. The closed signal refers to a signal which is to fix the position of the economizer feed-water valve to a closed position so that feed-water is not at all supplied through the economizer. If the second transfer logic 182 receives the low power section signal TL from the power determiner 140, the second transfer logic 182 generates the economizer valve position demand signal which is to fix the position of the economizer feed-water valve to close the economizer feed-water valve according to the closed signal. If the second transfer logic 182 does not receive the low power section signal TL, the second transfer logic 182 generates the economizer valve position demand signal to supply feed-water through the economizer according to the flow demand signal. Accordingly, in the low power section, the economizer valve controller 180 controls the economizer feed-water valve to be closed. In the high power section, the economizer valve controller 180 controls the remaining feed-water flow to be supplied through the economizer except for the feed-water flow of the total feed-water flow supplied through the downcomer. FIG. 3 is a table illustrating a ratio of feed-water which is distributed through the downcomer feed-water valve and the economizer feed-water valve according to the power section of the reactor. Referring to FIG. 3, a feed-water flow is distributed through the downcomer feed-water valve and the economizer feed-water valve according to the power section of the reactor. FIG. 4 is a view illustrating the information provider 120 and the information transformer 130 in detail. The information provider 120 provides PI information comprising the gain and the integral time constant, which are set according to the reactor power, to the PI controller 110. For this purpose, the information provider 120 receives the reactor power signal corresponding to the reactor power and provides information regarding the gain and the integral time constant, which are set based on the reactor power signal, to the PI controller 110. FIGS. 5 and 6 are graphs respectively illustrating the gain and the integral time constant set based on the reactor power signal. Referring to FIGS. 5 and 6, the gain and the integral time constant of the PI controller 110 are optimal setpoints for preventing several excessive phenomena which may occur in full power. In low power, the gain is low, and the integral time constant is relatively high. At the second transfer time when the reactor power transfers from the high power section to the low power section, the economizer feed-water valve is completely closed. Also, the downcomer feed-water valve, which is fixed to a valve position corresponding to a feed-water flow of about 10% of a demand flow during the 100% reactor power operation, rapidly rises to a valve position corresponding to a flow demand signal of 20% reactor power. Therefore, an excessive phenomenon in which the level of the steam generator rapidly increases may occur. FIG. 7 is a graph illustrating changes in the steam generator level at the second transfer time when the reactor power transfers from the high power section to the low power section. Referring to FIG. 7, an excessive phenomenon in which the steam generator level rises from 44% to 61%, i.e., by about 15%, at the second transfer time, occurs, wherein 44% corresponds to a normal level. Some nuclear power plants manually control downcomer feed-water valves due to an excessive rise of a level of a steam generator at a second transfer time to prevent a reactor from stopping due to a high level of a steam generator. Examples of manual operations increasing burden on an operator have been recently frequently reported. FIG. 8 is a graph illustrating changes in the steam generator level at the first transfer time when the reactor power transfers from the low power section to the high power section. At the first transfer time when the reactor power transfers from the low power section to the high power section, the downcomer feed-water valve rapidly transfers to the valve position corresponding to the feed-water flow of about 10% of the demand flow during the 100% reactor power operation, and the economizer feed-water valve rapidly moves to the valve position corresponding to the flow demand signal of the 20% reactor power. Since the downcomer feed-water valve rapidly transfers, the steam generator level decreases on an initial stage and then increases due to rapid opening of the economizer feed-water valve. Referring to FIG. 8, the excessive phenomenon in which the feed level of the steam generator decreases from 44% to 34%, i.e., by about 10%, at the first transfer time occurs, wherein 44% corresponds to a normal level of the steam generator. At the first transfer time, the open degree of the downcomer feed-water valve variously appears according to a pressure of the steam generator. Therefore, since the economizer feed-water valve rapidly opens after a valve transfer, the reactor may stop due to the high level of the steam generator. However, since the system 100 has a relatively slow response characteristic under the low gain and the high integral time constant of the PI controller 110, it may be difficult to relieve the excessive phenomenon of the steam generator level. Accordingly, if the main feed-water control valve transfers, it is necessary to prevent the excessive phenomenon of the steam generator level in order to relieve the burden on the operator due to the excessive phenomenon of the steam generator level and reduce the possibility of the stop of the reactor. The description will return to FIG. 4. The information transformer 130 provides transformed PI information, which includes information regarding a transformed gain and a transformed integral time constant, to the PI controller 110 only for a predetermined timer time at a transfer time of the power section of the reactor power. The information transformer 130 includes a timer 131 and a transformation information provider 132. When the timer 131 receives a transfer signal from the power determiner 140, the timer 131 provides the transformed PI information to the PI controller 110 only for the predetermined timer time. After the predetermined timer time has elapsed, the timer 131 provides the gain and the integral time constant, which are provided from the information provider 120, to the PI controller 110. Since the timer time and the transformed PI information may be differently set at the first and second transfer times, the timer 131 includes first and second timers 1311 and 1312. When the first timer 1311 receives a first transfer signal from the power determiner 140, the first timer 1311 provides the transformed PI information to the PI controller 110 only for a predetermined first timer time. After the predetermined first timer time has elapsed, the first timer 1311 provides the gain and the integral time constant, which are provided from the information provider 120, to the PI controller 110. When the second timer 1312 receives a second transfer signal from the power determiner 140, the second timer 1312 provides the transformed PI information to the PI controller 110 only for a predetermined second timer time. After the predetermined second timer time has elapsed, the second timer 1312 provides the gain and the integral time constant, which are provided from the information provider 120, to the PI controller 110. In other words, the first and second timers 1311 and 1312 provide the gain and the integral time constant, which are provided from the information provider 120, to the PI controller 110, and if the transfer signal is received, provide the transformed PI information to the PI controller 110 only for the predetermined timer time. The transformation information provider 132 provides the transformed gain and the transformed integral time constant. Since the transformed PI information may be differently set at the first and second transfer times, the transformation information provider 132 includes first and second transformation information providers 1321 and 1322. The gain and the integral time constant provided from the information provider 131 are values which are preset to vary according to the reactor power. However, the transformed gain and the transformed integral time constant provided from the transformation information provider 132 are values which are preset to vary with time not the reactor power. FIGS. 9 and 10 are graphs respectively illustrating the transformed gain and the transformed integral time constant with respect to a timer lapse time. Only when the timer 131 having a fixed constant (variable) time operates, the transformed gain and the transformed integral time constant shown in FIGS. 9 and 10 are provided to the PI controller 110. After the operation of the timer 131 is completed, the gain and the integral time constant provided from the information provider 120 are provided to the PI controller 110. This logic is applied only for a constant timer time of the timer 131 at the transfer time, and operation conditions of the timer 131 are limited only to a moment when the downcomer feed-water valve and the economizer feed-water valve reach transferred power (reactor power corresponding to about 20%). The transformed gain, the transformed integral time constant, and the timer time are not limited to the embodiment of the graphs of FIGS. 9 and 10 but may be appropriately set to control increase and decrease degrees of the steam generator level at the transfer time. FIGS. 11 and 12 are graphs respectively illustrating effects of the present invention at the first and second transfer times. In FIGS. 11 and 12, Case A is a graph illustrating changes in the feed-watwer level of the steam generator when the information provider 120 provides the PI information comprising the gain and the integral time constant to the PI controller 110. And Case B is a graph illustrating changes in the feed-watwer level of the steam generator when the information transformer 130 provides the transformed PI information comprising the transformed gain and the transformed integral time constant to the PI controller 110 only for a predetermined timer time at the transfer time. Referring to FIGS. 11 and 12, increases and decreases in the steam generator level at the first and second transfer times are rapidly coped with to relieve the excessive phenomenon of the steam generator level in the case B. Therefore, the possibility of the stop of the reactor caused by an excessive change in the steam generator level at the transfer time is reduced to relieve the burden on the operator and improve an operation rate and economic feasibility of the nuclear power plant. As described above, according to the present invention, an excessive phenomenon of a level of a steam generator is relieved at a transfer time of reactor power between low and high power sections. Therefore, a possibility of the stop of a reactor caused by the excessive phenomenon of the steam generator level is reduced to relieve burden on an operator and improve an operation rate and economic feasibility of a nuclear power plant. While the present invention has been particularly shown and described with reference to exemplary embodiments thereof, it will be understood by those of ordinary skill in the art that various changes in form and details may be made therein without departing from the spirit and scope of the present invention as defined by the following claims. |
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044118589 | description | DETAILED DESCRIPTION OF PREFERRED EMBODIMENT Referring now to the drawings in detail, FIG. 1 illustrates diametrically a local power monitoring system for the fuel assembly of a nuclear power reactor, the system being generally referred to by reference numeral 10. The raw data acquisition for the system is located in the reactor fuel core and consists of a plurality of local power rate sensors generally referred to by reference numeral 12 in FIG. 1. The signal outputs of the sensors are fed along two parallel paths through a direct analog signal processing line 14 and a precision computer line 16 in order to produce an averaged power readout 24 and a precision power readout 18, respectively. Data input to the precision computer is obtained from other sources, including for example, information calculated from sensor sensitivity models, power shock models, and various correction factors such as core condition and time domain corrections. In the processing path of computer 16, the sensor signal lines are individually biased by precalibration to obtain precision heat rate measurements from the sensor signals which are converted into local fuel power outputs corrected in accordance with various plant condition parameters from the data sources denoted by reference numeral 20 in FIG. 1. The precision power readout 18 so obtained may be fed to an analyzer 22 to provide fuel power failure forecasts and power distribution recommendations for failure avoidance purposes. The analyzer may alternatively receive its input from the averaged power readout 24 to which the signal processor 14 feeds its output in the form of local fuel power rate measurements. A calibrator 26 is connected to the signal processor 14 through which on-line correction of the processed signal output thereof may be effected by comparison of the readout 24 with the precision readout 18 while it is in operation. Thus the readout 24 may be operated continuously to provide the necessary information to the utility operator while computer 16 is non-functional in order to avoid power plant shut down because of interruptions in the supply of data to computer 16 for various reasons such as data updating. The type of sensor utilized in the power monitoring system of the present invention is very critical. As shown in FIG. 2, neutron flux sensors heretofore utilized for precision power monitoring purposes exhibited a significant change with time in signal level for a constant linear heat generation rate for a unit fuel rod length, as depicted by curve 28, assuming no emitter burn-out. With emitter burn-out compensation, the change in signal level for the neutron flux sensor is denoted by curve 30. In contrast thereto, the signal level change for a gamma sensor of the type disclosed in the aforementioned prior application is depicted by curve 32 in FIG. 2, requiring less drastic time domain correction. FIG. 5 illustrates one of the gamma sensors 34 extending through a guide tube 36 from a reactor installation to the seal flange connector 38 located at an instrument removal zone, of a pressure water reactor, for example. The sensor extends through the seal flange 40 and the thermocouple signal cables 42 thereof project through the seal plug 44 to the power monitoring hardware. Thus, gamma radiation produced by fission products in the reactor fuel assembly cause internal heating of the inner core 46 of the sensor to generate the signals in the thermocouple cables 42. While these signals provide for more accurate determination of linear heat generation rate because of its substantially direct relationship thereto, there is a signal response delay when a change in power occurs, as exhibited by the signal characteristic curve 48 shown in FIG. 3. In accordance with the present invention, the signal is modified to compensate for such slow signal response as indicated by deconvoluted heating rate signal curve 50. Referring now to FIG. 4, the signal cables from each of the sensors 34 are shown connected to a terminal box 52 through which signals of millivolt level are fed to a scanner or multiplexer 54. By way of example, eight sensors 34 are associated each fuel rod assembly of a reactor core and each sensor has two signal cables associated therewith to provide sixteen signal cables from each fuel rod assembly. In a light water reactor, between 350 to 450 of such signals are present to provide the local power rate measurements through the terminal box 52 to the scanner 54. The scanner may be a solid state multiplexer from which a signal sequence is fed to a first analog signal corrector 56 through which the signals are calibrated to provide a plurality of analog signals in signal path 58, representing local heat rates in the sensors. The signal path 58 represents a plurality of signal lines fed in parallel to the direct analog processing line in the precision signal processing line as aforementioned in connection with FIG. 1. In the precision processing line, the input analog signals enter a precision signal converter 60 through which the sensor signals are given in individual bias and corrected in accordance with a signal sensitivity model through calibrator 62 in order to obtain precision heat rate signals that are fed to a plant process computer 64 to which input data is also fed from model data storage 66 and plant condition data source 68. The signal output of the computer is then modified through a dynamic filter 70 to compensate for slow signal response as discussed with respect to FIG. 3. The signal output is then applied to the precision readout 24 in the form of a precision power display monitor furnishing local fuel power rate information for each fuel rod. Direct conversion of the signals in path 58 to local power rate information is effected through a second signal converter 72 to produce outputs reflecting the local fuel power rate for fuel rods adjacent to each of the sensors. The signal outputs are then modified by a dynamic filter 74 and passed through an extrapolator 76 to the continuous display monitor 24 as averaged power rate information. Signal correction may be effected through a calibrator 78. A substantially accurate readout is obtained by such direct signal processing only because of the more accurate signal information furnished by sensors 34 and the measures taken to compensate for slow signal response and on-line calibration through calibrator 28 from comparison with data obtained from the precision monitor 18. FIG. 6 illustrates a less complex version of the system insofar as the direct continuous signal processing path is concerned. The outputs of the sensors are fed through analog signal lines directly to calibrated voltmeters 82. The signals in these lines 80 are also fed to the contacts of a scanner 84 from which the signals are fed in sequence to the precision signal processing line 16 as hereinbefore described with respect to FIGS. 1 and 4. |
040070859 | abstract | A nuclear reactor fuel element has within its protective sheath an insert whereby it may be uniquely identified. The insert bears an array of markings at spaced positions to produce a series of clock pulses on presentation of the insert to a detector responsive to each marking and further markings at some only of the spaced positions to vary the series of signals produced on presentation of the insert to the detector, thereby providing numerical identification pulses. |
abstract | A nuclear reactor module includes a reactor vessel and a reactor housing mounted inside the reactor vessel, wherein the reactor housing comprises a shroud and a riser located above the shroud. The nuclear reactor module further includes a heat exchanger proximately located about the riser, and a reactor core located in the shroud. A steam generator by-pass system is configured to provide an auxiliary flow path of primary coolant to the reactor core to augment a primary flow path of the primary coolant out of the riser and into the shroud, wherein the auxiliary flow path of primary coolant exits the reactor housing without passing by the heat exchanger. |
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053012186 | description | DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS Referring to FIG. 1, the fuel element or pin 10 comprises an elongated body of fuel 12, such as a rod sometimes referred to as a pin or slug, containing a fissionable material in the form of metal or metal alloy. The fuel body 12 (or several aligned fuel bodies) is housed within a sealed, tube-like metal container 14, sometimes referred to as "fuel cladding". Cladding 14 isolates the fuel body 12 from the coolant which flows over the exterior surface of the fuel element 10 to transfer heat away therefrom, thereby protecting the fuel from reaction with or contamination from the coolant. The cladding also seals in fission products and precludes their escape into the coolant. The fuel body 12 is designed in configurational dimensions for radial expansion of about 25 to 30 vol. % due primarily to internally generated, fission-produced gases. Thus, the initially produced metal fuel units are of substantially smaller cross-sectional area than the internal cross-sectional area within the fuel cladding 14. As a result, an intermediate space 16 is formed between the exterior surface of the fuel body 12 and the interior surface of the fuel cladding 14. This initial intermediate space 16 is designed to accommodate the expansion of the fuel body 12 attributable to the gases produced during service and to protect the fuel cladding 14 from physical stress and possible rupture which would otherwise arise due to internal pressure of a confined body of expanding fuel. The intermediate space 16 of the fuel element 10 is initially filled during the fuel fabrication process with a liquid metal bonding material such as sodium, potassium, lithium or their alloys in solid form, which becomes molten at normal reactor operating temperatures and is displaced by the fuel as it expands. The bonding material enhances heat transfer from the fuel outward to the cladding while the fuel is expanding to fill the intermediate space. The fuel units of a typical fuel element for service in a liquid metal-cooled nuclear reactor are cylindrical bodies having a diameter of approximately 0.19 inch. The stainless steel cladding which surrounds such fuel units has a wall thickness of about 20 mils and an outside diameter of about 0.26 inch. The barrier in accordance with the invention is a multi-layer expandable body 18 of expendable alloying metal which is positioned between the fuel unit and fuel cladding. As a result, the metal alloy fuel interacts with the barrier material, not the stainless steel of the cladding. The barrier may be composed of any alloying metal having the property of increasing the melting temperature of metallic fuels, such as zirconium, titanium, niobium, molybdenum, vanadium, chromium and the like. Zirconium is a preferred alloying metal for the practice of the invention. In accordance with the preferred embodiment of the invention, the barrier is a multi-layer rolled foil 18 of zirconium or functionally equivalent metal. As best seen in FIG. 2, metal foil is rolled into the shape of a multi-layer tube and then laser tack-welded to hold the foil in the rolled state. The laser welding penetration is adjusted so that a foil weld 20 fuses only two or more of the outermost layers. At the weld point, the rolled foil must comprise at least one more layer than the number of foil layers penetrated by the laser weld. The weld 20 is designed to fail in response to fuel or blanket alloy swelling during irradiation. The fused material holding the outermost and one or more of the middle layers together fails at a level of stress which is less than the level of stress at which the material of said foil would rupture. After weld failure, the overlapping layers slip and the multi-layer foil unrolls, as shown in FIG. 3. For the purpose of illustration, FIG. 3 shows a failed weld area 22 in the outermost layer circumferentially shifted relative to a weld residue 20' still fused on the middle layer. (Obviously the positions of the failed weld area and the weld residue could be reversed.) If the barrier were comprised of only two layers at the point of tack welding, the relative movement of the two foil layers could give rise to a direct "line of sight" for radiation through the failed weld area 22 if the latter cleared the edge of the second layer. However, in accordance with the invention, a third foil layer is provided to assure that a barrier of at least two foil layers is presented along every " line of sight" from the fuel body 12 to the cladding 14. Thus, unrolling of the foil masks defects in the individual foil layers arising from small undetected manufacturing defects or weld failures, thereby providing a barrier which accommodates swelling associated with the metal fuel and blanket alloys. In the initial step, the tube is formed with at least 21/2 turns of the foil and then tack-welded. Laser weld development with 0.002-inch-thick pieces of zirconium foil revealed that the weld power could be adjusted so the weld penetrates the outer layers of foil in a stack without affecting the innermost layer. The geometric arrangement of cladding and barrier must satisfy the following relationship to assure that the foil movement masks the weld defects with one less foil layer than is initially provided at the weld point: EQU (C-2T-B).pi.+3S<L where C is the cladding diameter; T is the cladding thickness; B is the barrier diameter; S is the tack weld diameter; and L is the inside foil lap past the weld area. This expression assures that the inside foil layer laps the weld area enough that the failed weld areas in the unrolled or slipped foil barrier will not provide a direct "line of sight" between the cladding and fuel or blanket alloy. The preferred embodiment shown in FIG. 1 has been disclosed for the purpose of illustration only. It will be readily appreciated that the invention is not limited to a rolled foil barrier having only three foil layers at the point of tack welding. The barrier may have more than three foil layers at the point of tack welding provided that at least two but less than all of the foil layers are fused together by tack welding from the outermost side. |
052256855 | summary | BACKGROUND OF THE INVENTION This invention relates to encapsulated construction modules such as shielding panels, bricks and other the like including molded plastic material having an interior filled with a shielding media and method of making same. The prior art includes containers for housing radio active material which are thereafter wrapped with plastic material. Also bricks filled with lead pellets have been provided. Such soft bricks are beanbag designs filled with lead pellets. Such bricks are useful for packing around objects such as pipes, but they are not well suited to wall construction. Furthermore, lead pellets do not have the density of solid lead, so they offer reduced shielding. At joints between the soft bricks, radiation leakage is even more pronounced. Moreover, contact with shielding material such as lead should be avoided so that the molded plastic form protects against contact by the individual with the shielding material. The plastic encapsulating material also avoids contamination of the shielding material because the smooth surface is easily decontaminated and contact with liquid wastes, for example, is avoided. Accordingly, it is an important object of this invention to provide a molded encapsulating plastic form containing shielding material in order to facilitate manufacture and protect against contact with the shielding material while providing a surface which is easily decontaminated. SUMMARY OF THE INVENTION Depending on each user's needs, the shielding media utilized in the construction modules may be concrete, water, sand, or solid lead and the like. A special enlarged interfitting joint provides full shielding, and the use of an exterior as of molded polyethylene and the like over the shielding media avoids contamination of the shielding material while providing a surface which facilitates decontamination. For example, solid shielding bricks for construction of walls of various heights and lengths are provided of similar constructions. The shape of the bricks will improve shielding efficiency at the joints between bricks. The joints may be of chevron design, 45.degree. angle configuration providing the shielding media, usually solid lead, and will be encased in polyethylene polymer. A panel construction includes a molded polyethylene form filled with concrete. The use of a polyethylene exterior provides several benefits over existing designs. First, personnel are protected from contact with hazardous media such as lead, and second, it is possible to remove surface contaminants. The construction of the molded polymer shapes permit molten lead or concrete and the like to be poured into the molded plastic forms. |
description | This application is a Continuation of U.S. patent application Ser. No. 14/231,354 filed Mar. 31, 2014, which is a Continuation of U.S. patent application Ser. No. 13/893,056 filed May 13, 2013, which is a Continuation of U.S. patent application Ser. No. 12/280,079 filed Feb. 6, 2009, which is a U.S. National Phase Application under 35 U.S.C. §371 of International Patent Application No. PCT/EP2007/001424, filed Feb. 19, 2007, which claims priority to German Patent Application No. 102006008023.8, filed Feb. 21, 2006, each of which is herein incorporated by reference in its entirety. The present invention relates to a Method for purification of 225Ac from irradiated 226Ra-targets provided on a support according to claims 1 to 3. Furthermore, the invention relates to an 225Ac-containing radionuclide composition in accordance with claim 21. In particular, the radionuclide 225Ac can be successfully used in nuclear medicine—bound to tumorspecific antibodies—in various clinical trials in the treatment of cancer, particularly in form of its daughter nuclide 213Bi. Already in 1993, criteria for the selection of radionuclides for immunotherapy with α-emitters and β-emitters were provided for the first time (GEERLINGS, M. W. (1993): Int. J. Biol. Markers, 8, 180-186: “Radionuclides for radioiminunotherapy: criteria for selection”) where it turned out due to the difference in energy that the radioactivity of α-emitters to be applied may be more than 1000 times lower than that of β-emitters, if a comparable effect is to be achieved. Moreover, in the above literature, the α-emitting radionuclides 225Ac and its daughter isotope 213Bi turned out to be highly promising for the objects of radioimmunotherapy alongside the in principle usable, however relatively poorly available or instable antibody conjugate producing α-emitters: 211At, 255Fm, 212Bi/212Pb, 224Ra, 233Ra. One of the fundamental studies for the foundation of a radioimmunotherapy with αemitters is disclosed in GEERLINGS, M. W., KASPERSEN, F. M., APOSTOLIDIS; C. and VAN DER HOUT, R. (1993): Nuclear Medicine Communications 14, 121-125, “The feasibility 225Ac as a source of α-particles in radioimmunotherapy”. Here it is described that 225Ac produced from 229Th and the daughter isotope of 225Ac, namely 213Bi is suitable as isotope for the radioimmunotherapy with α-emitters. As indications there are described in particular cancer treatment and the treatment of micrometastases of malign tumors using tumor-specific monoclonal antibodies as carriers for α-emitters. A further study of KASPERSEN, F. M., BOS, E., DOORNMALEN, A. V., GEERLINGS, M. W., APOSTOLIDIS, C. and MOLINET, R. (1995): Nuclear Medicine Communications, 16, 468-476: “Cytotoxicity of 213Bi- and 225Ac-immunoconjugates” confirms and quantifies the cytotoxic effect of 213Bi and 225Ac with in vitro tests using the human epidermoid tumor cell line A431. Moreover, it is suggested to use 213Bi for the treatment of malignant diseases of the blood system. Further, in KASPERSEN et al. 1995 a process can be found with which antibodies can be bound chemically to a chelator suitable for 213Bi and 225Ac. It has proved that for example p-isothiocyanatobenzyl-diethylentriamine-pentaacetate (benzyl-DTPA) is particularly suitable. Another chelator, namely Cyclohexyl-DTPA is, for example, described in NIKULA, T. K., McDEVITT, M. R., FINN, R. D., WU, C., KOZAK, R.W., GARMESTANI, K., BRECHBIEL, M. W., CURCIO, M. J., PIPPIN, C. G., TIFFANY-JONES, L., GEERLINGS, M. W., Sr., APOSTOLIDIS, C., MOLINET, R., GEERLINGS, M. W., Jr., GANSOW, O. A. UND SCHEINBERG, D. A. (1999): J Nucl Med, 40, 166-176: “Alpha-Emitting Bismuth Cyclohexylbenzyl DTPA Constructs of Recombinant Humanized Anti-CD33 Antibodies: Pharmacokinetics, Bioactivity, Toxicity and Chemistry”. An overview over chelator chemistry can be found for example in HASSFJELL, S, and BRECHBIEL, W. (2001): Chem. Rev., 101, 2019-2036: “The Development of the α-Particle Emitting Radionuclides 212Bi and 213Bi, and Their Decay Chain Related Radionuclides, For Therapeutic Applications” In the meantime, various radioimmunotherapeutic approaches with 225Ac and 213Bi for the treatment of cancer are in various phases of clinical trials. The medical-clinical significance of the present invention may be seen for example from two promising therapeutic approaches: On the one hand, JURCIC, J. G., LARSON, S. M., SGOUROS, G., McDEVITT, M. R., FINN, R. D., DIVGI, C. R. Ase, M. B., HAMACHER, K. A., DANGSHE, M., HUMM, J. L., BRECHBIEL, M. W., MOLINET, R., SCHEINBERG, D. A. (2002) in Blood, 100, 1233-1239 report a significant success in the treatment of patients with acute myelogenous leukaemia (AML) and chronic myelogenous leukaemia (CML) by using 213Bi, which is bound to HuM195, a formulation of a monoclonal anti-CD33-antibody, which was developed for the humane medicine. This study was the first proof-of-concept where a human being was treated with a systemic radioimmunotherapy comprising an α-emitter, which is transported to a tumorspecific cellular target. On the other hand, HUBER, R., SEIDL, C., SCHMID, E, SEIDENSCHWANG, S., BECKER; K.-F., SCHUMACHER, C., APOSTOLIDIS, C., NIKULA, T., KREMMER, E., SCHWAIGER, M. and SENEKOWITSCE-SCHMIDTKE, R. (2003): Clinical Cancer Research (Suppl.) 9, 1s-6s: “Locoregional α-Radioimmunotherapy of Intraperitoneal Tumor Cell Dissemination Using a Tumor-specific Monoclonal Antibody” report the therapeutic effectivity of 213Bi-d9MAB—with low bone marrow toxicity—and the possible application of a locoregional therapy for patients who suffer from gastric carcinoma, who express d9-E-Cadherine. More results of studies and partial aspects in this matter are shown in: Roswitha HUBER, doctorate dissertation in the Faculty of Veterinary Medicine submitted to the Ludwig-Maximilians-University of Munich, Jul. 18, 2003: “Bewertung der lokoregionalen Radioimmuntherapie disseminierter Tumorzellen des diffusen Magenkarzinoms mit einem 213Bi gekoppelten tumorspezifischen Antikörper inn Mausmodell” (Evaluation of a locoregional radioimmunotherapy of disseminated tumor cells of the diffuse gastric carcinoma with a 213Bi bound tumor specific antibody in the mouse model). This dissertation was originated from Nuklearmedizinische Klinik and Poliklinik of the Technical University of Munich, the University hospital “Klinikum rechts der Isar”, dean: Prof. Dr. M. Schwaiger. The dissertation was prepared under the supervision of Prof. Dr. med. Dr. phil. Reingard Senekowitsch-Schmidtke and was presented to the veterinary faculty via Prof. Dr. med. vet. K. Tempel, Institute for Pharmacology, Toxicology and Pharmacy of the Faculty of Veterenary Medicine of the Ludwig-Maximillans-University of Munich, director: Prof. Dr. med. vet. R. Schulz. According to HUBER 2003, each year 18 out of 100 000 Germans come down with gastric carcinoma alone. In Japan, even 126 out of 100 000 people are affected. This means about 156 000 incidences per year in Japan alone. There, as well as in China, Taiwan and Korea, gastric carcinoma is one of the most frequent causes of death in consequence of a tumor. When a peritoneal carcinomatosis, the consequence of diffuse expansion of tumor cells in the abdominal cavity, is diagnosed, the life expectancy of a patient is at present about 12 months. Even with resectable gastric carcinoma, this means with carcinoma, which have not yet disseminated and with negative diagnostic findings with respect to lymph nodes, the relapse-free three-year survival-rate is at about 45%, only. Up to now the application of cytostatica within a chemotherapy seemed to be the most promising therapeutic way. However, the side effects range from immunosuppression, coagulopathy, metabolic anoxia, mucositis and hyperuricaemia to the danger of cytostatica induced secondary tumors. Particularly affected is here quickly proliferating tissue as bone marrow and the epithelium of the gastrointestinal tract as well as of the oral mucosa. The radioimmunotherapy, in contrast, uses protein structures located on the membrane, that are expressed by tumor cell lines in order to bind cytotoxic active substances via a carrier. Mostly, an overexpression of the binding molecule at the tumor cell is central to a radioimmunotherapy. The target molecule for the tumor associated antibodies is thus also expressed to a lower extend in physiologic cells of the organism. This implies that any therapeutic agent for radiotherapy also binds to these cells. Particularly, in the treatment of acute or chronic myelogenous leukaemia the meaning of the present invention takes effect, namely for the preparation of a suitable α-emitter, namely 225Ac which forms through decay reaction the bound, for example, to a tumorspecific antibody. The 213Bi atom decays via β-decay to 213Po, which releases its α-decay energy of 8,4 MeV with a half life of 4 μs in the tissue within a distance of 80 μm when decaying and thus kills effectively cells in its immediate neighborhood due to its high linear energy transfer. The so called locoregional application enables a quick binding of 213Bi bound tumor specific antibody to the tumor antigenes with maximal therapeutic success and minimal toxicity. Not before the late 80s was the α-emitting nuclide pair 213Bi/213Po was discovered for radioimmunotherapy. However, in the standard textbook of Schicha and Schober, 1997 “Nuklearmedizin-Basiswissen and klinische Anwendung” (nuclear medicine—basic knowledge and clinical application) it can be read: “The linear energy transfer of α-rays is so big that the likeliness for the creation of irradiation damages is bigger than a therapeutic effect. For this reason, nuclides, which release α-rays, are not applied in the nuclear medicine . . . ”, (“Der lineare Energietransfer ist bei α-Strahlen so groβ, daβ die Wahrscheinlichkeit für die Erzeugung von Strahlenschäden gröβer ist als ein therapeutischer Effekt. Aus diesem Grunde werden Nuklide, die α-Strahlen emittieren, in der Nuklearmedizin . . . nicht eingesetzt.”) However, in the clinical application of such α-emitters in combination with tumorspecific antibodies, exactly the opposite has proved to be true (cf. JURCIC et al. 2002). Consequently, the question arose which isotope it was best to use and how it could be prepared reliably and continuously. Most of the over hundred available α-emitters can already be excluded from in vivo application for practical reasons (cf. GEERLINGS 1993). These α-emitters have to meet requirements like sufficient chemical and physical purity, economic availability and an adequate half-life. The latter has to be long enough for binding to the antibodies and for the biologic allocation and has to be short enough so that the patient is not put at an unnecessary risk due to excessive exposition to the rays. One of the few α-emitter which fulfill these criteria is the nuclide pair 213Bi/213Po with a half-life of 45,6 min (213Bi). The photon emission of 213Bi with 440 KeV additionally permits an in vivo scintiscanning of the patient as well as an easy measurement of the activity using an α-ray counter. Moreover, in radiation protection it is important that the radiation can be detected easily. Furthermore, also traces of further daughter nuclides of 225Ac/213Bi as for example 221Fr or 209Pb can be determined by new methods of measurement and can also be included into the dosimetry alongside the quality control. In the meantime, 213Bi has become available via the production of 225Ac, for example according to EP 0 752 709 B1 and EP 0 962 942 A1 and particularly via the so called “thorium cow” according to U.S. Pat. No. 5,355,394. However, the production via the above-mentioned “thorium cow” is very expensive, as it derives from a neutron irradiation of 226Ra over several years, whereby finally among others an isotope mixture of 228Th and 229Th is assembled, whereby 229Th again decays via 225Ra into 225Ac, which decays to 213Bi. Thus, the mother-daughter nuclide pair 225Ac/213Bi is available in principle, however, neither in an adequate quantity and continuously nor at an acceptable price, however—as mentioned initially—first clinical studies with 225Ac/213Bi conjugated to HuM195, a humanized anti-CD33 monoclonal antibody are very successful against myeloid leukaemia. The first clinical phase I trials with 213Bi-HuM195 were carried out with excellent therapeutic results at leukaemia patients at the Memorial Sloan-Kettering Cancer Center in New York (JURICIC et at 2002). In cyclotrons, developed for the first time 1931, electrically charged particles are moving on spiral shaped orbits in magnetic flux lines. In particular, protons can be accelerated with the help of a cyclotron with currents that are high enough to such high velocities that they can be used in experimental and applied nuclear physics for the production of isotopes in a quantitative scale. EP 0 752 709 B1 describes, for example, a method for producing Actinium-225 from Radium-226, whereby accelerated protons are projected in a cyclotron onto a target of radium-226, characterized in that protons accelerated in a cyclotron are projected onto a target of radium-226 in a cyclotron, so that the instable compound nucleus 227Ac is transformed into Actinium-225 while emitting two neutrons (p,2n-reaction), whereby after a waiting period, during which the Actinium-226, which has been created simultaneously due to the emission of only one neutron, decays mostly due to its considerably shorter half-life and Actinium is chemically separated, so that a relatively pure isotope Ac-225 is obtained. Nevertheless, the final product contains unconverted 226Ra and other Ra isotopes. In addition, different decay products of Actinium occur as well as nuclear conversions of contaminating elements of the Al. Particularly important is to minimize the content of Sr and Ba which lead to the production of radioisotopes of Y and La, respectively. Several radioisotopes are produced as a result of nuclear reactions type (p,n) or (p,2n) on main impurities like Ba, Fe, Zn, Sr, Pt, V, Ti, Cr and Cu which are present in the Al carrier (foil, mesh) and/or in the Ra deposit. The radionuclides of major contribution to the total gamma activity excluding 226Ra and daughters are typically the following: 135La, 55Co 56Co, 67Ga, 57Ni, 135mBa, 133mBa, 131Ba, 129Cs, 51Cr, 48V, 52Mn, 54Mn, 65Zn. In addition, disturbing radiochemical impurities are 210Po and 210Pb resulting from the following decay chain: Ra-226 (alpha).→Rn-222(alpha) →Po-218 (alpha)→Pb-214 (beta)→Bi-214 (beta)→Po-214 (alpha)→Pb-210 (beta)→Bi-210 (beta)→Po-210 (alpha)→Pb-206 (stable). The 226Ra target used according to the procedure of EP 0 752 709 B1 is not specified in detail there. EP 0 962 942 A1 also describes a method for producing Ac-225 by irradiation of 226Ra with cyclotron accelerated protons having an energy of 10 to 20 MeV. According to the prior art of EP 0 962 942 A1, the target nuclide 226Ra is used in the form of RaCl2, which can be obtained for example by precipitation with concentrated HCl or radiumcarbonate (RaCO3). These radium substances are then pressed into target pellets. Prior to irradiation of the radium salts with protons, the pellets are heated to about 150° C. in order to release crystal water and are then sealed in a silver capsule. The capsule is then mounted on a frame-like support and connected to a water cooling circuit. The target itself exhibits a window, which is arranged in a way that the proton beam hits the target through the window. According to EP 0 962 942 A1, the target exhibits a surface of about 1 cm2. Although it is already possible to achieve good Actinium-225-yields with the targets according to EP 0 962 942 A1, it has turned out in practice that this target construction can heat itself under certain conditions due to the proton beam in such a way that the silver capsule tears open and might thus both destroy the target and contaminate the peripheral compounds. In order to solve these target problems, the inventors of the present inventions have designed two different improved radium targets for the production of radionuclides by means of accelerated protons, on the basis of the prior art of EP 0 962 942 A1. The one target preparation, a method of electrodeposition of 226Ra-material is disclosed in Applicant's DE 103 47 459 B3, the other one, an evaporation-dispensing system, is disclosed in Applicant's DE 10 2004 022 200 A1. Both application papers are herewith incorporated by reference in their entirety. Applicant's methods of target preparation provide the finally desired 225Ac-product on an Aluminium surface, and in a mixture of different radionuclides. Preferably, Al-mesh targets can be used as carrier of Ra in the targets. Al-mesh targets have an advantage in the achieved yield during electrodeposition. With the introduction of the Al-mesh disc as cathode in the electrodeposition process and as carrier of Ra in the target, the amount of Ra that can be deposited per disc could be increased. While, e.g. on an Al-foil disc the amount of Ra (experiments conducted at mg levels with Ba and at microgram levels with Ra-226) deposited was below 10 mg (2-3 mm at the eddies of one disc), in the case of the mesh disc, the amount of Ra was to approximately 70 mg (depending on the thickness of the deposit and other parameters, thicker deposits were not well adhered to the mesh anymore). Consequently the number of Ra/Al mesh discs that need to be introduced into the target cup was reduced to five or six instead of 10 or more as it was required by the use of Al-foil discs. The better yield of electrodeposition on Al mesh compared with the yield of Al foil is associated with the higher surface of the mesh. The fact that more Ra is electrodeposited on the Al also assures that the proton beam is hitting with higher probability the Ra and not much loss occurs in Al. The improvement by using an Al-mesh also facilitated the automation of the process. Preferably, a 99% pure Al provided by Good Fellow is used. The neutron activation results carried out on the mesh at the institute are reported below: Impurities in the Al mesh measured by ko-INAA are given in Table 1 TABLE 1ContentElement[μg/g]Fe1302Cr701Ni0.2Ga145Zn39Na9Mo3.5U1.3Co2.0Ce1.8La0.69W0.2Sb0.07Th0.18Br0.11Sm0.08As0.06Sc0.02Au0.002 As in the case of the Al-foil targets, the results from processing hundreds microCi of Ra/Al-mesh discs targets indicated that the selective leaching of Ra and Ac from the Al mesh (developed for the Al disc target) can be also performed. Already during the dissolution of the target it is possible to separate most of the Al and impurities from the Ac. A special advantage of the radium targets as described in DE 103 47 459 B3 and DE 10 2004 022 200 A1 is that they exhibit basically pure radium material in their radium containing coating. Hereby it is achieved that the targets are free of carriers or diluents, for example barium salts, which had to be added to the conventional radium targets of the prior art, in order to homogenize the radium-containing material. Due to the possibility to be able to work without such carrier materials as barium compounds, the chemical separation and purification of the created 225Ac becomes substantially more simple and the yields of irradiation are optimized, as competitive nuclear reactions, as for example those from barium nuclei, are not possible. To summarize, however, despite the already optimized target systems as provided by Applicant's DE 103 47 459 B3 and DE 10 2004 022 200 A1, the final 225 Ac-product still contains significant amounts of inorganic, radionuclide and organochemical impurities, which render the obtained 225Ac product unsuitable for direct medical or pharmaceutical application. In other words, the achieved product cannot be used immediately to prepare a pharmaceutical grade 225Ac-product for the manufacture of the radiopharmaceutical agents described in the introductory part of the present specification for cancer therapy. As a result, it is the object of the present invention to provide a purified and pharmaceutically acceptable 225Ac-containing radionuclide composition for further processing in the manufacture of 225Ac-containing therapeutic agents. With respect to a method, the above object is independently achieved by the characterising features of claims 1, 2, and 3. A pharmaceutically acceptable 225Ac-containing radionuclide composition in accordance with claim 21 also solves the above problem. In particular, the present invention suggests a method for purification of 225Ac from irradiated 226Ra-targets provided on a support, comprising the following steps: a) at least one leaching treatment of the 226Ra-targets for leaching essentially the entirety of 225Ac and 226Ra with nitric or hydrochloric acid under refluxing conditions; b) removing HCl if the solvent in step b) is hydrochloric acid and redissolving the resulting material in nitric acid; c) concentrating the 225Ac and 226Ra containing extracts; d) separating 225Ac from 226Ra and other Ra-isotopes by means of at least one first extraction chromatography with a solid support material having a first extractant system coated thereon, comprising at least one compound in accordance with general formula I in at least one compound in accordance with general formula II, wherein in formula I: R1, R2 independently is octyl, n-octyl, phenyl, or phenyl substituted with C1 to C3 alkyl; R3, R4 independently is propyl, isopropyl, butyl, or isobutyl; wherein in formula II: R5, R6, and R7 independently is C2-C5 alkyl, in particular, butyl, or isobutyl; e) eluting 225Ac which is retained on the solid support from the stationary phase with diluted nitric or hydrochloric acid, whereas 226Ra is passing through; f) separating 225Ac from 210Po and 210Pb by means of at least one second extraction chromatography with a solid support material having a second extractant system coated thereon, comprising at least one compound in accordance with general formula III in at least one compound in accordance with general formula IV, wherein in formula III: R8 and R9 independently is H, C1-C6 alkyl, or t-butyl; and wherein in formula IV: R10 is C4 to C12 alkyl; g) using 2M HCl as mobile phase; and h) recovering 225Ac from the flow-through, whereas 210Po and 210Pb are retained on the solid support. Alternatively, the method of the present invention for purification of 225Ac from irradiated 226Ra-targets provided on a support, comprises the following steps: a) at least one leaching treatment of the 226Ra-targets for leaching essentially the entirety of 225Ac and 226Ra with nitric or hydrochloric acid under refluxing conditions; b) removing HCl if the solvent in step b) is hydrochloric acid and redissolving the resulting material in nitric acid; c) concentrating the225 Ac and 226Ra containing extracts; d) separating 225Ac from 226Ra and other Ra-isotops by means of at least one first extraction chromatography with a solid support material having a first extractant system coated thereon, comprising at least one compound in accordance with general formula IA, wherein in formula IA: R1a, R2a, R3a, R4a independently is octyl or 2-ethylhexyl; eluting 225Ac which is retained on the solid support from the stationary phase with nitric acid within a concentration range of 0.3 M to 0.01 M or 1 M to 0.05 M hydrochloric acid, whereas 226Ra is passing through; f) separating 225Ac from 210Po and 210Pb by means of at least one second extraction chromatography with a solid support material having a second extractant system coated thereon, comprising at least one compound in accordance with general formula III in at least one compound in accordance with general formula IV. wherein in formula III: R8 and R9 independently is H, C1-C6 alkyl, or t-butyl; and wherein in formula IV: R10 is C4 to C12 alkyl; g) using 2M HCl as mobile phase; and h) recovering 225Ac from the flow-through, whereas 210Po and 210Pb are retained on the solid support. A further alternative method for purification of 225Ac from irradiated 226Ra-targets provided on a support, comprises the following steps: a) at least one leaching treatment of the 226Ra-targets for leaching essentially the entirety of 225Ac and 226Ra with nitric or hydrochloric acid under refluxing conditions; b) removing HCl if the solvent in step b) is hydrochloric acid and redissolving the resulting material in nitric acid; c) concentrating the 225Ac and 226Ra containing extracts; d) separating 225Ac from 226Ra and other Ra-isotops by means of at least one first extraction chromatography with a solid support material having a first extractant system coated thereon, comprising a compound in accordance with formula IB, de) eluting 225Ac which is retained on the solid support from the stationary phase with nitric acid having a concentration lower than appr. 0.1 M and higher then appr. 0.02 M, whereas 226Ra is passing through; f) separating 225Ac from 210Po and 210Pb by means of at least one second extraction chromatography with a solid support material having a second extractant system coated thereon, comprising at least one compound in accordance with general formula III in at least one compound in accordance with general formula IV, wherein in formula III; R8 and R9 independently is H, C1-C6 alkyl, or t-butyl; and wherein in formula IV: R10 is C4 to C12 alkyl; g) using 2M HCl as mobile phase; and h) recovering 225Ac from the flow-through, whereas 210Po and 210Pb are retained on the solid support. In a preferred method according to the invention, said nitric acid in step a) has a concentration range of appr. 0.001 M to 2 M, preferably appr. 0.1 M and said hydrochloric acid has a concentration range of 0.001 M to 2 M, and/or said acids are used at elevated temperatures, in particular, from appr. 30 to 90° C. Preferably, extracts from the leaching treatment are pooled and used for further processing. In concentration step c), typically, a final concentration of 1.5 M to 10 M of nitric acid is achieved. In a preferred embodiment of the invention, the first extractant system is octyl(phenyl)-N,N-diisobutylcarbamoylphosphine oxide [CMPO] in tributyl phosphate [TBP]. The second extractant system can be very efficiently a crown ether in accordance with formula V: Preferably, the crown ether of Formula V is used in 1-octanol. In a particularly preferred method of the invention, the second extractant system is 4,4′-bis(t-butylcyclohexano)-18-crown-6 in 1-octanol. An alternative second extractant system is 4,5′-bis(t-butylcyclohexano)-18-crown-6 in 1-octanol. In order to improve the final purification method, the first extraction chromatography of step d) can be repeated several times, in order to remove trace amounts of Ra-isotopes, depending on the desired purity of the 225Ac. In an analogues manner, the second extraction chromatography of step f) can be repeated several times, depending on the desired purity of the 225Ac. In a case of need, the first and the second extraction chromatography steps can be repeated several times for higher purification grades. In the method according to the present invention, it is preferred to remove Radon which is contained in the Al-support and/or in the converted products from the 225Ac products and the Al-support during the leaching process by means of suitable traps. Radon removing can be achieved for example easily by guiding Rn into a first alkaline trap to, neutralize acidic vapors, into a subsequent silica trap to absorb water, and into a final activated coal trap, wherein the activated coal trap is optionally cooled. Due to its value and hazardous potential, not converted 226Ra starting material is recovered from the flow-through of step e). 210Po and 210Pb impurities are eluted from the solid support of the second extraction chromatography in step h) by means of concentrated nitric acid or hydrochloric acid. In the present method of the invention, each purification step and/or fraction is preferably checked by means of α- and/or γ-spectroscopy. Respective fractions containing: a. 225Ac; or b. Ra-isotopes; or c. 210Po; and d. 210Pbare evaporated to wet or dry residues and redissolved, if necessary. For removing organochemical impurities, it is preferred to pass the prepurified 225Ac solutions through a resin filter which contains a non-ionic acrylic ester polymer. The final product as obtainable with the method of the present invention is a pharmaceutically acceptable 225Ac-containing radionuclide composition which can be used to prepare 225Ac-bearing radiopharmaceuticals as disclosed in the introductory part of the present specification. The present invention further comprises all combinations of all disclosed single features together, independent from their AND- or OR-linkage. A Ra batch sealed 226Ra source is pre-checked by γ-spectrometry, ampoule is broken. The Ra salts or compounds are dissolved and the solution is separated from glass by filtration. The filter and glass particles are leached out with 0.5 M HNO3 and pooled with 226Ra-containing liquid. This solution is subjected to an at least one extraction chromatography step, which results in a purified Ra fraction. The latter fraction is used—after a further concentration step—for preparing the 226Ra targets. Further details of 226Ra purification for cyclotron target preparation for 225Ac manufacture are described in the not prepublished DE 102005043012, filed on 9 Sep. 2005. The disclosure of this patent application is herewith incorporated by reference in its entirety. The present invention will be illustrated by way of an example of a target preparation by means of an electrodeposition according to DE 103 47 459 B3, “Radium-Target sowie Verfahren zu seiner Herstellung”. The person having average skill in the art will understand that the invention also works in targets prepared by the evaporation method in accordance with DE 10 2004 022 200 A1 “Method for producing 226Ra targets by the droplet-evaporation methods for irradiation in the cyclotron”. For the preparation of a 226Ra target, aluminium discs with a thickness of 0.015 mm and a diameter of about 5 cm with a minimal 99% purity of the aluminium are punched out and fixed on a stainless steel support. The support facilitates the handling of the aluminium foils and is removed after the electrodeposition itself, before the positioning of the radium-coated foil in the target itself. For the electrodeposition on the aluminium foil, a solution of a radium-226-nitrate is used, whereby in particular 226-radium chloride or 226-radium carbonate are absorbed beforehand for the transformation into the corresponding nitrate in about 0.05 M HNO3. Subsequently, the stainless steel support, on which the aluminium foil is fixed, is weighted and the net weight of the aluminium foil is determined. 150 ml (for electrodeposition on aluminium foils with a diameter of up to 15 cm) or 10 to 11 ml isopropanol are added into an electrodeposition cell (for aluminium foil discs with a diameter up to 2 cm). Then the required amount of radium-226 solution is filled into the electrolytic cell and 1-2 ml 0.05 M HNO3 are added. The total volume of the radium solution and 0.05 M HNO3 should not exceed about 2 ml, if aluminium foil discs with a diameter of up to 2 cm are used, and 20 ml at the most, if aluminium foil discs with a diameter of up to 15 cm are used. When high radium concentrations are used, a white precipitates may be formed. If this happens, 0.05 MHNO3 is further added until the precipitation has dissolved. The pH value of the depositing plating solution should preferably be between 4 and 5. For the electrodeposition of 226Ra containing material out of the plating solution the electric current is adjusted to about 60 mA and a voltage of about 200V is applied, monitored for a few minutes and, if necessary, readjusted. After the electrodeposition of the 226Ra solution has been completed, the plating solution is poured out, the support is rinsed with 2 to 3 ml isopropanol and the cell is disassembled and the aluminium foil is additionally rinsed with about 1 to 2 ml isopropanol. Afterwards, the support with the 226Ra coated aluminium foil arranged on it is dried under an infrared lamp until the weight remains constant, in order to render the radium-containing coating anhydrous. Afterwards, the stainless steel support with the fixed, coated aluminium foil is weighted and the net mass of the coated aluminium foil is determined. Then the yield is determined from the weighted mass of the 226Ra containing layer. An alternative way to monitor the yield of the electrodeposition—instead of weighing—is to measure the γ-activity of 226Ra by means of a high resolution γ-spectrometer. Subsequently, the stainless steel support and the aluminium foil are separated from each other. The dry aluminium fail coated with radium compounds is carefully covered with a new aluminium foil and the edges of the aluminium foil with which the Aluminium foil carrying the active layer is fixed are cut off, in order to minimize the amount of aluminium in the target itself. For the use as radium target in the proton beam of a cyclotron, a pile of the of the circular disc shaped aluminium foils prepared according to present examples, which are coated with radium-containing material in a ring shaped manner, are piled in a so called target cup. For the production of a folded radium target, one or more aluminium foils, in the case of this example, coated on one whole surface with 226Ra are covered in a way with another aluminium foil that the radium containing film is covered entirely. Then, the aluminium foil is folded several times until stripes of about 2 mm are obtained. The folded aluminium foil, which contains the layers of radium-containing material, in particular radium oxides, is then placed into the target for proton irradiation in the cyclotron or in the linear accelerator. With the above methods according to DE 103 47 459 B3 and DE 10 2004 022 200 A1, it is possible to obtain highly potent 226Ra targets on aluminium foil of a different thickness with different 226Ra-amounts. This method assures in particular to deposit s that are highly homogenous on the aluminium-226Ra target. This is particularly important for the irradiation of the target in the cyclotron, as the atomic nuclei of radium are thereby exposed homogenously to the proton flux. The use of aluminium as substrate for 226Ra offers various advantages for the irradiation in a cyclotron and the subsequent radiochemical processing of the irradiated target. The advantages of the aluminium lie in the nuclear physical and chemical properties of the aluminium: Nuclear properties: Aluminium has just one single stable isotope. The activation products formed from the aluminium are very short-lived. The formation of only short lived radionuclides on aluminium facilitates the radiochemical purification of Ac-225 and reduces the coaling time of the target after irradiation. As the loss of energy of protons in aluminium is very low, it is possible to use several thin films of aluminium without substantial reduction of the proton energy. Physical properties: Aluminium is a light metal with good thermal and electrical conductivity. It is easy to handle and can be adapted easily to the required geometry. Chemical properties: Aluminum can easily be dissolved in mineral acids and it can be easily separated from the resulting Actinium. Aluminum foils are available with a high degree of chemical purity and at reasonable prices. The deposition of 226Ra, e.g. as oxide or peroxide, allows to obtain a layer with a high content of radium, in particular about 70% of the deposited material per cm2. The electrodeposition yield is high. In practice it has turned out that about 4 to 5 g/cm2 226Ra with goad adhesive properties can be deposited on the aluminum foil. A. Selective Leaching of Ac and Ra from Irradiated Ra/Al Targets Prepared by the Electrodeposition Technique After the irradiation at the cyclotron, the target containing Ac and Ra is transferred to a shielded glove box and positioned in the disassembling and dissolution position. For leaching Ra and Ac from the irradiated Al discs or rings, a refluxing/distillation arrangement is used. This set up enables the condensation of hot water and acids vapours and their continuous reflux into the dissolution vessel and the collection of condensates when this is required. Using this arrangement any Rn which could be still present in the irradiated Al discs can be trapped in a series of traps. The traps are assembled in the following sequence: a NaOH bath to neutralize acid vapors, a silicagel trap to absorb water vapours and finally an activated cooled-coal trap to capture Rn. The arrangement used for leaching Ra and Ac from irradiated disc targets is a Refluxing/Destillation arrangement. Typically, the discs or rings are inserted in the flask and they are treated first with 30 ml hot 0.1-0.2 M HNO3 and then with 30 ml boiling 2M HNO3 or HCl. The leaching processes are repeated two-three times to wash out any remaining activities of Ra or Ac attached to the discs or to the walls of the glass vessel. The leaching solutions are first subjected to gamma-spectrometry and then combined if required. As a result of the leaching process at least two fractions are obtained: the first one contained the Ac, the Ra and part of the activation products (0.1-0.2 M HNO3) and the second contained most of the matrix (Al) and part of the activation products (2M HNO3 or concentrated HCl). The 0.1-0.2 M HNO3 fraction is taken for the Ac extraction process. This solution is converted to 2M HNO3, during this conversion any particles which can be suspended in solution should be dissolved. The volume of this fraction is generally set to 30 ml. The results indicate that more than 99% of Ac and Ra is contained in this fraction. Only trace amounts of Ac and Ra are found in the second leaching solution of 2M HNO3 which contains most of the Al from the Al discs. The activation products are found almost equally distributed between these two leaching fractions. This procedure facilitates the purification and recycling of Ra because both Ac and Ra are extracted from the foil or mesh without the total dissolution of the Al. In addition, the lower beta and gamma activity associated with activation products in the Ac/Ra leaching solution reduces the risk of radiation damage of the used resins, in particular RE resin. B. Selective Leaching of Ac and Ra from Irradiated Ra Targets Prepared by the Droplet-Evaporation Technique The Ra and Ac are removed from the irradiated Al cup by washing it with a 0.1 M HNO3 solution in an ultrasonic bath. After irradiation at the cyclotron and target disassembling in a shielded glove box; the Al target-cup which carries high radiation dose is transferred and placed into a 250 ml glass beaker (chosen for this specific target cup). This beaker is inserted in an ultrasonic bath. Once the target is inside the beaker or container, 100 mL 0.1M HNO3 are added into the Al-cup. This volume of 100 ml was selected to completely immerse the target into the leaching solution (the volume depends on the geometry and size of the target cup). The ultrasonic bath is then activated and the temperature of the water bath is kept at approximately 80 C during the process. The leaching process with the ultrasonic bath is conducted two times for short time (not more than 20-30 minutes). AN liquid fractions containing the Ra and Ac are combined in a glass beaker and evaporated to wet residues. Experiments with Ba nitrate has previously indicated that Ba at these conditions (setup, leaching volume, duration of US bath) is completely removed. The experiments with Ba also demonstrate that some particulate material associated with Al oxide is released from the target cup. Consequently before starting the separation process, this particulate fraction has to be dissolved either in hot 2M HNO3 or, if necessary, in 6M HCl and then converted to 2M HNO3. This solution is taken for the radiochemical separation. The recovery of Ra and Ac from the irradiated target by using this technique is always higher that 90%. Studies are being currently carried out to minimize the volume of 0.1M HNO3 solution needed to quantitatively recover the Ra and Ac from the target cup with a high chemical purity. These studies are conducted using also a new target design. Using this target we will be able to leach out the Ac and Re from the target cup without the need of disassembling it. The chemical purity of the leaching solution will define the complexity of the Ra recycling and purification process and therefore, it is important to obtain a chemical pure Ra solution already at this stage. C. Separation of Ac from Ra and Most of the Activation Products by Extraction Chromatography using the RE Resin as a First Extractant System The Ac/Ra separation is based on the use of the extraction chromatography resin RE Resin (EiChrom). In the RE resin, the stationary phase consist of octyl(phenyl)-N,N-diisobutylcarbamoylphosphine oxide in tributylphosphate. This extractant has the property to extract trivalent actinides and lanthanides from nitric acid solutions (e.g. 2M HNO3). The Ac can be eluted from the stationary phase by washing the column with diluted solutions of nitric or hydrochloric acid (e.g. 0.05M HNO3). Background Information The extraction of trivalent actinides especially transplutonium elements with bidentate organophosphorus compounds was extensively studied in the USA and the former USSR. In the USA, for example Horwitz et al. (1984, 1993) studied the extraction of Am and other elements with a great number of carbamoylphosphonates and carbamoylphosphine oxides. It was established that both kinds of extractants form trisolvates with lanthanides and trivalent actinides. The high extraction coefficient from nitric acid medium was explained by the bidentate coordination and cycle chromatography versions of the extraction system CMPO/TBP (e.g. TRU resin or RE resin, distributed by EICHROM). On both resins the tetravalent actinides show high retention from nitric acid solutions, having for example capacity factors (CF) in the range of 104-106 from 2-3 M HNO3 for the TRU Resin. In the same range of concentration, the CFs for lanthanides is in the order of 100 on the TRU Resin and between 100-200 on the RE Resin. For the RE, the CFs are higher for all relevant elements. The low retention of trivalent actinides from HCl and from diluted nitric acid solutions is the basis for their selective elution. According to Horwitz (1993); Ca, Fe (II) and commonly occurring polyatomic anions do not show significant effect on the Am retention from HNO3. Based on these properties, the TRU Resin has been applied for the separation of Am from Sr, Ca and Ba in environmental samples (e.g. Burnett et al.; 1995; Moreno et al.; 1997 and 1998). Burnett et al. (1995) applied the RE Resin in the combined determination of very small quantities of both 226Ra and 228Ra in environmental samples. In an entirely novel approach, in the present invention, the inventors have used the RE Resin for the separation of Ac from 226Ra, Al and from most of the activation products produced at the cyclotron by selectively extracting the Ac from 2 M HNO3. Ac is eluted from the stationary phase using 0.03-0.05 M HNO3. Separation of Ac from Ra, Al and Activation Products after the Irradiation of Ra/Al Targets at the Cyclotron FIG. 1 shows the flow chart of processes used for the extraction of Ac from irradiated targets. The size of the columns used (8 ml-bed volume) is chosen to maximize the retention of Ac on the RE resin from a large volume of loading solution and consequently to reduce the breakthrough of Ac in the Ra fraction. Assuming that a maximum of 0.5 g Ra and 0.5 g (extreme conditions) of Al can be present in the leaching/loading 2 M HNO3 solution, a total volume of up to 70-80 ml is needed for the total dissolution of Ra and Al. The results from experiments conducted with synthetic solutions and also with irradiated targets (ma of Ra and hundreds μCi of 225Ac) indicate that under similar conditions, most of the Ra can be removed by washing the column with approximately 50 ml 2 M HNO3 without a significant breakthrough of Ac. Meanwhile, most of the Ac can be eluted with 50 ml 0.05 M HNO3. The typical decontamination factor Df (Ac/Ra) is found to be in the order of 104 (one stage). D. Purification of the Ac D1. From Tracer Quantities of Ra by using a Repeated Extraction Chromatography Column with the RE Resin After the separation of the bulk of Ra, Al and activation products; 210Po (FIG. 3a) and some small quantities of Ra and isotopes of transition elements still remain in the Ac fraction. Therefore, a second separation of Ac from these remaining impurities is necessary. As shown in FIG. 1, the purification process consists of two stages: the first is a repetition of the Ac/Ra separation using the RE resin to provide an additional decontamination of Ac from Ra. The experiments have shown that the total decontamination factor Ac/Ra is approximately 106-107 by repeating twice the Ac/Ra separation with the RE resin under the described conditions. A further purification step enables the Ac/Po, Ac/Pb and Ac/Rn separation using a second extractant system, the Sr Resin (Eichrom) and this process is described below in section D2. D2. From Po and Pb Isotopes by using the Sr Resin as a Second Extractant System Background Information In the Sr Resin of the present example, the extractant in the stationary phase is a crown ether: 4,4′(5′)-bis(t-butylcyclohexaneno)-18-crown-6 in 1-octanol, Horwitz (1991, 1992) proposed this crown ether in 1-octanol to selectively extract Sr from concentrated nitric acid solutions. The extraction chromatography system is commercially available as Sr Resin (Eichrom) and has been applied to the determination of very low activities of 210Pb in environmental samples (Vajda et al,; 1995). Indeed, this resin has been also frequently used for the separation and purification of 90Sr from Ca, Mg and Ba in the radiochemical analysis of environmental samples (Vajda N. et al., 1992; Moreno et al, 1997 and 1998). In the present invention, the inventors have used the Sr Resin as second extractant system to purify Ac from Po, Pb and also Rn in 2 M HCl solutions: while Pb and Po are retained by the stationary phase from 2 M HCl, Ac passes through. Separation of Ac from Po and Pb in the Purification Scheme The presence of Po in the Ac (FIG. 3a) is observed on the alpha spectrum obtained from the Ac after the RE Resin separation. The presence of both Pb and Po in the Ac can be confirmed by measuring the gross alpha beta activity of aliquots taken from the Ac fraction. Without performing the purification with the Sr Resin, this parameter (gross alpha and beta activities) is much higher that the expected gross activity associated with the Ac and its decay products. Experiments carried out in dynamic conditions demonstrate that while Ac passed through the column, both Po and Pb were retained from 2M HCl acid. The 2M HCl fraction (loading and washing 2M HCl solutions) contained the Ac (FIG. 3b) while Po and Pb were retained by the stationary phase. E. Final Purification and Pre-Concentration of the Purified Ac Fraction Before proceeding with the final preconcentration step, the Ac fraction in 2M HCl acid from the Sr Resin is subject to quality control. At this stage, the radioisotopic purity is generally very high and it depends mainly on the presence of the short living 135La. Consequently the purity quickly increases within a few days after the end of production to more than 99.7%. The activity ratio 226Ra/225Ac (and also the activity ratio in relation to other long-lived isotopes) is checked and this ratio was usually below 5.10−4 in the Ac fraction. If the conditions for radioisotopic purity were not fulfilled, then a further purification of Ac from Ra and other relevant components is required. For this purpose, the Ac fraction obtained after concentration of the 2 M HCl solution is subject to a fast purification from Ra using a 2 ml-bed volume column with the RE resin. Usually, there is also a need to purify the Ac from soluble or dispersed organic materials. To separate the organic material, the solution is passed through a pre-filter 2 ml-bed volume resin (Eichrom) which contains a non-ionic acrylic ester polymer. The results indicate that the content of soluble organics is decreased in one order of magnitude and all the Ac can be filtered through this resin without retention. The results from the manual reprocessing of irradiated Ra/Al targets show that the recovery of Ac and Ra (excluding the recycling and further purification) are higher than 98% and 96% respectively. For processes conducted with 2- to 3 mg of Ra and hundreds μCi of 225Ac and using almost fully automated processes, the recovery factor of Ra is slightly lower but generally higher than 90-92%. This factor is intended to be increased by optimizing parameters associated with the automatic processes (e.g. liquid transfer, dead volumes, etc). F. Radioisotopic Impurities Measured by γ-Spectrometry The radioisotopic purity and the chemical purity of the Ac depend on the applied radiochemically procedures and also on the purity of the materials (mesh carrier, TC, etc) arid reagents (Ra solution, acids, etc). Particularly important is to minimize the content of Sr and Ba which lead to the production of radioisotopes of Y and La respectively that behave similarly to Ac during the separation process. As already mentioned in the introduction, several radioisotopes are produced as a result of nuclear reactions type (p,n) or (p,2n) on main impurities like Ba, Fe, Zn, Sr, Pt, V, Ti, Cr and Cu which are present in the Al carrier (foil, mesh) and/or in the Ra deposit. As an example, the γ-spectrum of a Ra fraction is shown in FIG. 2a. The radionuclides of major contribution to the total gamma activity excluding 226Ra and daughters were typically the following: 135La, 55CO, 56CO, 67Ga, 57Ni, 135mBa, 133mBa, 131Ba, 129Cs, 51Cr, 48V, 52Mn, 54Mn, 65Zn. Except for RE isotopes, most of these radionuclides are separated from the Ac. The typical radioisotopic purity of the purified Ac fraction is higher than 99.8% (see Table 2). The γ-spectrometry measurements of the purified Ac fraction (FIG. 3b) showed the presence of small quantities of rare earth radioisotopes, namely 87Y, 88Y, 139Ce. Small quantities of 194Au were some times observed (Pt anode) when the target was prepared by electrodeposition (Pt anode). Radioisotopic Impurities Measured by γ-Spectrometry The γ-spectrometry results after radiochemical separation of Ra in the aliquot sample indicate that the combined decontamination factor of 225Ac in relation to 226Ra (Df) is 106-107. This factor can be significantly improved by optimizing relevant parameters associated with the purification process. FIG. 3b shows the spectrum of the purified Ac extracted from an irradiated target. The spectrum clearly shows the peaks of 225Ac and decay products. No impurities of 210Po were observed which indicate that decontamination of 225Ac from 210Po is also very high by applying the described radiochemical scheme (see FIG. 3a). The content of impurities will decrease by increasing a proper selection of high purity reagents and materials (e.g. Al foils/mesh of better purity). In addition, when Bi is eluted from the Ac/Bi generator, the rare earth radioisotopes Ce, Ln, Y, and any 226Ra will remain on the stationary phase along with Ac (Ac/Bi generator) thus providing additional purification of 213Bi. TABLE 2Radioisotopic impurities measured in a purified Ac fraction fromirradiated target (electrodeposition).ActivityActivityRadioisotopicRadionuclideActivity [Bq]ratio ai/aAcbratio ai,t/aAccpurity [%]88Y4.661.57 × 10−44.1 × 10−499.96139Ce7.822.64 × 10−4226Ra0.4a 1.3 × 10−5209Tl562 221Fr2.93 × 104213Bi2.91 × 104225Ac2.96 × 104Except for 226Ra, all results were obtained by high resolution gamma-spectrometryaα-spectrometry after radiochemical separation of Ra (two independent analyses)bai/aAc impurity/actinium activity ratiocai,t/aAc ratio of the activity of all impurities to the activity of 225Ac55Co, 56Co, 57Co, 58Co, 67Ga, 194Au, 206Bi, 205Bi, 51Cr, 87Y, 48V, 54Mn, 65Zn, 226Ra, 214Pb and 214Bi were not detectable by γ-spectrometry.Chemical Impurities Measured in the Purified Ac Fraction The typical content of total inorganic impurities in the Ac purified fraction is generally below 100 μg. The following elements have been detected and quantified in the Ac fraction: Al, Ba, Ca, Cr, Cu, K, La, Mg, Mn, Na, P, Rb, Si, Sr, Ti, Zr, Zn and Zr. Thus, with the method according to the invention a pharmaceutically acceptable 225Ac preparation can be obtained, and the 226Ac can be used for the preparation of nuclear drugs for treatment of cancer as described in the introductory part of the present specification. |
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046735453 | claims | 1. In a nuclear plant facility having irradiated clips securing together a plurality of fuel rods of an irradiated nuclear fuel rod assembly contained underwater in a tank, a clip removing apparatus comprising: a tool having: jaw support block means slideably located between said first and second fuel rod support means for providing opposed pivotal mounting of said first and second jaws to permit said first and second jaws to pivot between open and closed positions as said block means is moved between forwardmost and rearwardmost positions relative to said first and second fuel rod support means; and jaw actuator means connected between said jaw support block means and said frame, remotely operable for selectively moving said jaw support block means between forwardmost and rearwardmost positions, whereby with said tool prepositioned with a clip to be removed between said first and second fuel rod support means, and said block means in a forwardmost position, and said first and second jaws open over the top and bottom edges of said clip, respectively, said jaw actuator means is operated to move said block means rearward, causing said first and second jaws to pivot closed for grasping said clip, and thereafter pulling said clip free of said fuel assembly as said block means continues to move rearward. television camera means mounted upon said frame and being prepositioned for forming and transmitting video signals of said visual image to said remotely located operating station. mirror means mounted upon said frame for reflecting back to said television camera means an image of areas about said first and second jaws not in the direct field of view of said television camera means. first and second movable mounting means for mounting said first and second mirrors to said frame, respectively, said first and second movable mounting means being independently remotely operable for selectively positioning said first and second mirrors, respectively, for obtaining a desired field of view. an operator support platform suspended above the water in said tank; and tool positioning means connected to said platform and rigidly attached to the top of said frame, for permitting an operator on said platform to position said tool for removing one of said clips. a support pole; said operator support bracket including a longitudinal slotway through its top and bottom portions overlying an area of the surface of said water in said tank; and carriage means slideably mounted upon the slotway of said operator support bracket, said carriage means including a hole for receiving said support pole, and means for both rigidly securing a portion of said pole within said hole, and permitting rotation of said pole, whereby one end of said pole protrudes above support platform for permitting access thereto by an operator on said support platform, and the other end of said pole extends below said support bracket for rigid attachment to said frame of said tool, therebypermitting said operator to push or pull said pole in an appropriate direction for moving said tool back and forth within a range in a horizontal plane within said tank, for adjusting the distance between said tool and a fuel assembly, and for further permitting said operator to rotate said pole for rotating said tool, all for positioning said tool to remove a clip from said fuel assembly. clip disposal basket means suspended from said operator support bracket to a position in said tank means proximate said tool, whereby after operating said tool to remove one of said clips from said fuel assembly, said operator manipulates said tool positioning means for positioning said tool with its jaws either directly over or within an opening in said basket means, and thereafter operates said jaw actuator means of said tool for moving said jaw support block means to a forwardmost position for opening said first and second jaws for releasing the clip into said basket means. clip ejection means remotely operable for pushing a previously retrieved clip away from said first and second jaws in an open position. plunger means slideably mounted on said jaw support block means, between said first and second jaws, for longitudinal movement between a retracted position away from the front of said first and second jaws, and an extended position at or about the front of said jaws, said plunger means being operable to rapidly move towards the extended position for ejecting a clip from said jaws. first and second ejector cylinders rigidly mounted to a back face of said jaw support block means, behind and centered upon two holes through the back face of said jaw support block means, the holes being located in a plane between said first and second jaw means, said first and second ejector cylinders being remotely operable for rapidly extending respective plungers for ejecting a clip from said jaws. television monitor means located at said remotely located operating station, for receiving said video signals and converting the same to said visual image for viewing by said operator. an operator support bracket secured to said platform and having a portion extending from said platform to a position over said water, said operator support bracket including a longitudinal slotway; a support pole; carriage means slideably mounted upon said slotway, said carriage means including a hole for receiving said support pole, and means for securing a portion of said pole within said hole, but permitting rotation of said pole whereby one end of said pole protrudes above said carriage means to a height permitting access thereto by an operator on said platform, the other end of said pole extending below said operator support bracket to a predetermined level within said tank; rotating means secured to said pole for permitting an operator to rotate said pole; tool means including a frame connected to the other end of said pole, said tool means including: television monitor means mounted upon said platform and positioned in the field of view of said operator, for receiving said video signals from said television camera means, and displaying a visual image thereof to said operator; and clip disposal basket means positioned near said tool means, for receiving said irradiated clips from said tool means; said apparatus being operable by said operator positioning said jaw means opposite one of said clips to be removed, said positioning being accomplished by the operator observing the image on said television monitor means, and operating said fuel elevator to obtain the necessary vertical position of said fuel assembly concurrent with moving said rotating means to place said jaw means opposite said clip to be removed, and thereafter operating said jaw means to engage and remove said clip, followed by moving said rotating means to position said jaw means with said clip within said basket, and releasing said clip into said basket via operation of said jaw means to an open position. 2. The apparatus of claim 1, further including tool positioning means rigidly connected to said frame, for permitting an operator to remotely position said tool for removing a clip from said fuel assembly. 3. The apparatus of claim 2, further including observation means for providing at a remotely located operating station a visual image of said first the second jaws and a portion of the associated surrounding area thereto, thereby assisting said operator in using said tool positioning means and operating said tool. 4. The apparatus of claim 3, wherein said observation means includes: 5. The apparatus of claim 4, wherein said observation means further includes: 6. The apparatus of claim 5, wherein said observation means further includes movable mounting means for mounting said mirror means to said frame, said movable mounting being remotely operable for selectively positioning said mirror means, for obtaining a desired field of view. 7. The apparatus of claim 5, wherein said mirror means includes first and second mirrors juxtaposed to said first and second jaws, respectively, for providing fields of view proximate said first and second jaws, respectively. 8. The apparatus of claim 7 said movable mounting means further includes: 9. The apparatus of claim 1 further including: 10. The apparatus of claim 9, wherein said tool positioning means further includes: 11. The apparatus of claim 10, wherein said tool positioning means further includes a spoked wheel rigidly secured to said one end of said support pole for permitting an operator easy manipulation of said tool positioning means. 12. The apparatus of claim 1 further including: 13. The apparatus of claim 1, wherein said tool further includes: 14. The apparatus of claim 13, wherein said clip ejection means includes: 15. The apparatus of claim 14, wherein said plunger means includes: 16. The apparatus of claim 4, further including: 17. In a nuclear plant facility having irradiated clips securing a plurality of rods of a nuclear fuel assembly together, the nuclear fuel rod assembly being contained underwater in a storage tank, said storage tank including an operator platform over a portion of said tank, and a fuel elevator to move fuel assemblies to different locations in said tank, a remotely operable apparatus for removing said clips comprising: |
description | The present invention generally relates to the field of data storage systems, particularly a system and method for generating test plans and identifying new test requirements for quality assurance in a hardware or software product. During design and development of a hardware or software product, manufacturers often put great effort to assure desired quality of the hardware or software product. The production validation (quality assurance) process involves various activities such as test plan and management, test execution, test report, and the like. Extensive testing is often completed on the product to assure performance and that the product meets customer requirements, industry standards for hardware, software, and BIOS (basic input/output system) compatibility. Testers may need to perform tests on a product with new features desired by a customer. Conventionally, a set of predetermined test requirements and recorded results for legacy features of the product may be readily available by a quality assurance tool. However, the legacy test plan may or may not cover verification of new features. This may cause testers to conduct improper tests (ad hoc testing) and/or waste valuable resources during a product validation process. Thus, it would be desirable to provide a method and system to address the foregoing-described problems. Accordingly, the present invention provides a quality assurance system and method for identifying new test requirements, maintaining a test database and generating test plans for different versions of a product by mapping new requirements, for which a user assigns a name for each requirement using a unique naming convention, to a test database, which uses the same naming convention for legacy test cases, to derive the test cases required to verify different aspects of the new features and to identify what new test cases are required to be scoped for aspects that are not covered in the test database. The present method may automatically generate a test plan from the selected legacy tests and new test cases defined by a test engineer. In an exemplary aspect of the present invention, a method for ensuring desired quality of different versions of a product is provided. Old versions of the product may be previously tested, and test plan documents associated with previously tested versions of the product may be stored in a database. The database may store test plans, test configurations, test scopes, and the like for previously tested versions of the product. Customer desired features including new and/or modified features of a new version of the product may be received. Product design requirements may be determined based on the received customer desired features. Then, an initial test plan (first test plan) for the product may be generated by querying a database based on said product design requirements. The method may determine whether the initial test plan does not include a test for each new and/or modified feature. If the initial test plan does not include a test for a certain new feature, a new test for the certain new feature may be provided by a tester. The database may be updated by adding the provided new test. Next, a test plan document for the product may be generated based on the updated database. The generated test plan document may be verified through a comparison with a test plan generated by querying the updated database with the product design requirements. In an additional aspect of the present invention, a Graphic User Interface (GUI) may be provided to allow a user (tester) to provide product design requirements and new test procedures. New feature requirements for the new version of the product may be identified from a drop menu on the GUI, which based on the name of the new features. The menu has the capability of adding entries to new keys to the database if the key does not exit in the database. Then, descriptive keys for the verified new features, sub-features, or modified features of a new product may be created. In another aspect of the present invention, the test plan may be extracted, or derived from existing test cases stored in the database. New tests required to verify the new features of the product under development may be identified. Test areas, test cases, and/or test steps which require modification to handle new features or modified legacy features in the product under design may be identified. In this manner, the granularity of the existing test cases stored in databases may be enhanced. It is to be understood that both the foregoing general description and the following detailed description are exemplary and explanatory only and are not restrictive of the invention as claimed. The accompanying drawings, which are incorporated in and constitute a part of the specification, illustrate an embodiment of the invention and together with the general description, serve to explain the principles of the invention. Reference will now be made in detail to the presently preferred embodiments of the invention, examples of which are illustrated in the accompanying drawings. In the following description, numerous specific descriptions are set forth in order to provide a thorough understanding of the present invention. It should be appreciated by those skilled in the art that the present invention may be practiced without some or all of these specific details. In some instances, well known process operations have not been described in detail in order not to obscure the present invention. Referring now to FIG. 1, a block diagram of a Quality Assurance (QA) system 100 in accordance with an exemplary embodiment of the present invention is shown. The QA system 100 may include a QA planning module 102, a user interface module 104, a test database 106, and a test database management module (not shown). The QA system 100 may be configured to ensure desired quality of different versions of a product. Old versions of the product may be previously tested, and test plan documents associated with previously tested versions of the product may be stored in the test database 106. The test database 106 may store test plans, test configurations, test scopes, and the like for previously tested versions of the product. The QA planning module 102 may be configured to query the test database 106 with product requirements for a new version of the product, build a test plan based on the queried information, identify a product requirement requiring new tests to be defined, and the like. In an embodiment, GUIs 108 and 110 may be provided for several users to interact with the QA system 100. The GUIs 108 and 110 may be provided by the user interface module 104 to the users. The GUI 108 or 110 may be suitable for receiving various user inputs and for displaying menu options and QA system outputs including general test plan output and test plan documents for the new version of the product. The test plan output on the GUI 108 or 110 may include an indication for lack of predefined tests for new features in the test database. The indication may be highlighted, blinked or the like to be visually noticeable. The user may provide new tests for the new features through the GUIs 108 and 110. Upon reception of the new tests through the GUI 108 or 110, the test database management module (not shown) may update the database 106 accordingly. Then, the QA planning module 102 may generate a test plan document for the new version of the product by querying the updated test database 106. Referring now to FIG. 2, a flow diagram of a method 200 for ensuring desired quality of a product in accordance with an exemplary embodiment of the present invention is shown. The method 200 may be implemented in the system 100 shown in FIG. 1. Customer desired features for a new product may be received 202. The customer desired features may be translated into high level product requirements suitable for being used in a QA system 204. The customer desired features for the new product may include legacy features, new features, modified features, and the like. A product design requirement associated with a new feature may be identified 206. Test plan information for the product may be retrieved through querying a test database with the high level product requirements. A first general test plan may be generated based on the test plan information 208. In some instances, the test plan information does not include test information for new features or modified features. If the new tests are not found in the test plan information retrieved from the database, the new tests are to be defined for the new feature. The new tests may be provided by the user if the first general test plan does not include the new test for the new feature 210. Upon reception of the new tests, the test database may be updated accordingly 212. A test plan document for the new product may be generated based on the updated database 214. The test plan document may include test scope, detail test cases, a test configuration, test resources, test duration, a test summary and the like. The generated test plan document may be verified 216. FIG. 3 illustrates a screen shot of a Graphic User Interface 300 displayed for a user in accordance with an exemplary embodiment of the present invention. The GUI 300 may include a requirements section 302, a query database option 304, a build test plan option 306, a generate test plan option 308, a clear test plan option 310, a test plan content section 312, and a test database section 314. A user may select requirements from the requirements section 302 to check whether there are test cases in the database to verify these product requirements. For example, as shown in FIG. 4(i), the O/S Specific requirement, the Santricty Software requirement, the New HBA requirement, and the Host Interface requirement have been selected, and the query database option 304 has been selected. After querying the database, it may be reported in the test database section 314 that the database includes test plan information (i.e., test cases which includes test steps required to verify the requirement under question) for some of the selected requirements and no test plan/test case information for the rest of the selected requirements. For example, in FIG. 4(i) the database includes test plan/test case information for the Host Interface requirement and the Santricty Software requirement, but no test plan/test case information for the O/S Specific requirement and the New HBA requirement. Under this scenario, a user may need to define the test case information for the selected requirements that have no corresponding pre-existing test case information in the database. After the test plan information is received, the database may be updated. Upon the build test plan option 306 is selected, the test plan content (i.e., test file) may be displayed in the test plan content section 312 (see FIG. 5(ii)). In an exemplary embodiment, the structure of the database for the test database section 314 is as follows: <?xml version=“1.0” encoding=“UTF-8” ?>−<TEST_DATA>−<TEST><NAME>HSWx-HostInterface-BASC-xxxx-0001</NAME><OBJECTIVE>Increase volume capacity</OBJECTIVE>−<CONFIGURATION><NAME>2 × 2 × 2 × 4 ** CFG 8 **</NAME><SERVER>2</SERVER><SWITCH>2</SWITCH><CONTROLLER_MODULE>2</CONTROLLER_MODULE>−<TRAYS>−<TRAY><COUNT>3</COUNT><DRIVES>18</DRIVES></TRAY>−<TRAY><COUNT>1</COUNT><DRIVES>12</DRIVES></TRAY></TRAYS></CONFIGURATION>−<STEPS><STEP>Run SANSCAN</STEP><STEP>Create 5 RAID volumes</STEP><STEP>Run I/O</STEP><STEP>Save MEL</STEP></STEPS><DURATION>.2</DURATION>−<RESOURCES><ENGINEER>.2</ENGINEER><STUDENT>1</STUDENT></RESOURCES></TEST>−<TEST><NAME>HSWx-RAID Platform-BASC-xxxx-0006</NAME><OBJECTIVE>Volume creation Verify that Increase volume capacitycommands can be issued successfully on snapshots andrepositorires using SMcli.</OBJECTIVE>−<CONFIGURATION><NAME>2 × 1 × 2 × 3 ** CFG 5 **</NAME><SERVER>2</SERVER><SWITCH>1</SWITCH><CONTROLLER_MODULE>2</CONTROLLER_MODULE>−<TRAYS>−<TRAY><COUNT>3</COUNT><DRIVES>18</DRIVES></TRAY>−<TRAY><COUNT>1</COUNT><DRIVES>12</DRIVES></TRAY></TRAYS></CONFIGURATION>−<STEPS><STEP>Run SANSCAN</STEP><STEP>Create 5 RAID volumes</STEP><STEP>Run I/O</STEP><STEP>Reset controller</STEP><STEP>verify that config are ACCESSED</STEP></STEPS><DURATION>.2</DURATION>−<RESOURCES><ENGINEER>.2</ENGINEER><STUDENT>1</STUDENT></RESOURCES></TEST>−<TEST><NAME>HSWx-Santricty Software- xxxx-0001</NAME><OBJECTIVE>healthStatusTest Purpose: This is intended to be asystem test in a het host environment with severalservers.</OBJECTIVE>−<CONFIGURATION><NAME>4 × 1 × 2 × 3 ** CFG 13 **</NAME><SERVER>4</SERVER><SWITCH>1</SWITCH><CONTROLLER_MODULE>2</CONTROLLER_MODULE>−<TRAYS>−<TRAY><COUNT>3</COUNT><DRIVES>18</DRIVES></TRAY>−<TRAY><COUNT>1</COUNT><DRIVES>12</DRIVES></TRAY></TRAYS></CONFIGURATION>−<STEPS><STEP>Run SANSCAN</STEP><STEP>Create 5 RAID volumes</STEP><STEP>Run I/O</STEP><STEP>Reset controller</STEP><STEP>verify that config are accessed</STEP></STEPS><DURATION>.5</DURATION>−<RESOURCES><ENGINEER>.2</ENGINEER><STUDENT>1</STUDENT></RESOURCES></TEST>−<TEST><NAME>HSWx-HostInterface-BASC-xxxx-0002</NAME><OBJECTIVE>Increase Configuration capacity</OBJECTIVE>−<CONFIGURATION><NAME>2 × 0 × 2 × 4 ** CFG 8 **</NAME><SERVER>2</SERVER><SWITCH>0</SWITCH><CONTROLLER_MODULE>2</CONTROLLER_MODULE>−<TRAYS>−<TRAY><COUNT>3</COUNT><DRIVES>18</DRIVES></TRAY>−<TRAY><COUNT>1</COUNT><DRIVES>12</DRIVES></TRAY></TRAYS></CONFIGURATION>−<STEPS><STEP>Run SANSCAN</STEP><STEP>Create 5 RAID volumes</STEP><STEP>Run I/O</STEP><STEP>Save MEL</STEP></STEPS><DURATION>.2</DURATION>−<RESOURCES><ENGINEER>.2</ENGINEER><STUDENT>1</STUDENT></RESOURCES></TEST>−<TEST><NAME>HSWx-New Cluster-BASC-xxxx-0001</NAME><OBJECTIVE>Increase volume capacity</OBJECTIVE>−<CONFIGURATION><NAME>2 × 1 × 2 × 2 ** CFG 8 **</NAME><SERVER>2</SERVER><SWITCH>1</SWITCH><CONTROLLER_MODULE>2</CONTROLLER_MODULE>−<TRAYS>−<TRAY><COUNT>1</COUNT><DRIVES>12</DRIVES></TRAY>−<TRAY><COUNT>1</COUNT><DRIVES>6</DRIVES></TRAY></TRAYS></CONFIGURATION>−<STEPS><STEP>Run SANSCAN</STEP><STEP>Create 5 RAID volumes</STEP><STEP>Run I/O</STEP><STEP>Save MEL</STEP></STEPS><DURATION>.2</DURATION>−<RESOURCES><ENGINEER>.2</ENGINEER><STUDENT>1</STUDENT></RESOURCES></TEST></TEST_DATA> The present invention may have the following advantages. It may identify the new features requirement from a drop-down menu, which is based on the name of the new features. This menu has the capability of adding entries to new keys, which correspond to requirements, if the key does not exit. Moreover, it may automatically create descriptive keys for all new feature, sub-feature, or modified feature of a new product in the key. Additionally, it may automatically generate a test plan, which is required to verify the functionality and features of a new RAID product or an updated RAID product which is scoped to meet customer needs. The test plan is extracted and/or derived from an existing test cases database. Further, it may identify new tests required to verify the new features of the product under development. In addition, it may identify test areas, test cases, and/or test steps which require modification to handle new features or modified legacy features in the product under design. Moreover, it may enhance the granularity of the existing test cases stored in databases. It is to be noted that the foregoing described embodiments according to the present invention may be conveniently implemented using conventional general purpose digital computers programmed according to the teachings of the present specification, as will be apparent to those skilled in the computer art. Appropriate software coding may readily be prepared by skilled programmers based on the teachings of the present disclosure, as will be apparent to those skilled in the software art. It is to be understood that the present invention may be conveniently implemented in forms of a software package. Such a software package may be a computer program product which employs a computer-readable storage medium including stored computer code which is used to program a computer to perform the disclosed function and process of the present invention. The computer-readable medium may include, but is not limited to, any type of conventional floppy disk, optical disk, CD-ROM, magneto-optical disk, ROM, RAM, EPROM, EEPROM, magnetic or optical card, or any other suitable media for storing electronic instructions. It is understood that the specific order or hierarchy of steps in the methods disclosed are examples of exemplary approaches. Based upon design preferences, it is understood that the specific order or hierarchy of steps in the method can be rearranged while remaining within the scope of the present invention. The accompanying method claims present elements of the various steps in a sample order, and are not meant to be limited to the specific order or hierarchy presented. It is believed that the present invention and many of its attendant advantages will be understood by the foregoing description. It is also believed that it will be apparent that various changes may be made in the form, construction and arrangement of the components thereof without departing from the scope and spirit of the invention or without sacrificing all of its material advantages. The form herein before described being merely an explanatory embodiment thereof, it is the intention of the following claims to encompass and include such changes. |
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048204773 | claims | 1. In a light-water cooled nuclear reactor, a method for preventing formation of tritium in a circulating light-water coolant medium for said nuclear reactor comprising diverting only a portion of said circulating light water coolant medium to means for removing deuterium to continuously remove deuterium from said portion of said circulating medium during normal operation of said reactor. 2. The method of claim 1 wherein said portion is substantially one percent of said circulating medium. 3. The method of claim 1 wherein said step of continuously removing deuterium comprises returning a remainder of said portion having a reduced deuterium content to said circulating medium. 4. The method of claim 3 wherein said step of removing comprises physically separating said portion from said circulating medium and directing said portion to a means for removing deuterium from said portion. 5. The method of claim 4 wherein said means for removing deuterium is a distillation column. 6. The method of claim 5 wherein the deuterium concentration of said remainder is less than about one part in 6,400. 7. The method of claim 5 wherein said step of separating comprises vaporizing light water from said portion and condensing said vaporized light water to produce said remainder. 8. In a light-water nuclear reactor, wherein predominantly light water is a circulating coolant medium in a coolant circuit, the improvement comprising means for directing only a portion of said circulating coolant from said coolant circuit to means for removing deuterium from said circulating coolant during normal operation of said reactor, and means for returning said portion to said circuit, whereby the formation of tritium is substantially prevented. 9. The reactor of claim 8 wherein said means for removing deuterium is a distillation column which separates heavy water in said portion from light water in said portion, and means for returing said separated light water to said circulating coolant. 10. The reactor of claim 9 wherein said portion is substantially one percent of said circulating coolant. |
abstract | A quick disconnect for a control rod drive mechanism seismic support tie rod system that is remotely operable from a nuclear power plant's operating deck. A wall mounted anchor in the reactor cavity contains one half of a disconnect coupling that interfaces with the other half of the disconnect coupling on the ends of the tie rods employing a remote winching system that is actuated from the top of the reactor head assembly. A latching mechanism is then actuated from the refueling cavity operating deck to lock the tie rod in place and prevent displacement during a seismic or pipe break event. The tie rod may similarly be unlocked from the wall anchor and raised above the reactor head assembly as part of a reactor head disassembly operation to gain access to the core of the reactor vessel for refueling. |
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abstract | The disclosure relates to a device for checking a fuel rod comprising a testing container having first and second chambers, a first checking device arranged in the testing container, wherein the testing container has at least one inlet opening, at least one outlet opening, and an insertion opening for inserting the fuel rod into the second chamber, and wherein a valve is arranged in a connecting channel connecting the first and the second chamber. A method is disclosed for checking a fuel rod in a water-filled basin of a submerged nuclear plant having such a device, wherein the fuel rod is inserted into the second chamber of the testing container through the insertion opening while the valve is closed, wherein a fluid is fed in via the at least one inlet opening, and wherein the valve is opened in order to check the fuel rod with the first checking device. |
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description | The disclosed system and method relate to semiconductor processing. More specifically, the disclosed system and method relate to wafer scanning and annealing. Semiconductors devices are increasingly being scaled down and gate dielectrics become thinner. At such a small dimension, any tunneling through a gate dielectric layer to the underlying channel region significantly increases gate-to-channel leakage current and increases power consumption. Gate dielectrics are therefore required to have high density and fewer pores. High-k materials are commonly used as gate dielectrics for MOSFET devices. However, high-k materials have the disadvantage that their densities are lower than conventional thermally grown, low-k silicon dioxide. One of the methods of improving density is annealing, by which the material density is increased and thus electrical properties are improved. Some conventional methods of gate-dielectric annealing are performed by rapid thermal annealing (RTA) or furnace annealing, which requires temperature as high as around 700° C. Since wafers are typically kept at high temperature for a long period, conventional rapid thermal annealing and furnace annealing have drawbacks of agglomeration formation, high thermal budget cost, and high diffusion of impurities. Laser spike annealing (LSA) has been developed to overcome the shortfalls of RTA. Conventional methods of LSA involve arc scanning in which the laser is scanned in an arc across the semiconductor wafer. For example, FIGS. 1A and 1B illustrate a conventional LSA arc-scanning process of a semiconductor wafer 100. As shown in FIG. 1A, a semiconductor wafer 100 is placed on a pedestal 102 which may move as indicated by the arrows. A laser source 104 directs a beam of light 106 onto the semiconductor wafer 100 at an angle θ from an axis normal to the plane of the semiconductor wafer 100. FIG. 1B illustrates the conventional scanning paths 108a-108g of the laser beam 106 across the surface of the semiconductor wafer 100. In a conventional arc-scanning process, the laser beam 106 will scan path 108a first followed by 108b and so on until the final inverted “fill-in” scan 108g is performed. While these conventional methods of LSA arc-scanning overcome some of the disadvantages of RTA, a semiconductor wafer 100 may have large variations in material characteristics that are difficult to account for using the conventional LSA methodology. The material variations may extend from one die to another which may negatively affect the performance of the integrated circuits formed on the dies and wafer 100. Accordingly, an improved method of laser annealing for semiconductor wafers is desirable. In some embodiments, a method includes dividing a semiconductor wafer into a plurality of dies areas, generating a map of the semiconductor wafer, scanning a first one of the plurality of die areas of the semiconductor wafer with a laser, adjusting a parameter of the laser based on the map of the semiconductor wafer and a value of the first measurement associated with a second one of the plurality of die areas, and scanning the second die area. The map characterizing the die areas based on a first measurement of each individual die area. The adjusting being performed after scanning the first die area. In some embodiments, a system includes a laser light source, a pedestal, and a processor in signal communication with the laser light source and the pedestal. The pedestal is configured to hold a semiconductor wafer. One of the laser light source and the pedestal is configured to move in relation to the other of the laser light source and the pedestal. The processor is configured to divide the semiconductor wafer into a plurality of die areas, control the relative movement between the pedestal and the laser light source, and adjust a parameter of the laser light source individually for scanning each of the die areas based on a map of the semiconductor wafer. The map characterizes the individual die areas of the semiconductor wafer based on a respective value or a first measurement taken in each respective die area. In some embodiments, a machine readable storage medium is encoded with program code. When the program code is executed by a processor, the processor performs a method. The method includes dividing a semiconductor wafer into a plurality of dies areas, generating a map of the semiconductor wafer, scanning a first one of the plurality of die areas of the semiconductor wafer with a laser, adjusting a parameter of the laser based on the map of the semiconductor wafer and a value of the first measurement associated with a second one of the plurality of die areas, and scanning the second die area. The map characterizing the die areas based on a first measurement of each individual die area. The adjusting being performed after scanning the first die area. An improved system and method for performing a laser spike annealing (LSA) scan is now described. The LSA scan may be controlled by a processor 601 as shown in FIG. 6. FIG. 6 is a block diagram of an exemplary system. As shown in FIG. 6, a semiconductor wafer 600 is placed on a pedestal 602 which may move as indicated by the arrows. A laser source 604 directs a beam of light 606 onto the semiconductor wafer 600 at an angle θ from an axis normal to the plane of the semiconductor wafer 600. Processor 601 controls at least one parameter of the laser source 604, and receives information from the laser (which may include the laser parameter settings and/or measurements). The processor 601 is also coupled to the pedestal 602 for controlling the pedestal and receiving position data from the pedestal. The processor 601 reads data and computer program instructions from a computer readable storage medium 603 and stores data in the computer readable storage medium. FIG. 2 is a flow diagram of one embodiment of an improved method of performing a LSA scan 200. At block 202, the semiconductor wafer 300 is divided into one or more dies 302 as shown in FIG. 3A. At block 204, pre-annealing sheet resistance measurements 304 and/or thermal wave measurements 306 may be performed on test structures at several locations in each of the one or more dies 302, and the measurements stored in a computer readable storage medium 603. In some embodiments, the one or more sheet resistance measurements 304 may be performed over 1 mm by 1 mm areas of dies 302, and the one or more thermal wave measurements 306 may be performed over 50 μm by 50 μm areas of dies 302. However, one skilled in the art will understand that other dimensions for the one or more sheet resistance measurements 304 and thermal wave measurements 306 may be performed. Additionally, one skilled in the art will understand that other measurements including, but not limited to, Photo Luminescence Imaging (PLI) measurements, may be performed. The sheet resistance and thermal wave measurements may be used to create a map of the semiconductor wafer 300 on a per die basis. The map of the die 302 may identify defects within each die 302 of the wafer 300. Some of these defects detected during the sheet resistance measurements 304 or the thermal wave measurements 302 may be alleviated by LSA scanning. For example, thermal wave imaging of the wafer 300 may identify areas of a die 302 having high concentrations of dopant impurities that may be the result of ion implantation of the wafer 300. These high concentrations of dopant impurities may be alleviated by the LSA scanning, which may diffuse or even out the high dopant concentrations. The map generated by the pre-scanning measurements of the wafer 100 may be stored in a computer readable storage medium 603 and used during the LSA scanning to adjust the properties of the laser beam as described below. At block 206, the map generated by the pre-annealing measurements may be used by processor 601 to adjust the parameters of the laser individually for each die while the LSA scanning is performed. For example, the intensity of the laser may be adjusted individually for each die, to provide the amount of annealing needed to correct defects or to activate dopants. The map may be implemented in a feed-forward system such that the adjustable parameters of the laser (e.g., wavelength, intensity, duration of exposure, etc.) may be varied as the laser moves across a semiconductor wafer 300 or a die 302a-302c. At block 208, field-by-field LSA scanning is performed. FIG. 4A illustrates one example of a scanning sequence (e.g., scanning sequence lines 310a-310d) of the semiconductor wafer 300. The dashed line in FIG. 2 indicates that the steps 206 and 208 can be repeated (i.e., the laser can be adjusted each time another die is to be annealed). As shown in FIG. 4A, the LSA scan of the wafer 300 may be performed through a plurality of linear scanning passes 310a-310d in which the dies 302 are sequentially scanned. For example, a row or column of the dies 302 may be sequentially scanned as shown by scanning sequence lines 310a-310d. One skilled in the art will understand that the scanning sequence may be performed from left-to-right, right-to-left, top-to-bottom, bottom-to-top, or the like. Additionally, one skilled in the art will understand that the dies 302 may be scanned in a nonlinear or non-sequential pattern as well. Detail A of FIG. 4A is shown in FIG. 4B and illustrates one example of the scanning paths 308a-308c of the laser beam. Note that, although the laser is described as moving in relation to wafer 300, the wafer 300 may be on a pedestal that moves the wafer 300 in relation to the laser beam. An example of a commercially available pedestal includes, but is not limited to, an Ultra LSA 100 pedestal. As shown in FIG. 4B, the scanning paths 308a-308c of the laser beam may wind from one portion (e.g., top portion 314) of die 302a to another portion (e.g., lower portion 316b). Once a die 302a has been scanned, then the laser beam may move to another die, e.g., adjacent die 302b, as identified by the scanning sequences 310a-310d shown in FIG. 4A. The scan path 308b of the next die 302b may be identical to the scan path 308a of the previous die 302a. In some embodiments, scan paths 308a-308c are not identical to each other and may be implemented such that the laser may seamlessly move from one die 302a to a second die 302b without turning off. FIGS. 5A-5D illustrate various scanning paths 408a-408c, 508a-508c, 608a-608c, and 708a-708c that may also may be implemented. One skilled in the art will understand that other in-die scanning paths may be implemented that scan the entire wafer. At block 210, post-scanning sheet resistance measurements or thermal wave measurements of the wafer 300 may be performed. These post-scan measurements may be used to confirm parameters of the laser scanning and to ensure that the advanced process control is optimal. For example, if the post-scan measurements, e.g., the Rs, TW, and/or PLI measurements, identify that the wafer does not have uniform characteristics, then a parameter of the laser scanning (e.g., length of scan, wavelength of laser, intensity of laser) may be adjusted using a feed forward control system. At block 212, the wafer 300 may undergo resistance protective oxide (RPO) formation as well as additional processing steps necessary to finish the integrated circuitry. Dividing the semiconductor wafer into a plurality of dies 302a-302c and scanning the semiconductor wafer 300 on a per die 302a-302c basis advantageously enables each die 302a-302c to be annealed in such a manner that the properties of the dies 302a-302c are more uniform than may be achieved through conventional arc scanning. The present invention may be embodied in the form of computer-implemented processes and apparatus for practicing those processes. The present invention may also be embodied in tangible machine readable storage media encoded with computer program code, such as random access memory (RAM), floppy diskettes, read only memories (ROMs), CD-ROMs, blu-ray disk, DVD ROM, hard disk drives, flash memories, or any other machine-readable storage medium, wherein, when the computer program code is loaded into and executed by a computer, the computer becomes a particular machine for practicing the invention. When implemented on a general-purpose processor, the computer program code segments configure the processor to create specific logic circuits. The invention may alternatively be embodied in a digital signal processor formed of application specific integrated circuits for performing a method according to the principles of the invention. Although the invention has been described in terms of exemplary embodiments, it is not limited thereto. Rather, the appended claims should be construed broadly, to include other variants and embodiments of the invention, which may be made by those skilled in the art without departing from the scope and range of equivalents of the invention. |
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description | The present application is a Continuation-in-Part of and claims priority of U.S. patent application Ser. No. 12/423,909, filed Apr. 15, 2009, which is a Continuation-in-Part and claims priority of International Application No. PCT/EP2008/003183, filed Apr. 21, 2008, the content of which is hereby incorporated by reference in its entirety. The present invention relates to multi-leaf collimators. Radiotherapeutic apparatus involves the production of a beam of ionising radiation, usually x-rays or a beam of electrons or other sub-atomic particles. This is directed towards a cancerous region of the patient, and adversely affects the tumour cells causing an alleviation of the patient's symptoms. Generally, it is preferred to delimit the radiation beam so that the dose is maximised in the tumour cells and minimised in healthy cells of the patient, as this improves the efficiency of treatment and reduces the side effects suffered by a patient, A variety of methods of doing so have evolved. One principal component in delimiting the radiation dose is the so-called “multi-leaf collimator” (MLC). This is a collimator which consists of a large number of elongate thin leaves arranged side to side in an array. Each leaf is moveable longitudinally so that its tip can be extended into or withdrawn from the radiation field. The array of leaf tips can thus be positioned so as to define a variable edge to the collimator. All the leaves can be withdrawn to open the radiation field, or all the leaves can be extended so as to close it down. Alternatively, some leaves can be withdrawn and some extended so as to define any desired shape, within operational limits. A multi-leaf collimator usually consists of two banks of such arrays, each bank projecting into the radiation field from opposite sides of the collimator. The leaves on the MLC leaf bank need to be driven in some way. Typically, this is by a series of lead screws connected to geared electric motors. The leaves are fitted with a small captive nut in which the lead screws fit, and the electric motors are fixed on a mounting plate directly behind the leaves. Rotation of the leadscrew by the motor therefore creates a linear movement of the leaf. The leaf drive motors are inevitably wider than a single leaf thickness, so in order to be able to drive each leaf the motors have to be mounted in a particular pattern as shown in FIG. 1. This shows a housing 10 for an array of adjacent MLC leaves 12. Behind the array, a motor mount 14 is fixed in place to housing 10 via bolts 16 so that it lies behind the leaves 12. A motor 18 for each leaf 12 is fixed to the motor mount 14. Each motor 18 is generally tubular and from one end (as shown in FIG. 1) therefore appears circular. The motors are wider than an individual leaf and are therefore arranged in a staggered pattern. In this example, the motors 18 are arranged in four offset rows so that the centre of a motor is aligned with each leaf. As a result of this, the leadscrew nuts therefore have to be fixed to the leaves in one of a variety of positions, meaning that (in this case) four different leaf shapes need to be manufactured. In an alternative system referred to as the “Beam Modulator” and shown generally in FIG. 2, leaves are driven by a rack and pinion system. A gear rack 20 is machined into the top or bottom of the leaves 22 and is driven by motors 24 fixed to the side of the leaf bank. The motor gear pinions 26 are mounted to an extension shaft 28 of a suitable length to enable the drive to be carried across to the appropriate leaf to be actuated. In our earlier patent application GB-A-2423909, we describe a modular design similar to the Beam Modulator drive system. The application describes a design where a system of miniature gears and racks are incorporated into a detachable module. The linear motion is transmitted to the leaf via a slotted feature in the rack and engages in a leaf drive coupling fitted to the rear of the leaf. The choice of drive system is influenced by the quantity and thickness of the leaves in the leaf bank. For example, the MLC leaf bank has 40 leaves per side and has an average leaf thickness of 3.6 mm. This thickness and number of leaves allows for a conventional solution of placing the motors directly behind the leaves and actuating them via a leadscrew which passes through the centre of the leaf. The diameter of the leadscrew in this design is limited to 2.5 mm, as this is largest diameter that can pass into the leaf without interfering with neighbouring leaves. Conveniently, it is also a standard ISO thread size. The leadscrew has to drive a leaf weighing around 800 g, and at certain head/gantry angles the full weight of the leaf is suspended by the thread alone. Due to the small engagement area of the thread, the leadscrew therefore experiences high frictional loads and requires regular lubrication to maintain an acceptable service life. The performance of the leadscrew is also adversely affected by a whipping motion that can arise when the leaf nut is close to the motor, in which the long free end of the leadscrew can oscillate as it rotates. In addition, the leadscrew experiences a buckling load when the leaf is pushed to the far end of the leadscrew. There is also a certain degree of noise due to this motion of the leadscrew. The Beam Modulator design employs a thinner leaf in order to increase the resolution of the leafbank. This leaf thickness of only 1.75 mm influences the selection of the drive system. A lead screw system as used on the MLC would not be a viable solution as it would require a 1.5 mm diameter leadscrew; as the leaf travel is longer, the leadscrew would suffer increased whipping and buckling. Leadscrews with a high aspect ratio are also extremely difficult and costly to manufacture and are likely to fracture if they are not adequately supported. In addition, the number of motors required (40 per side) could not be fitted in behind the leaves due to their size. The drive system for Beam Modulator therefore incorporates a rack and pinion system, with the motors disposed on either side, top and bottom of the leaf bank. The motors are fixed to the side of the leaf bank, and pinions are driven from the motors on extension shafts requiring 10 different lengths, in addition a staggered bearing block is incorporated in which the extension shafts runs. 8 such bearing blocks are required for the leaf bank. Because the motors are dispersed along the 4 sides of the leaf bank, the bank has to be removed for motor servicing. Removal of the leaf bank is a lengthy process, and problems can occur with radiation performance if the leaf bank is not replaced in the same position. The rack is machined into the top or bottom of the tungsten leaf; the bearing surface that would be positioned at the top of the leaf therefore has to be offset in order to make way for the rack. This has the undesired effect of reducing the shielding effect of the leaf, as some 8 mm is lost off the top/bottom of the leaf for the rack and bearing surface. In order for smooth operation of the rack a certain amount of clearance has to be maintained between the rack and pinion. Each of the 80 motors therefore has to be checked when assembling the leaf bank. This clearance can vary leaf to leaf, depending on manufacturing tolerances, and can lead to unwanted backlash once the pinion and motor gearbox begin to wear. Such backlash will affect the positional accuracy of the leaves. GB-A-2423909 describes a removable module which alleviates many of the service issues problems experienced with the beam modulator. However, as it incorporates a rack and pinion system it will suffer from backlash in the same way. The MLC Rack and Pinion System was originally designed around a 160 leaf MLC, but limitations in available space in the treatment head above and below the leafbank as well as restrictions on the overall head diameter create problems for fitting this type of Actuator. The gear racks in the actuator are positioned to match the leaf pitch; during operation the racks extend into the radiation beam, which may have effects on beam performance—particularly if there is an error in the pitching. The Actuator module also contains a high part count, including many precision cut gears and racks making this expensive to produce. Thus, the leaf thickness/pitch and motor size affects the method in which the actuation is carried to the leaf, and once a suitable method is derived (of the 2 practical drive solutions, leadscrew and rack and pinion) the design can have inherent problems with wear, noise, production and assembly costs, backlash and servicing issues. The present invention therefore seeks to provide a compact MLC actuator, that addresses many of the problems associated with a conventional leadscrew system, with the potential to drive a greater number of leaves without relying on a complex drive design and a high part count (relative to the number of leaves). This has the benefit of reducing production costs and assembly times. The drive mechanism should ideally not reduce the shielding effect of the tungsten leaves or interfere with the radiation beam. A modular design would also improve servicing issues by allowing the complete removal of the drive system from the leafbank. The MLC actuator of the present invention is designed for use on a 160 leaf MLC, but can of course be applied to MLCs with a lesser or greater number of leaves. The drive will ideally be capable of moving the leaves faster than previous MLCs to offer better dynamic treatment therapies, and will be useable for MLCs with smaller width and/or pitch of the leaves of, say, 1.5 mm as compared to the 10 mm diameter of the drive motors even within a limited overall head height. The width above the leaves (i.e. on the source side) is generally smaller than that below the leaves, due to the tapered design of the leafbank. Therefore, any design should ideally encompass this difference in leaf width and available space without complicating the design and increasing the required numbers of component parts. The present invention therefore provides a multi-leaf collimator for a radiotherapy apparatus, comprising at least one array of laterally-spaced elongate leaves, each leaf being driven by an associated motor connected to the leaf via a drive means so as to extend or retract the leaf in its longitudinal direction, the drive means comprising a sub-frame on which at least a subset of the motors are mounted, the sub-frame being mounted at a location spaced from the leaf array in a direction transverse to the lateral and longitudinal directions, and including a plurality of threaded drives disposed longitudinally, each being driven by a motor and being operatively connected to a leaf thereby to drive that leaf. The threaded drives will typically be leadscrews, but other arrangements such as a ballscrew can be used. Mounting the drive motors in this way allows them to be distributed more space-efficiently, and allows the drive system to be modular, without requiring rack and pinion gears. To take advantage of the ability to distribute the motors in a more space-efficient manner, we therefore prefer that a plurality of the motors mounted on the subframe are mounted at a first longitudinal end, and the remainder of the motors mounted on the subframe are mounted at a second, opposing, longitudinal end. Those leadscrews not at an edge of the array are preferably neighboured on either lateral side by one leadscrew driven by a motor mounted at the same longitudinal end and a second leadscrew driven by a motor mounted at the opposite longitudinal end. This results in the motors being arranged in pairs with a gap between which provides space for mounting the motors. The pairs of motors can be arranged one above the other to allow the necessary clearances, meaning that the leadscrews will be mounted in the subframe at one of two spacings from the leaf, with laterally neighbouring leadscrews being mounted at alternating spacings. The leadscrews can be mounted within a bore in the subframe. Still greater space efficiency can be achieved by including a lower subframe, mounted at a location spaced from the leaf array in an opposite direction to that of the upper array and on which the remainder of the motors are mounted. This can be designed in a generally similar manner to that of the (upper) subframe, except as regards the leaf pitch which will need to be adjusted as a result of the varying inclination of the leaves. We prefer that half of the leaves are driven from the subframe and half are driven from the lower subframe. Adjacent leaves in the array can be driven alternately from the subframe and from the lower subframe. The leaves are preferably mounted in a machined guide thereby to allow longitudinal motion. The subframe(s) can be mounted on the guide. In this way, drive can be supplied to the leaves from an elongate edge thereof. This drive can be transmitted to the substantially radio-opaque leaves via a drive coupling attached to the rear of each leaf. This can be located outside the radiation beam and can therefore be of a lightweight non-radio-opaque material. The drive means can further include a threaded member on the leadscrew. This is preferably restrained from rotation around the leadscrew by the remainder of the operative connection between it and the leaf. One way of doing so is for the threaded member to urge a laterally extending lug, thereby to connect to the drive coupling. The lug can engage with a recess on the drive coupling, and can include laterally-spaced flanges positioned to lie adjacent the drive coupling to prevent rotation of the lug around the threaded member. The lug ideally has a reasonable length in a direction parallel to the threaded member, to prevent rotation around axes transverse to the threaded member. A length that is at least 50% of its length transverse to the threaded member will typically suffice. The lug can alternatively be held in a machined slot in the subframe; that slot can be machined with non-parallel sides to assist in guiding the lug in the light of the offset nature of the load that it needs to carry. If desired, a collimator can be provided with 160 leaves, for future expansion, but operated as an 80 leaf collimator for compatibility purposes, by grouping adjacent leaves (such as in pairs), each leaf of a group being identically oriented and driven in unison by the same drive means. The inherent limitation on the minimum length of the rack and pinion-type system is the number of motors mounted on the side of the module. For example, assuming that each module is designed to drive 40 leaves, that each motor is 10 mm in diameter and (therefore) spaced 14 mm apart in a double row, then the length of the module will have to be 14×(40/2), i.e. 280 mm, plus the distance over which the leaves are expected to travel. If we take a rough figure of 70 mm for this distance, this makes an overall length for the system of 350 mm. The minimum overall height will be the motor diameter plus the height of the rack, i.e. about 32 mm. A rack and pinion module when mounted on the leafbank will therefore increase the treatment head diameter significantly. The MLC actuator described herein features a lead screw that runs parallel to the leaf, which means that the length of the drive modules are shorter overall, as the leadscrew only needs to be a slightly longer than the required leaf travel. The overall length of actuator including motors can therefore be about 200 mm, with a height of about 24 mm. This however faces the difficulty noted above, i.e. that the leadscrew needs a minimum diameter in order to be economic to produce and sufficiently rigid in operation. For MLC arrays in which the individual leaf thickness falls close to or below this diameter, this raises difficulties in accommodating both the leadscrews and the motors that drive them. The MLC actuator described herein incorporates a leadscrew drive assembly which actuates the leaf indirectly via a lug which projects out from the drive assembly and engages with a drive coupling for the leaf. The leadscrews and lugs run in machined guide slots in a bearing block which both houses the lugs (etc) and provides mounting for the drive assemblies. It still remains, of course, that the leadscrews may be wider than the leaves, and it will usually be the case that the motors are wider. Accordingly, each leaf will (generally) only be a fraction of the width of its associated drive mechanism. An alternative way of viewing this is that laterally arrayed drive mechanisms will only be able to drive a fraction of the leaves. Therefore, a number of such arrays can drive all of the leaves, if the drive from each array can be transmitted to the leaves satisfactorily. A specific pattern of drive mechanisms is therefore needed in order to mount the leadscrews drives into a compact removable module. We have chosen to divide the drive to the leaves in a number of ways so as to distribute the drive mechanism arrays. First, leaves can be driven from their upper edge or their lower edge. This is defined by the convention that MLC arrays are usually described as having a top that is closest to the radiation source and a bottom that is closest to the patient. Such a convention is necessary since the MLC array is mounted in a radiation head that rotates around the patient, and therefore in use the array may take up any orientation. Thus, an upper subframe can carry half of the drive mechanisms and drive every other leaf, and a lower subframe can carry the other half to drive the remaining leaves. Next, each subframe can carry two rows of leadscrews, one above the other. The lugs associated with each leadscrew can be of a corresponding length. This spaces the motors and allows them to drive laterally adjacent leadscrews. Finally, the leadscrews do of course have two ends and can be driven from either. Accordingly, half the leadscrews in each subframe can be driven from the front (which we define as the end most distant from the beam) and half from the rear (defined correspondingly). These three binary divisions allow 23 combinations, i.e. each situationally identical drive means drives one in eight leaves. This division can be as follows: LeafSubframeRowBank 1*Lowerbottomfront 2Uppertopfront 3Lowertopfront 4Upperbottomfront 5Lowerbottomrear 6Uppertoprear 7Lowertoprear 8Upperbottomrear 9*Lowerbottomfront10Uppertopfront11Lowertopfront12Upperbottomfront13Lowerbottomrear14Uppertoprear15Lowertoprear16Upperbottomrear17*Lowerbottomfront18Uppertopfront19Lowertopfront20Upperbottomfront21Lowerbottomrear22Uppertoprear23Lowertoprear24Upperbottomrear25*Lowerbottomfront26Uppertopfront27Lowertopfront28Upperbottomfront29Lowerbottomrear30Uppertoprear31Lowertoprear32Upperbottomrear The precise pattern of the leadscrews, lugs, and guiding slots in the bearing block is derived from the angle and pitch of the leaf and the required space for the drive motor. Such a pattern can also allow the drive motor axis to match the leaf centre line, ensuring an efficient transfer of linear motion. By mounting the drive motors on the front and rear surfaces of the drive modules (upper and lower subframes) the area required to mount the drive motors can be dispersed over 2 faces. This also has the advantage of only requiring 2 sizes of drive mechanism, thereby maintaining a low parts count. Thus, the drive system is split into 2 modules; 2 per side, upper and lower. Each of these modules contains 40 motor/leadscrew drives, allowing for 80 leaves in total. Each module has 20 motors mounted on the front face and 20 on the rear face. The method for mounting of the motor/leadscrew drives is designed specifically to fit the pattern of machined slots in the modules. This leadscrew design incorporates a precision machined leadscrew with an Acme thread form. The leadscrew nut is injection moulded in a low friction plastic material, which allows the assembly to run quietly without lubrication. The leadscrew nut fits into the lug, and can be easily replaced by removing the motor assembly. The machined guide slots for the lugs can also be formed with non-parallel sides, and the lugs profiled correspondingly. Thus, viewed along the guide slot, the profile can be akin to that of a key for a cylinder lock. This provides non-vertical surfaces which act as bearings, removing from the leadscrew nut the side and moment loads which will occur in moving the mass of the tungsten leaf. On previous designs, these loads adversely affected the life of the nut. The leadscrew is also supported in this way, reducing both whipping and buckling tendencies. The guide slot profile may also feature a “V” or fir tree shape in the leg of the slot, which will increase the bearing surface area of the key and reduce friction. A lower portion of the lugs are exposed below the drive module. These sections engage into the top or bottom of a drive coupling for the leaf via a mating cut-out in the drive coupling. Referring to FIG. 3, this shows a single leaf and its associated drive. The tungsten attenuation portion 100 is relatively thin in a lateral direction in order to allow good resolution, is long in its longitudinal direction to allow a wide range of movement, and is deep in the beam direction to allow good attenuation of the beam. A front edge 102 of the attenuation portion 100 is curved in a generally known manner so as to provide a sharper penumbra. A rear edge of the attenuation portion 100 is vertical, and is joined to a drive coupling 104. The drive coupling 104 has one edge, in this case the upper edge, which is co-linear with the corresponding edge of the attenuation portion 100 except for a recess 106 into which a lug 108 fits snugly. The opposing edge of the drive portion 104 is rebated back from the corresponding edge of the attenuation portion 100 in order to reduce the overall weight of the device and to avoid interference with the drive mechanism on the other side. It will be apparent that the relative orientations of the attenuation and drive portions can be reversed to allow the leaf to be driven from the top edge (as shown) or from the bottom edge. The lug 108 fits snugly in the recess 106 of the drive coupling 104 but is not fixed in place. The lug 108 is however attached to a pair of cylinders 110, 112 through which a leadscrew 114 passes, and between which a leadscrew nut 116 is fixed. Thus, as the leadscrew 114 is rotated, the nut 116 is forced in one direction or another and takes with it the cylinders 110, 112, the lug 108, the drive coupling 104 and the attenuation portion 100. The cylinders offer rigidity to the structure retaining the leadscrew nut 116, and also offer lateral support to the leadscrew 144 to inhibit both whipping and buckling. Finally, at one end of the leadscrew 114, a motor 118 is provided in order to drive the leadscrew. Thus, by simple reversal of the orientations of the drive coupling 104 and/or the motor 118/leadscrew 114, two of the above divisions can be achieved. The remaining third division is achieved by substitution of a longer lug 108. Accordingly, the spatial distribution of the various drive motors is achieved with an exceptionally low parts count. FIG. 4 shows one leaf bank from one end. The side-by side (i.e. laterally arrayed) leaves 100 are supported at their top and bottom edges in a leaf guide (not visible). Counting the leaves from the left hand side of FIG. 4, the odd-numbered leaves are driven from their lower edge and the even-numbered leaves are driven from their upper edge. Thus, an upper subframe 120 carries leadscrews, lugs, motors etc for the even-numbered leaves and a lower subframe 122 carries leadscrews, lugs, motors etc for the odd-numbered leaves. Apart from dimensional issues relating to the divergent nature of the leaves 100, the two subframes are functionally and structurally identical. Within each subframe, for example the upper subframe 120, the first two leaves that are controlled (i.e. leaves 2 and 4) are connected via lugs 108 of varying lengths to a leadscrew running in a guide machined in the otherwise solid block that forms the subframe. These two guides are placed at differing heights so as to separate the motors 118. The next leaf (i.e. leaf 6) is then connected to a leadscrew at the same upper level as leaf 2. To provide sufficient space, the motor for leaf 6 is located at the other end of the subframe 120 and drives its associated leadscrew from its other end. The pattern then continues, so that the next leaf that is driven in a manner identical to leaf 2 is leaf 10. FIG. 5 shows one subframe, with the leaf bank and leaf guide removed. An array of motors 118 can be seen at one end, distant from the beam, and an opposing array of motors 124 can be seen at the other end, closest to the beam. The lugs 108 can be seen projecting from the guide slots 126; when this sub-assembly is replaced under (or over) the leaf array then these lugs will project into the recesses 106 of the drive portions 104 of the leaves 100. In this way, the drive mechanism can be easily removed for service, repair or replacement. FIG. 6 shows how the motors 118 are retained on the subframe 122. Each motor has a pair of flanges projecting outwardly in two opposed directions around a part (but not all) of the circumference of the motor 118. Fortuitously, there will be a pair of guide slots 126a and 126b either side of the motor 118 which contain a leadscrew that is driven from the other end of the subframe 122. Thus, the ends (at least) of these slots 126a and 126b will be empty, and thus a mushroom-head screw 128a and 128b respectively can be screwed into the end of these slots 126a and 126b by providing a suitable tapping in the ends of the slots. In this way, by rotating the motor 118 so that the flanges are located under the mushroom-headed screws, then tightening the screws, the motor 118 will be retained securely. To remove the motor 118, both screws can be loosened, and the motor rotated in the direction of arrow 130 to move the flanges clear of the screw heads and allow the motor to be withdrawn in the direction of arrow 132. In this arrangement, each screw will retain two motors, one on either side. This still permits individual motors to be removed, since the motors either side will still be retained by one screw, on their other side. This is generally preferable to providing each motor with a single flange and a single retaining screw; whilst this could be done, and would mean that each screw only held one motor, it would weaken the retention of the motors generally. There could of course be further layers of leadscrews and motors beyond the two illustrated. Although this will incur a cost in terms of a greater complexity, it will permit a still greater ratio of motor spacing to leaf thickness to be achieved. FIGS. 7 to 10 show alternative profiles for the lug and 108 and the guide slot 126 in which it slides. FIG. 7 shows the simplest option, a parallel-sided guide slot 126 formed in the subframe 122, with an enlarged root 134. The leadscrew 114 sits in the enlarged root 134 and is surrounded by the leadscrew nut 116. The lug 108 extends from the leadscrew nut 116, along the guide slot 126 and out of the subframe 122, to engage with the drive portion 104 of the leaf 100. This arrangement is obviously easiest to manufacture. However, it then requires the lug 108 to support the leaf 100 despite the fact that the centre of mass of the leaf 100 is offset from the line along which the lug 108 is driven. This will create a rotational moment on the lug 108 which will seek to rotate the lug 108 within the plane of the guide slot 126. This will create an uneven wear pattern on the lug 108, the leadscrew nut 116, and the leadscrew 114 and may be detrimental to the long-term performance of the drive mechanism. FIG. 8 therefore shows an adjustment to this design to alleviate this. The lug 108 is no longer parallel-sided, but includes a step 136 to one side part way along its length. The thickness of the lug 108 remains the same through the step; that is, the outward bulge 138 on one side is matched by a corresponding recess 140 on the other side. Matching formations are provided in the guide slot 126, to accommodate the outward bulge and to project into the recess. By providing a non-flat surface to the lug 108 and a corresponding shape to the guide slot 126, rotation of the lug 108 in the guide slot 126 is inhibited. Support for the lug 108 against rotation is provided by the interaction of the bulge 138 and the recess 140 with the corresponding formations in the guide slot 126. Some lubrication may be useful in these areas, and a coating of graphite is suitable. The arrangement shown in FIG. 8 is a simple and straightforward one which illustrates the concept. In practice, the bulges and recesses could be located elsewhere along the height of the lug 108/guide slot 126, and/or they could be duplicated so that multiple such formations are present. Where several such formations are provided, they could be oriented in the same direction, or in different orientations such as alternate directions or a mix of directions. FIG. 9 shows a further variation. In this arrangement, the lug 108 has a pair of adjacent bulges 142, 144 on one side, duplicated on the other side. Corresponding recesses are formed in the guide slot 126. This arrangement has the advantage of being symmetrical as compared to that of FIG. 8, and also avoids any narrowing of the lug 108 that might cause it to be weakened. FIG. 10 shows a further alternative. A pattern of recesses 146 are formed in the sides of the lug 108, in this case four on each side in two groups of two each. Corresponding bulges are provided on the internal surfaces of the guide slot 126. The shapes described above can be formed at the necessary scale by processes such as wire discharge machining. FIG. 11 illustrates an alternative embodiment which may be simpler to manufacture in that the potentially complex shapes illustrated in FIGS. 7 to 10 are avoided. In the above embodiments, the leafbank comprises a set of leaves that run in a leafguide, driven via separately attached drive couplings in the form of ‘tails’ that can be made of a lighter and cheaper material. A separate drive module uses guided ‘keys’ running in accurately machined slots, which fit into slots in the drive couplings. This allows the drive module to be removed and replaced very quickly. In the alternative embodiment, the keys are made with slots that fit over the edges of the slots in the drive couplings. It is therefore no longer necessary to constrain the keys against movement in their roll axis (around the axis of the leadscrew). This allows the drive couplings to be fitted with a looser tolerance, reducing manufacturing time and cost. This also allows the key drive profile to be greatly simplified. With the key restrained in roll, it is possible to use the leadscrew to constrain the key in pitch and yaw, eliminating the need for the sliding contact and complicated machining of the drive module. The key can be simplified in material and form, reducing cost further. Thus, referring to FIG. 11, a plurality of leaves 200 are provided in the usual side-by-side relationship. FIG. 11 shows a single leaf for clarity purposes, but this will be supplemented by many other leaves on either side—typically making up a bank of 20, 40 or 80 leaves in total on each bank. The leaves 200 are supported in a guide 202 which supports the upper edge 204 and the lower edge 206 of the leaves in slots 208 formed in the guide. The guide 202 can be fixed to one side of the radiotherapy beam so that the leaves 200 are extendable into the beam by sliding in the guide slots 208, thereby limiting the lateral extent of beam on that side to a desired shape. Alternatively, the guide 202 can be mounted on a moveable support, its position thereby being adjustable in rotation around the beam and/or longitudinally relative to the leaves so as to enable a wider range of adjustment of the leaf positions. A similar bank of leaves is usually provided on the opposite side of the beam in order to collimate the other lateral extent of the beam. The leaf 200 is illustrated in FIG. 11 in a partially advanced position, shown in solid lines, and a withdrawn position shown in dotted lines. The withdrawn position illustrated is one that lies beyond the normal fully retracted position, in which the leaf has been fully retracted and then withdrawn further so that is no longer supported by the guide slots 208. Such a position would only be reached during assembly, maintenance, or disassembly, but allows us to illustrate the construction of the leaf. Each leaf 210 is of a substantially radio-opaque material such as tungsten, and the drive couplings 212 can be of a lighter and less expensive material such as steel or aluminium. This allows the tungsten forward portion 210 to be projected into the beam, driven by a rearward drive coupling that never enters the beam and does not therefore need to be of a radiopaque material. The overall weight and cost of the unit is thereby minimised. The drive coupling 212 of each leaf 200 includes a rectangular cut-out section 214, visible more clearly in the dotted outline version of the leaf 200 shown in the withdrawn position. This receives a corresponding drive lug 216 that is threaded onto a leadscrew 218. The leadscrew 218 is, in turn, mounted in a subframe 220 and provided with a drive motor (not shown) in a pattern similar to that described above. In this embodiment, the leadscrew 218 is supported by the subframe 220 at either end. The drive lug 216 has an extent in the longitudinal direction (i.e. parallel to the leadscrew 218 and the leaf 200) of (for example) 10 mm or more, generally at least 50% of its extent transverse to the leadscrew 218. It is therefore constrained against rotation about axes transverse to the leadscrew 218. The drive lug 216 extends transversely away from the leadscrew 218 toward the cut-out 214 of the leaf tail 212. The lug 216 ends with an interface region that keys with the cut-out 214; in this example it comprises a solid rectangular section 222 that matches the rectangular cut-out 214 and (when assembled) fits into the cut-out 214. On either longitudinal side of the rectangular section 222, there are laterally-spaced flanges 224 that fit snugly either side of the leaf tail 212 and prevent the lug 216 from rotating around the leadscrew 218. Thus, the drive lug 216 is prevented from movement in all axes other that longitudinal translation along the leadscrew 218 as the leadscrew 218 rotates. This movement of the drive lug 216 will then cause a corresponding movement of the leaf 200. Through the use of the above-described embodiments, it is possible to produce a reliable 160-leaf multi-leaf collimator, that is a collimator with 80 leaves on each side of the beam. Current commercially-available large-aperture MLCs have a total of 80 leaves, i.e. 40 leaves per side as illustrated in FIG. 4, but the increased space efficiency achieved by the present invention allows this to be doubled by appropriate thinning of the leaves. This means that instead of a projected width at the isocentre of 1 cm, each such leaf will have a resolution of 5 mm—with an attendant improvement in resolution and accuracy of delivery. An improvement of the resolution to 160 leaves instead of 80 will also require improvements in the treatment planning systems and software, and the associated control systems and software in order to take advantage of the additional degrees of freedom offered by doubling the number of leaves. In the longer term, this does not present a particular difficulty, but in the short term clinics may wish to replace hardware and other systems incrementally. Accordingly, there may be advantages in an MLC that retains the ability to operate in a 160-leaf mode but which is fully compatible with 80-leaf control systems. This is indeed possible through the present invention. If the same leaves are inserted into the same leaf guide, but oriented so that they are organised in identical pairs, then these leaf pairs can be driven together, in unison, by providing suitable upper and lower subframes 120 as illustrated in FIG. 3 et seq. Adjacent leaf pairs will have co-located recesses 106 in their associated drive couplings, into both of which the same lug 108 can project. Some care may need to be taken in designing the appropriate width for the lug 108 to ensure that an adequate drive is transmitted to both leaves. Thus, the device will operate as an 80-leaf collimator and can be controlled and driven in the same way. However, as and when the clinic is able to upgrade other aspects of their radiotherapy equipment, the upper and lower subframes can be replaced with items adapted for 160-leaf operation and the leaves removed and re-inserted in the pattern appropriate to independent operation of each leaf. Another use of the described collimator drive is for a variable-pitch collimator. Such a collimator includes leaves having a plurality of different thicknesses, such as a group of narrow leaves in the central region flanked on either side by relatively thicker leaves. Thus, a fine resolution is available in the central area of the aperture where it is usually needed, but the full aperture of the MLC is available when needed. Such collimators are limited by (inter alia) difficulty in driving the various leaves accurately and the present invention can assist with this. It will of course be understood that many variations may be made to the above-described embodiment without departing from the scope of the present invention. Although the present invention has been described with reference to preferred embodiments, workers skilled in the art will recognize that changes may be made in form and detail without departing from the spirit and scope of the invention. |
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059819642 | summary | BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention pertains to the field of shielding structures for radiation procedure tables. In particular, the present invention pertains to adjustable X-ray shields for procedure tables and adjustable X-ray shielding systems using on-line radiation dosimetry. 2. Description of the Related Art During certain procedures, for example, those involving the insertion of an intravascular catheter (procedures such as atherectomy, balloon angioplasty, stent placement and the like), patients are required to lay prone on an X-ray procedure table. The physician then inserts a guidewire or a catheter device through a small incision, often near the patient's groin, and advances the device through an artery to the target site. To facilitate imaging of the intravascular device during use, the device may include a radio-opaque distal tip to allow the physician to guide the device under fluoroscopic observation to the desired site. During such procedures, the physician controls an X-ray tube, the source of the radiation used to visualize the target arterial site and the intravascular device, using a foot pedal. The X-ray tube is generally located underneath the table surface. The X-ray beam generated by the tube travels through the procedure table and the patient's body. The majority of the photons generated by the X-ray tube are scattered by the cells within the patient's body. A small portion of the generated X-ray beam, however, succeeds in traversing both the procedure table and the patient's body. A portion of the X-ray beam that traverses both the table and the patient's body reaches an image intensifier located above the patient. The image intensifier detects the photons, channels them through a photon multiplier and transfers the resulting intensified image to a display screen, which allows the physician to accurately guide the device. Interventional procedures such as described above often take hours to complete, and may expose the physician and the patient to a significant amount of radiation that may, over time, pose serious health risks. To reduce the physician's exposure to scattered X-ray radiation, a number of protective measures have been implemented. The first of such protective measures is the use of X-ray shields. Several types of X-ray shields have been developed. Lenhart, for example, in U.S. Pat. No. 5,006,718, proposes a shield assembly wherein shielding material extends from the side of the table adjacent the physician to the floor. However, such shield, while affording an unobstructed view of the patient, does not appear to protect the physician from the large amount of radiation scattered from the patient's body and above the level of the table. Other solutions generally involve the use of horizontally disposed strips of radiation shielding material, such as disclosed by Collica et al. in U.S. Pat. No. 3,984,696, or the use of vertically disposed strips of radiation shielding material hung from a position above the patient's body, as disclosed in Lenhart, U.S. Pat. No. 4,581,538 and Stivender et al., U.S. Pat. No. 4,062,518. However, both of these approaches appear to share common disadvantages. The first of these is that the physician's view of the patient, and the patient's view of the physician is severely impaired by the strips of shielding material. The second of these disadvantages appears to be an incomplete shielding of scattered radiation, in that significant portions of the patient's body are directly exposed to the physician. During the interventional X-ray procedure, a significant amount of scatter from these exposed portions may reach the physician, with deleterious health consequences. Moreover, with such conventional X-ray shields, unintended gaps in coverage may occur due to improperly positioned or repositioned shields, unknowingly exposing the physician to greatly increased radiation levels until the gap is discovered, if ever. Indeed, the shield often must be re-positioned during the procedure when the orientation of the X-ray tube is changed to change the imaging angle. Failure to properly reposition the shield thereafter may result in gaps in the shielding material. As X-ray radiation cannot be seen, heard or felt, such gaps may remain undiscovered and may expose the physician and other personnel in the procedure room to needlessly increased radiation levels. The second of such protective measures to protect the physician from unwanted scatter is the so-called lead apron. The lead apron is an article of clothing donned by the physician prior to a catheter procedure. The lead apron generally covers at least the neck including the thyroid, the torso, and the thighs. Leaded eyeglasses may also be worn. The apron is effective in significantly reducing the physician's instantaneous and cumulative exposure to X-ray radiation over time, at least with respect to the areas covered by the apron. In contrast, uncovered areas, such as the physician's hands, arms and lower legs, are protected, if at all, solely by the table's X-ray shield. Common lead aprons are equivalent to a lead barrier of about 0.25 to 0.5 millimeters in thickness. However, such protection, while effective, does not come without discomfort to the physician. Indeed, such lead aprons commonly weigh about fifteen to twenty pounds. As the physician may be required to bend over the patient for periods of time often measured in hours, a significant proportion of the weight of the lead apron may be borne by the physician's lower back, shoulder and neck area. Supporting such weight can, over time, cause intense lower back, neck and shoulder pain, particularly for those suffering from already weakened backs. What is needed, therefore, is an X-ray shield for an X-ray procedure table that more effectively protects the physician from X-ray scatter. What is also needed is an X-ray procedure shield for an X-ray procedure table that is highly effective in attenuating X-ray scatter while affording the physician a relatively unobstructed view of the patient during the X-ray procedure. What is also needed is an X-ray shield that is sufficiently effective in attenuating radiation to allow the physician to wear a lighter apron (e.g. one having a lower lead equivalency rating) or to allow the physician to safely forego wearing a lead apron or other uncomfortable protective gear altogether. Also needed are means for insuring, at all times, the proper positioning of the X-ray shield to achieve the greatest possible attenuation of radiation. SUMMARY OF THE INVENTION In accordance with the above-described objects and those that will be mentioned and will become apparent below, the adjustable X-ray shield for an X-ray procedure table, according to an embodiment of the present invention comprises: a main support bar mounted substantially vertically on the table near a patient's waist level, the main support bar extending above and below a surface of the table; PA1 a first transversal support bar, the first transversal support bar being attached to the main support bar just below the table and extending substantially across a width of the table; PA1 a second transversal support bar, the second transversal support bar being pivotally mounted to the main support bar above the table and extending substantially across a width of the table; and PA1 a first longitudinal support bar, the first longitudinal support bar being pivotally mounted to the main support bar above the table and extending along a length of the table past the patient's chest level. PA1 a pair of center pivots, each attached near respective ends of the first transversal support bar; a pair of rectangular blocks, each pivotally attached to respective ends of each of the center pivots closest to an underside of the table; and a pair of threaded cylinders, each attached to respective other ends of each of the center pivots and disposed within respective threaded through bores in each of the main support bar and outer support tube. Tightening the pair of threaded cylinders within the threaded through bores causes the pair of rectangular blocks to exert a force on the table to secure the X-ray shield thereto. PA1 a main support bar, mountable substantially vertically on the table near a patient's waist level, such that the main support bar extends above and below a surface of the table; PA1 a first transversal support bar, attachable to the main support bar just below the table to extend substantially across a width of the table; PA1 a second transversal support bar, pivotally mountable to the main support bar above the table to extend across a width of the table; PA1 a first longitudinal support bar, pivotally mountable to the main support bar above the table to extend along a length of the table past the patient's chest level; and PA1 at least one sheet of radio-opaque shielding material attachable to at least one of the first transversal support bar, the second transversal support bar and the first longitudinal support bar. PA1 a first sheet of radio-opaque shielding material adjustably disposed alongside the procedure table, from a distance above a surface of the table to a floor surface; PA1 a second sheet of radio-opaque shielding material adjustably disposed across and above the surface of the table at a level of a patient's waist; and PA1 a third sheet of radio-opaque shielding material disposed across a width of the table from just underneath the table to about the floor surface, the first, second and third sheets of radio-opaque shielding material being supported by an adjustable frame assembly attached to the table. According to other exemplary embodiments, at least one sheet of radio-opaque shielding material is attachable to the first transversal support bar, the second transversal support bar and/or the first longitudinal support bar. The main support bar may include an outer support tube portion below the table; an upper support tube portion above the table; and a telescoping support bar slidingly mounted at least within the upper support tube, the telescoping support bar allowing adjustment of a height of the second transversal and first longitudinal support bars above the surface of the table. The first transversal support bar may include a telescoping extender bar to accommodate X-ray procedure tables of varying widths. The first longitudinal support bar may comprise an articulated joint. The adjustable X-ray shield may further comprise a first sheet of shielding material attached to the first transversal support bar that substantially spans the width of the table from an underside thereof to a floor surface. A second sheet of radio-opaque shielding material may be attached to the second transversal support bar adjacent the main support bar, the second sheet shielding material draping alongside the patient's pelvic and thigh areas to attenuate X-ray scatter therefrom. A third sheet of radio-opaque shielding material may be attached at least to the second transversal support bar, the third sheet of shielding material being of a generally rectangular shape and having a cutout portion for the patient's waist. The sheets of radio-opaque shielding material may provide overlapping coverage along the vertical axis of the main support bar. The main, first transversal, second transversal and first longitudinal support bars may be formed of metal and/or a radiolucent material. The radiolucent material may be a carbon fiber material. According to other embodiments, the X-ray shield may be removably secured to the table by means of a clamping assembly. The clamping assembly may comprise at least two mounting pads to be positioned on the table surface and a bell crank mechanism attached underneath the table surface to the first transversal support bar. A first end of the first transversal support bar may be attached to the main support bar and a second end thereof may be attached to an outer support tube. The bell crank mechanism may comprise: According to still further embodiments, the second transversal and first longitudinal support bars may be pivotally attached to the main support bar by respective clevis joints having respective pivot pins. In that case, tightening the respective pivot pins increases the friction in the respective clevis joints. A first end of the first transversal support bar may be attached to the main support bar and a second end thereof may be attached to an outer support tube and the X-ray shield may further comprise a second longitudinal support bar attached to the outer support tube such that the second longitudinal support bar is parallel to the first longitudinal support bar and disposed on an opposite side of the table; and a further sheet of radio-opaque shielding material attached to the second longitudinal support bar, to attenuate X-ray scatter on the opposite side of the table. According to another embodiment, at least one of the sheets of radio-opaque shielding material may comprise one or more radiation sensors. One or more of the sensors may be a semiconductor X-ray sensor, such as a Cadmium-Zinc-Telluride (CdZnTe) sensor in communication with a data processing and display device. A plurality of X-ray sensors may be attached to each of the one or more sheets of radio-opaque shielding material, to form an array of X-ray sensors, the array of sensors being connected to a data processing and display device to allow real time monitoring of X-ray radiation from the X-ray shield. One or more of the sensors may be attachable to the patient's back or to the table, to measure the patient's entrance dose of radiation. Moreover, one or more of the sensors may be attachable to a person other than the patient, such as the physician and/or other care givers, to measure scatter received by that person during a procedure on the patient. According to another embodiment, the present invention is a kit for an adjustable X-ray shield for a procedure table, comprising: According to another embodiment of the kit according to the present invention, the sheet or sheets of radio-opaque shielding material may comprise at least one X-ray sensor attached thereto. According to a further embodiment, the present invention is an X-ray shield for a procedure table, comprising: According to a still further embodiment, a procedure table radiation shield having on-line radiation dosimetry according to the present invention comprises a plurality of sheets of radio-opaque shielding material adjustably disposed on a frame assembly so as to protect a physician from radiation scatter during a radiation procedure, at least one of the plurality of sheets of radio-opaque shielding material including at least one radiation sensor. According to further embodiments, the array of radiation sensors may include at least one X-ray sensor, which may be a semiconductor X-ray sensor. Such a semiconductor X-ray sensor may be a CdZnTe sensor. The radiation sensor or sensors may be connectable to a data processing and display apparatus via a communication channel to allow the physician to view real time radiation readings from the display apparatus and to adjust a position of the plurality of sheets of radio-opaque shielding material in response to the readings to minimize exposure to radiation scatter during a radiation procedure. One or more of the radiation sensors may be attached to one of the sheets of radio-opaque shielding material and one or more of the radiation sensors may be adapted to be positioned anywhere within the procedure room or on a person within the procedure room. |
046684654 | claims | 1. Apparatus for transmitting signals generated by a plurality of detectors located in a hazardous environment enclosed within a containment structure to the outside of the containment structure, said apparatus comprising: means within the containment structure for generating a plurality of redundant sets of digital detector signals; a plurality of redundant communications buses located within the containment structure; interface means within the containment structure for applying each set of the redundant digital detector signals to each of the redundant communications buses; a plurality of bus controller/serial output devices in the containment structure, each of which is connected to one of said communications buses for controlling said interface means in applying said redundant sets of digital detector signals to the associated communications bus, and for serially outputting the redundant sets of digital detector signals received over the associated communications bus; serial data link means connected to each of said bus controller/serial output devices within the containment structure and extending through the containment structure to the outside thereof; and receiver means outside of said containment structure connected to said serial data link means, for receiving said redundant sets of digital detector signals serially transmitted over said data link means and for generating therefrom representations of the value of the detector signals. a plurality of detector means associated with the reactor within the containment structure each generating first and second redundant digital detector signals representative of a selected operating parameter of the reactor; first and second communications buses located within the containment structure; interface means within the containment structure for applying each of said first and second digital detector signals to each of said communications buses; first and second bus controller/serial output devices each connected within said containment structure to one of said communications buses, for controlling said interface means in applying said redundant digital detector signals to the associated communications bus, and for serially outputting the redundant detector signals received on the associated communications bus; first and second data link means each connected in the containment structure to one of said bus controller/serial output devices and extending through the containment structure to the outside thereof; and first and second receiver means connected to the first and second data link means for generating from the redundant digital detector signals representations of the value of the detector signals. a containment structure; a reactor core located inside the containment structure and having a plurality of drive rods for positioning neutron absorbing rods within the reactor core to control the reactivity thereof; a plurality of detector means each of which is located within the containment structure and generates redundant first and second multi-digit digital detector signals representative of the position of one of said drive rods; first and second communications buses located within said containment structure; interface means within said containment structure for applying each of said first and second digital detector signals to each of said first and second communications buses; first and second bus controller/serial output devices connected to the first and second communications buses respectively for controlling the interface means to apply sequentially all of said first and then all of said second digital detector signals to the associated communications bus and for serially outputting the first and second digital detector signals received over the associated communication bus; first and second data links connected respectively within the containment structure to each of the first and second bus controller/serial output devices and extending through the containment structure to the outside thereof; and first and second receiver means outside of the containment structure each selectively connected to the first data link from one bus controller/serial output device and the second data link from the other bus controller/serial output device for generating representations of the position of said drive rods from the first and second digital detector signals received over the selected data link. generating inside of the containment structure a plurality of redundant sets of signals representative of the measurement of each selected parameter; storing inside the containment structure in each of a plurality of microprocessors equal in number to the number of redundant sets, said plurality of redundant sets of signals; serially transmitting said plurality of redundant sets of stored signals from each microprocessor inside said containment structure to the outside of said containment structure over its own serial data link passing through said containment structure; and generating separately outside of said containment structure, each from the redundant sets of signals transmitted over one of said serial data links, redundant representations of the value of each of said selected parameters. 2. The apparatus of claim 1 wherein said receiver means includes a plurality of receiver devices and means for selectively connecting each of said receiver devices to one of said serial data link means to receive the redundant sets of digital detector signals from one of said communications buses through the associated bus controller/serial output device. 3. The apparatus of claim 2 wherein said receiver devices include means for selecting for each detector signal, one of said redundant digital detector signals and means for generating for each detector said representation of the value of the detector signal from selected redundant digital detector signal. 4. Apparatus for remotely monitoring the operation of a nuclear reactor enclosed in a containment structure, said apparatus comprising: 5. The apparatus of claim 4 wherein said receiver means each include means for selectively receiving the redundant digital detector signals from one of said first and second data link means. 6. The apparatus of claim 5 wherein said receiver means each include means for selecting for each detector signal one of said first and second redundant digital detector signals received through said one data link means. 7. The apparatus of claim 4 for use in monitoring the position of the drive rods which position neutron absorbing rods in the core of the reactor with each detector monitoring the position of one drive rod and wherein said detector means each include: a plurality of sensor located along the travel path of the associated drive rod with each sensor generating a sensor signal representative of the position of the drive rod relative to the sensor, and processing means for generating a multi-digit detector signal from said plurality of sensor signals. 8. The apparatus of claim 7 wherein said processing means includes means for reproducing the multi-digital signal to generate said first and second redundant digital detector signals. 9. The apparatus of claim 7 wherein the detector means includes first and second sets of interleaved sensors, first processing means for generating said first multi-digit digital detector signal from signals generated by the first set of sensors and second processing means for generating said second multi-digit digital detector signal from signals generated by the second set of sensors. 10. A nuclear power plant comprising: 11. The apparatus of claim 10 wherein each detector means includes a plurality of electrical coils positioned along the travel path of the associated drive rod for generating a plurality of electrical signals indicative of the position of the drive rod, processing means for generating said first and second multi-digit detector signals from said electrical signals, said processing means being spaced from the reactor core and said electrical coils, and electrical leads for transmitting said electrical signals from the coils to the processing means. 12. The apparatus of claim 11 in combination with a radiation barrier inside said containment structure between said reactor core and electrical coils on one side of the barrier and said processing means, communications buses, interface means, bus controller/serial output devices and data link means all on the other side of the barrier. 13. The apparatus of claim 12 wherein each receiver means includes means for selectively generating said representations of the rod positions from the first and second multi-digit digital detector signals received through only one of said data links. 14. The apparatus of claim 13 wherein each receiver means includes means for selectively, as to the position of each drive rod, generating said representation of position from one only of said first and second digital detector signals received through said one data link means. 15. The apparatus of claim 14 wherein one of said receiver means is a display device which generates a visual display of rod position from the multi-digit digital rod position signals and the second receiver means is a plant computer for monitoring operation of the nuclear power plant and includes means for generating representations of rod position from said multi-digit digital detector signals. 16. A method of transmitting reliable signals representative of the value of a plurality of selected parameters in a process carried out in a high radiation environment inside of a containment structure to the outside of said structure with a minimum number of penetrations through the structure, comprising the steps of: 17. The method of claim 16 wherein the step of generating redundant representations includes as to each representation the step of selecting from among the data links, the one from which said plurality of sets of signals is to be taken. 18. The method of claim 17 wherein the step of generating redundant representations includes as to each representation, selecting separately as to each parameter the signal to be used for generating the representation of that parameter from among the plurality of redundant signals transmitted over said one data link. 19. The method of claim 18 wherein said selected parameters are each of the position of a drive rod which positions control rods within the core of a nuclear reactor located within the containment structure and wherein the step of generating a plurality of sets of redundant signals comprises generating first and second sets of redundant multi-digit digital signals each representative of the position of the associated drive rod. 20. The method of claim 19 wherein the step of generating a representations of rod position comprises: generating a multipage visual display with a plurality of said pages each including analog visual representations of the position of selected groups of rods generated from said multi-digit digital signals with each of said plurality of pages displaying different rod groups, and with each of said plurality of pages also displaying digital representations of selected status conditions of all of said groups of rods. |
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abstract | The reflectivity and thermal stability of Mo/Si (molybdenum/silicon) multilayer films, used in soft x-ray and extreme ultraviolet region, is enhanced by deposition of a thin layer of boron carbide (e.g., B4C) between alternating layers of Mo and Si. The invention is useful for reflective coatings for soft X-ray and extreme ultraviolet optics, multilayer for masks, coatings for other wavelengths and multilayers for masks that are more thermally stable than pure Mo/Si multilayers |
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06078640& | summary | FIELD OF THE INVENTION AND RELATED ART This invention relates to an X-ray exposure apparatus for use in the manufacture of semiconductor devices, for example. Miniaturization of circuit patterns has been enhanced for an increase of operation speed and integration of a solid device such as LSI. For formation of a circuit pattern in LSI manufacturing processes, as a future exposure technique, a fine pattern forming process which is based on high luminance X-rays from a synchrotron orbit radiation (SOR), for example, is attractive. Generally, X-ray exposure methods include a proximity unit-magnification X-ray exposure method which uses those of soft X-rays having a wavelength of 0.5-2 nm, and a reduction projection exposure method which uses a reflection type mask and a wavelength of 4-20 nm. An example of the former is proposed in Japanese Laid-Open Patent Application, Laid-Open No. 100311/1990. In this method, because of a short exposure wavelength, in principle, there is a possibility of high resolution not larger than 0.1 micron. The proximity unit-magnification X-ray exposure method uses a transmission type mask, called a "unit-magnification X-ray mask". The portion of such a unit-magnification X-ray mask through which X-rays are transmitted is made of a light element material such as SiN.multidot.SiC, called a "membrane". It comprises a thin film having a thickness of about 2 microns, and a size 35 mm square. As for an X-ray absorbing portion of the unit-magnification X-ray mask, a circuit pattern made of heavy metal such as Ta or W, called an "absorptive material", having a thickness of about 0.5-1.0 micron, is formed on the membrane. Also, an optical system in the proximity unit-magnification X-ray exposure method serves to enlarge X-rays from a light source into a predetermined field size by use of an X-ray mirror, and through the X-ray mask it serves to transfer the pattern onto a wafer substrate, disposed opposed to the X-ray mask. However, an apparatus to be used in the proximity unit-magnification X-ray exposure method such as described above differs from a conventional optical exposure apparatus using ultraviolet rays, for example, and it is difficult to variably set the exposure transfer magnification, to be transferred onto a wafer substrate, by use of the X-ray optical system. More specifically, it is difficult to correct magnification with the X-ray optical system, whereas the magnification in an optical exposure apparatus can be corrected with an optical system having lens groups. As for correction of magnification in X-ray exposure apparatuses, a method has been proposed in which a mask substrate is positively deformed to thereby control the pattern magnification on a membrane. For example, an example wherein a plane stress is applied to a mask substrate is disclosed in Japanese Published Patent Application, Publication No. 66095/1992. In this method, a clamping mechanism for a mask chuck is provided with mechanical means (electrostrictive device) and, as a result of application of deformation to a mask supporting frame, the mask substrate on the supporting frame is elongated or contracted. With the pattern magnification correcting mechanism for an X-ray exposure apparatus such as described above, however, a stress along a plane or not along a plane is applied to the mask to press the mask against a reference surface, for example, to thereby cause deformation of it and correct the pattern magnification. This causes positional deviation of the pattern, and it raises a problem of degradation of pattern registration precision. SUMMARY OF THE INVENTION It is an object of the present invention to provide an X-ray exposure apparatus by which high precision exposure is assured regardless of mask magnification correction. These and other objects, features and advantages of the present invention will become more apparent upon a consideration of the following description of the preferred embodiments of the present invention taken in conjunction with the accompanying drawings. |
claims | 1. An x-ray apparatus which includes an X-ray source for producing X-rays, an X-ray detector for detecting the X-rays, a filter which is arranged between the X-ray source and the X-ray detector, which filter includes a plurality of tubular filter elements having a longitudinal direction z and a circumference, the X-ray apparatus also including an electrical device which is provided with at least one power supply source for controlling the individual filter elements, wherein each filter element is constructed with an internal volume for receiving a liquid filling having electrically conductive and X-ray absorbing properties, and wherein each filter element displays an X-ray absorptivity which is dependent on the quantity of X-ray absorbing liquid present in the internal volume, wherein each filter element includes a first electrode for applying a first electric potential to a wall of the filter element, and a second electrode for applying a second electric potential to the internal volume of the filter element, wherein the X-ray absorptivity of each filter element is adjustable by step-wise control of a level of the X-ray absorbing liquid in the longitudinal direction z of the filter element, and wherein the first electrode includes a series of electrode segments which extend at least over a part of the circumference of the filter element and succeed one another in the longitudinal direction z of the filter element, and said directly successive electrode segments are electrically insulated from one another and individually controllable by the electrical device. 2. A filter for use in the X-ray apparatus as set forth in claim 1 , wherein the electrode segments are subdivided into at least two sub-groups, each sub-group including at least two not directly successive electrode segments, the electrode segments in each sub-group being electrically interconnected. claim 1 3. The filter as set forth in claim 2 , wherein the electrode segments bear even sequence numbers in the series and the electrode segments bear odd sequence numbers in the series constitute respective sub-groups, the electrical device being provided with switches for controlling said sub-groups. claim 2 4. A filter for use in the X-ray apparatus as set forth in claim 1 , wherein each electrode segment is connected to the electrical device via a respective connection. claim 1 5. A filter for use in the X-ray apparatus as set forth in claim 1 , wherein facing edges of directly successive electrode segments are provided with meshing teeth. claim 1 |
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abstract | A specimen holder is offered which can reduce the amount of chemical sprayed over a specimen consisting of cultured cells. The specimen holder has an open specimen-holding surface. At least a part of the specimen-holding surface is formed by a film and a tapering portion formed around the film. The specimen can be cultured on the specimen-holding surface of the film. The presence of the tapering portion can reduce the amount of used reagent. The specimen can be irradiated via the film with a primary beam for observation or inspection of the specimen. Consequently, the specimen, such as cells, can be well observed or inspected in vivo while the specimen is being cultured. Especially, if an electron beam is used as the primary beam, the specimen can be well observed or inspected in vivo by SEM (scanning electron microscopy). |
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claims | 1. A system for fabricating a radiation-absorbing semiconductor substrate, comprising:a processing chamber having an inlet port for introducing a gas into the chamber,a movable substrate holder disposed in said chamber and configured for holding a semiconductor substrate,a motion controller coupled to said substrate holder for moving the holder,a window coupled to the chamber for allowing passage of radiation into the chamber for processing said semiconductor substrate,a laser radiation source for generating laser pulses, said source being optically coupled to said window such that the laser pulses impinge on a surface of the semiconductor substrate,a lens for focusing said laser radiation pulses onto the semiconductor substrate surface,a camera positioned outside the chamber,a mirror configured to redirect the laser radiation after its passage through the lens onto the camera when the mirror is positioned in the path of the laser radiation,wherein said camera is positioned at a distance from the lens that is substantially equal to a distance of the lens from the substrate surface, thereby providing a measure of a spotsize of the radiation pulses on the substrate surface, andwherein said controller is capable of moving the substrate while the surface of the substrate is irradiated by said radiation pulses in presence of the gas in the chamber. 2. The system of claim 1, wherein said substrate holder is configured for movement in at least two orthogonal directions. 3. The system of claim 1, wherein said laser radiation source generates radiation pulses having a duration of about 50 femtoseconds to about a few ns. 4. The system of claim 3, wherein said laser radiation source generates radiation pulses having duration of about 50 femtoseconds to about 50 picoseconds. 5. The system of claim 1, wherein said focused radiation pulses exhibit a fluence in a range of about 1 kJ/cm2 to about 8 kJ/cm2 at said semiconductor surface. 6. The system of claim 1, wherein said laser radiation source comprises an amplified, Ti:Sapphire laser system. 7. The system of claim 1, wherein said motion controller provides a micrometer precision in moving the substrate holder. 8. The system of claim 1, further comprising a roughing pump coupled to the chamber for evacuating the chamber. 9. The system of claim 1, wherein a repetition rate of said laser pulses and a speed by which said controller moves the substrate holder are configured such that each location of the irradiated surface is exposed to a number of laser pulses in a range of about 2 to about 2000. 10. The system of claim 1, further comprising a viewport for viewing the substrate in the chamber. 11. The system of claim 1, further comprising a single axis translation stage onto which the lens is mounted for moving the lens relative to the substrate surface so as to vary a spotsize of the laser radiation pulses on the substrate surface. 12. The system of claim 2, wherein said substrate holder is configured for movement along a third direction orthogonal to said at least two directions. 13. The system of claim 1, wherein said mirror comprises a flipper mounted mirror. 14. The system of claim 1, wherein said camera comprises a CCD camera. 15. The system of claim 1, further comprisinga source of visible radiation optically coupled to said window of the chamber for illuminating the substrate surface. 16. The system of claim 15, wherein said camera comprises a CCD camera. |
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053708270 | description | DETAILED DESCRIPTION OF THE INVENTION The present invention is directed to methods for treating various solutions which are contaminated with soluble and insoluble inorganic (including radioactive) species. While this process will be described primarily for removal of metals like uranium, thorium, lead, mercury, copper, cesium, barium, cadmium, and mixtures thereof, it is also suitable for removal of radium, arsenic, boron, chromium, silver, selenium, beryllium, nickel, antimony, molybdenum, vanadium, zinc, thallium, strontium, cobalt, plutonium, and the like. As used herein, the term "process stream" includes all forms of solutions in which contaminates may be found, such as, for example, groundwater, drinking water, soil washing extractants, leachates, effluents, etc. It also specifically includes uranium/thorium-containing waste streams produced by nuclear fuel handling/manufacturing facilities, and by mining facilities. The term "silicate" as used herein refers to the salt of silicic acid, or any compound that contains silicon, oxygen, and one or more metals, and may contain hydrogen. Silicates further include any of a group of minerals whose crystal lattice contains SiO.sub.4 tetrahedra, either isolated or joined through one or more oxygen atoms. The term "age" as used herein means to cause or allow to stand for a certain period of time (with or without stirring), or until certain transformations have taken place, namely, until the contaminant-containing silica matrix has gelled, polymerized and/or precipitated. The term "gel" as used in the present invention includes the formation of any jelly-like colloid, solution or suspension. For purposes of the invention described herein, "precipitates" include coprecipitates, and vice versa, and both terms include any substance precipitated from solution. The method of the invention begins by treating the contaminated process stream with precipitants, specifically silicate and ammonium hydroxide. In solution, the silicate becomes amorphous silica which has a large surface area and a high reactivity. It is believed that the amorphous silica reacts with the metal contaminants in solution by providing adsorption or exchange sites which capture the metal contaminants, thereby forming amorphous silica precipitates. The process stream may be treated with any suitable silicate known to the those skilled in the art, including, for example, sodium silicate, potassium silicate, tetraethylorthosilicate, tetramethylorthosilicate, or a mixture thereof. Preferably the stream is treated with sodium silicate having the formula: EQU Na.sub.2 O*nSiO.sub.2, where n=2 to 3.5 As the value of n increases, the silicate level increases, thereby increasing the likelihood that precipitation will occur. If n is outside of the these ranges, precipitation will not occur as efficiently, that is, more reactant will be required to accomplish the job or too much waste will be generated. Preferably, the sodium (or potassium) silicate is added as a liquid which comprises from about 1% to about 50% sodium (or potassium) silicate by weight, more preferably from about 2% to about 35%, even more preferably from about 2% to about 10%, and most preferably from about 4% to about 8%. The amount of silicate to be added is determined by the condition of the stream to be treated. Preferably, the amount of silicate added should be from about 0.5 to about 250 g/L of stream to be treated, more preferably from about 1 to about 100 g/L, and most preferably from about 5 to about 25 g/L. As shown in Table 2, increasing the amount of silicate (for example, from 5 to 15 g/L of solution to be treated) reduces the aging time required for successful gelation, polymerization and/or precipitation, as well as the filtration time. TABLE 2 ______________________________________ Effect of Silicate Concentration on Aging Time Required for Successful Filtration Filtration Silicate Aging Time, Filtration Level, g/L hr Time Contamination ______________________________________ 5 30 minutes 24 sec Highly Contaminated 5 180 minutes >2 hours Low (<1 ppm) 5 300 minutes 23 minutes Low (<1 ppm) 10 75 minutes 60 minutes Low (<1 ppm) 10 120 minutes 26 minutes Low (<1 ppm) 10 180 minutes 15 minutes Low (<1 ppm) 15 75 minutes 10 minutes Low (<1 ppm) ______________________________________ Although treatment with silicates will significantly reduce the solubility of the contaminants in the stream, it generally will not be adequate to precipitate the contaminants to a degree which will permit collection and removal of contaminants to environmentally acceptable levels. Due to the nature of the precipitate (which tends to be a slimy or sludge-like), it may be difficult to collect the precipitate and separate it from the solution. In this regard, the addition of ammonium hydroxide solution and hydrochloric acid enhances contaminant removal and aids in the separation and collection of precipitate from the cleansed solution. The ammonium hydroxide solution promotes precipitation because the solubility of many metal hydroxides is relatively low. The silicate, which has a high surface area, then acts as a scavenger for the precipitated metal hydroxide contaminants. Preferably the stream is treated with ammonium hydroxide solution or ammonia gas comprising from about 1% to about 30% ammonium hydroxide by weight, even more preferably from about 10% to about 30% by weight, and most preferably from about 20% to about 30%. If ammonia gas is used, it can be sprayed directly into the solution. In order to ensure precipitation of substantially all of the contaminant, the ammonium hydroxide solution should be added in an amount of from about 0.001 to about 100 g/L of stream to be treated, preferably from about 0.01 to about 10 g/L, and most preferably from about 0.1 to about 1 g/L. The silicate and ammonium hydroxide precipitants may be added sequentially and in any order, or they may be added concurrently. However, the pH of the process stream should not be lowered with addition of acid until after the precipitants have been added. In the next step the stream is treated with any suitable acid known to those skilled in the art. Upon addition of acid to the stream, the contaminant-containing silica matrix will begin to gel, polymerize and/or precipitate. Generally, upon the addition of acid, clear liquid will begin to cloud and/or thicken, and solid particles will begin to form. Eventually the particles may get large enough in size to settle out of solution. For the reasons indicated above, it is important to obtain and maintain the pH of process stream in this step. Thus, the pH of the stream should be continuously monitored as the acid is slowly added. Preferably, acid should be added in a drop-wise fashion in an amount sufficient to lower the pH to between about 6 and about 8.5, more preferably to between about 7 to about 8, and most preferably to between about 7 to about 7.5. Mineral acids are most suitable for use in this step. Mineral acids selected from the group consisting of hydrochloric acid, nitric acid, sulfuric acid, phosphoric, or a mixture thereof, are preferred. Hydrochloric acid is even more preferred. In addition, acetic acid, formic acid, or other suitable organic acids known to those of skill in the art also may be used alone or in combination with the above mentioned mineral acids. In the next step of the process, the mixture is given sufficient time for the desired gelling, polymerization, and/or precipitation reactions to occur. Allowing for sufficient aging time is essential in order to remove the desired amounts of contamination from the solution and to produce a waste-containing "solid" which readily filters at an acceptable filtration rate. "Solid" as used herein and for purposes of the present invention includes any filterable precipitate, gel or polymer. Preferably, the stream is aged for about 5 minutes to about 15 hours, more preferably for about 30 minutes to about 10 hours, and most preferably for about 1 to 5 hours. If adequate time is not permitted for aging, the filtration rate is likely to be extremely slow and the filtrate will contain excessive and undesirable levels of contaminant. The gelled, polymerized and/or precipitated solids are easily handled and separated from the clean solution using any suitable technique known to those of ordinary skill in the art, including flocculation, settling and/or filtration techniques. Filtration techniques can be used to separate the waste-containing solid from the filtrate without substantial plugging or clogging of the filtration device, and at relatively rapid rates, i.e., within minutes. Suitable filtration techniques include but are not limited to vacuum filtration, filter press, or filter membranes. Preferably, each of the above described steps are undertaken sequentially and in the order set forth above; namely, the precipitating agents are added first, followed by addition of mineral acid, followed by the aging step, followed by separation of the clean stream from the contaminant-containing precipitate. In one preferred embodiment, addition of sodium silicate is followed by treatment with ammonium hydroxide solution, which is followed by treatment with hydrochloric acid, which is followed by aging and separation. In another embodiment, the ammonium hydroxide solution is added prior to treatment with sodium silicate. In yet another embodiment, the precipitants are added simultaneously. Although the above described method may be used in-situ and/or as a continuous process, it is intended to be used off-site and above ground in any suitable batch process wherein the entire process is carried out in one mixing tank. Although the method of the invention is ideal for treating uranium/thorium contaminated effluents, it also is suitable for treating extracting solutions used in various soil washing processes, such as those described in U.S. Pat. No. 5,128,068, which issued on Jul. 7, 1992, from U.S. patent application Ser. No. 529,092, filed May 25, 1990; U.S. patent application Ser. No. 648,673, filed Jan. 31, 1991, U.S. Pat. No. 5,268,128, which issued on Dec. 7, 1993, from U.S. Pat. No. 5,045,240, issued on Sep. 3, 1991, from U.S. patent application Ser. No. 345,852, filed May 1, 1989; and U.S. patent application Ser. No. 722,458, filed Jun. 27, 1991, in the name of Grant, et al., the disclosures of which are incorporated herein in their entirety. With the method of the present invention, it is possible to lower the amount of inorganic contamination to environmentally acceptable levels as set forth in the Federal Primary Drinking Water Standard (40 C.F.R., Part 141). The ability to accomplish solution decontamination using the methods of the invention, and in particular the novel combination of process steps, is demonstrated in the following example. EXAMPLE 1 A sample of water (approximately 400 g) contaminated with about 20 milligrams of copper was successfully treated according to the method of the invention as follows. Approximately 1 g of ammonium hydroxide (10 weight percent) was added to the water sample with stirring. Then approximately 13.6 g of sodium silicate (6 weight percent) was added to the solution with stirring. Next, concentrated hydrochloric acid (about 38 weight percent) was added dropwise to the solution while monitoring the pH. The pH of the stream dropped to between about 7.1 and 7.2 after the addition of about 3.5 g of acid. No additional acid was added. The solution was then mixed for an additional 5 minutes and then allowed to remain undisturbed for about 3 hours. The precipitated solids were easily separated from the solution by vacuum filtration. The resultant solution contained less than 0.2 milligrams of copper per liter of solution. From the above, it can be seen that the invention provides a simple, yet highly effective method for treating solutions contaminated with inorganic and radioactive species. The process utilizes a novel combination of steps which maximize contaminant removal, minimize waste volume, and produce a manageable waste stream. In addition, the method of the invention results in a precipitate which is readily treated and separated from the cleansed solution. The invention having now been fully described, it should be understood that it may be embodied in other specific forms or variations without departing from its spirit or essential characteristics. Accordingly, the embodiments described above are to be considered in all respects as illustrative and not restrictive, the scope of the invention being indicated by the appended claims rather than by the foregoing description, and all changes which come within the meaning and range of equivalency of the claims are intended to be embraced therein. |
043022952 | description | FIG. 1 shows a nuclear fuel element according to one embodiment of this invention. It is seen that a plurality of pellets 2, prepared by molding uranium dioxide, followed by sintering, are loaded in a cladding tube 1. Loaded in plenum 3 of the cladding tube 1 are a metal foil 5 having a tag gas implanted thereinto by ion implantation method and a spring 4. Both open ends of the cladding tube 1 are sealed by stoppers 6 and 7. In general, the cladding tube 1 and the stoppers 6, 7 are formed of a zirconium alloy for a BWR and of stainless steel for a fast breeder reactor. The spring 4 serves to maintain the pellets 2 and the metal foil 5 at their proper locations. In the embodiment of FIG. 1, the metal foil 5 is disposed on the laminated pellets 2. But, the metal foil may be disposed beneath or between the pellets. The tag gas-implanted metal foil can be prepared by employing a known ion implantation apparatus and method disclosed in, for example, U.S. Pat. No. 4,051,063 granted to R. S. Nelson et al. and U.S. Pat. No. 4,124,802 granted to M. Terasawa et al. FIG. 2 illustrates how Kr gas is implanted into an aluminum foil by using an apparatus disclosed in the latter U.S. patent. As shown in the drawing, Kr gas is sent from a gas reservoir 10 through a pressure control value into an ion source 11 so as to be ionized into Kr.sup.+. The krypton ion (Kr.sup.+) is accelerated by an accelerator 12 to have an energy of about 50 KeV and runs through a magnetic deflection device 13 so as to be implanted into a metal aluminum foil 15 disposed within an implantation chamber 14. It is seen that an evacuation means 16 is provided for evacuating the ion source 11, the accelerator 12 and the implantation chamber 14. When implanted into an aluminum foil, Kr.sup.+ with 50 KeV of acceleration energy penetrates about 300 A into the foil and is distributed as shown in FIG. 3. The maximum amount of tag gas which can be implanted into a metal foil, i.e., saturation amount, is determined by the acceleration energy imparted to the tag gas. Where a tag gas imparted with 50 KeV of acceleration energy is implanted into an aluminum foil, the saturation amount of the tag gas is about 1.times.10.sup.7 Kr/cm.sup.2. The ionized tag gas implanted into a metal foil is retained within the metal foil in the form of individual atoms or aggregation of atoms forming bubbles and is thermally diffused and released from the foil to the outside when the foil has been heated. The release of tag gas is promoted in accordance with elevation of the heating temperature. But, all the tag gas implanted into the metal foil is not always released by the heating. For example, about 55% of Kr gas implanted into an aluminum foil is released at a temperature of 450.degree. C. to which the plenum of a fuel element is exposed at the initial stage of a nuclear reactor operation, with about 45% of Kr gas retained within the aluminum foil. Naturally, the residual amount within a foil should be taken into account in determining the amount of tag gas which is to be implanted into the foil. In other words, the foil should be enabled to release a tag gas in an amount large enough to be detected. Where Kr ions accelerated to have an energy level of 500 KeV were implanted into a stainless steel foil in an amount of 2.times.10.sup.15 Kr/cm.sup.2, the implanted Kr begins to be released from the foil at 820.degree. C. and substantially all the Kr atoms are released at 1,200.degree. C. In general, a tag gas is implanted into a metal foil in an amount about two times as much as that required for detection. The required area of a metal foil is determined by the saturation amount of tag gas determined by the acceleration energy imparted to the ionized tag gas. If a tag gas is implanted into both sides of a metal foil, it naturally suffices for one surface area of the foil to be half the required area. A metal foil having a thickness greater than the penetration range of the ionized tag gas can perform its function. FIG. 3 shows that it suffices for a metal foil to be about 1,000 A thick in view of its tag gas sealing function. However, it is practical in view of manufacturing process and mechanical strength to use a metal foil about 1 to 3 .mu.m thick. Also, the shape of a metal foil need not be restricted as far as the foil can be loaded in a cladding tube. For example, a plurality of circular foils about 5.5 mm in diameter can be loaded in the form of a laminate or apart from each other within a cladding tube having an inner diameter of 5.6 mm. It is also possible to wind a ribbon-shaped foil into a coil of a diameter smaller than the inner diameter of a cladding tube for loading of the coil within the cladding tube. A tag gas can be implanted into a metal foil formed into a desired shape, or a tag gas-implanted metal foil can be formed into a desired shape. Where a tag gas is implanted into a stainless steel foil, the foil should desirably be disposed between fuel pellets because somewhat high temperature is required for releasing the implanted tag gas. In general, isotopes of Kr and Xe are used in the form of a mixture or independently as a tag gas. It is possible to implant a mixed gas into a metal foil. It is also possible to implant the tag gas components individually and successively into a metal foil. According to a preferred embodiment of this invention, each component of the tag gas is implanted separately into a single metal foil having a particular shape and a plurality of gas-implanted foils are loaded in combination within a cladding tube. In this embodiment, it is possible to make arrangement such that the kind of the tag gas component can be distinguished by the shape of the metal foil. This renders it possible to load a desired amount of a desired tag gas into a cladding tube efficiently. In addition, the mixing ratio of the components can be readily adjusted by counting the number of metal foils. A tag gas of Kr was actually implanted by ion implantation method into a ribbon-shaped aluminum foil 3 .mu.m in thickness, 2 cm in width and 650 cm in length. The volume of the foil was: EQU 2 cm.times.650 cm.times.0.0003 cm=0.39 cm.sup.3. The Kr gas was implanted at an acceleration energy of 50 KeV into both surface area of the foil, followed by winding the foil into a coil 5.0 mm in diameter and 2 cm in height. The coil was small enough to be loaded into a cladding tube of a fuel element. Suppose Kr was implanted into the foil at the amount of 4.times.10.sup.16 Kr/cm.sup.2, which is somewhat lower than the saturation amount under the acceleration energy of 50 KeV, i.e., 1.times.10.sup.17 Kr/cm.sup.2. In this case, the total amount of Kr implanted into the entire foil is about 1.times.10.sup.20 Kr (the entire surface area of the foil is: EQU 2 cm.times.650 cm.times.2=2,600 cm.sup.2). As a matter of fact, about 5.5.times.10.sup.19 Kr atoms were released from the foil when heated at 450.degree. C. for 5 minutes. Incidentally, 2 cc of Kr gas at standard condition, which is sufficient for use as a tag gas, contains 5.4.times.10.sup.19 Kr atoms. In other words, the amount of Kr atoms actually released from the foil in the above-described experiment is sufficient for use as a tag gas. As described above in detail, a tag gas is sealed in a metal foil in this invention, resulting in that the sealed tag gas is not released unless the metal foil is heated. In other words, the sealed tag gas is not released by the impulse or vibration in the step of assembly or transportation of a fuel element, rendering it very easy to handle the fuel element. In addition, a conventional apparatus can be used for assembling a fuel element, a special apparatus need not be used. It should also be noted that metal foils of different shapes can be used for sealing different tag gas components separately, rendering it possible to distinguish the tag gas component by the shape of the metal foil. In this case, the tag gas components can be mixed at a desired ratio quite easily. To reiterate, a tag gas implanted into a metal foil is released from the foil in accordance with elevation of the ambient temperature caused by start-up of a nuclear reactor operation. As a result, a cladding tube is filled with the tag gas in an amount large enough to be detected, rendering it possible to carry out gas tagging. |
053655613 | summary | FIELD OF THE INVENTION AND RELATED ART This invention relates to exposure control in an exposure apparatus for use in the manufacture of semiconductor microcircuits. More particularly, the invention is concerned with exposure control suitably usable in an X-ray exposure apparatus, for example, for exposing a mask and a wafer with a radiation (X-rays) from a synchrotron or otherwise to print a pattern of the mask upon the wafer. In prior art X-ray exposure apparatuses, the amount of exposure is controlled in the following manner: (1) From the sensitivity, to X-rays used, of a radiation-sensitive material (called a "resist") applied to a wafer and from the illuminance in the field of irradiation of the X-rays from a source (the illuminance being determined from experience), the exposure time for the whole irradiation field on the wafer is determined. PA1 (2) An alternative is such that, as disclosed in Japanese Laid-Open Patent Application, Laid-Open No. Sho 60-198726, an X-ray detector is provided in the X-ray irradiation field and, while taking into account the variation in the X-ray illuminance with time, the exposure time for the whole irradiation field is determined. SUMMARY OF THE INVENTION In these prior art examples, the exposure time is determined uniformly over the entire irradiation field or an exposure region on a wafer. This involves a problem that, when there is non-uniformness in illuminance within the exposure region on the wafer, it is difficult to assure correct exposure throughout the exposure region. This causes inconveniences. For example, in X-ray lithography wherein a pattern of a linewidth on an order of 0.25 micron may be printed, it is possible that a semiconductor device formed on a wafer does not operate with a desired performance. The non-uniformness in illuminance may be reduced by placing a radiation source far remote from a mask and a wafer. With this method, however, the illuminance itself within the whole exposure region on the wafer is reduced. It is accordingly a primary object of the present invention to provide a system for controlling exposure in an exposure apparatus such as an X-ray exposure apparatus, by which the amount of radiation absorbed by a radiation-sensitive material in an exposure region can be made substantially uniform throughout every portion of the exposure region even if non-uniformness in illuminance is present within the exposure region. In accordance with one aspect of the present invention, to achieve this object, a shutter means having a leading edge and a trailing edge is provided and the moving speed of each of the leading edge and the trailing edge when they pass the exposure region for the controlled passage and interception of the radiation, is controlled to ensure that at different portions of the exposure region different exposure times are set as desired. These and other objects, features and advantages of the present invention will become more apparent upon a consideration of the following description of the preferred embodiments of the present invention taken in conjunction with the accompanying drawings. |
046876260 | claims | 1. In a pressurized water nuclear power reactor having at least one steam generator with a main feedwater line, a passive emergency steam dump comprising: a first storage means for coolant water, the storage means having a drainage means to prevent excessive build-up of coolant water therein; a steam jet ejector disposed within the first storage means and immersed in the coolant water the steam jet ejector having an inlet and an outlet; means for feeding pressurized steam from the steam generator of a nuclear power reactor into the steam jet ejector; heat exchange means for cooling the hot condensate from the steam jet ejector, the heat exchange means having an inlet and an outlet, the inlet of the heat exchanger means being connected to the outlet of the steam jet ejector, and the outlet of the heat exchange means being connected to the first storage means; and a second storage means for coolant water in which the heat exchange means is immersed. 2. The passive emergency steam dump of claim 1 wherein the outlet of the steam jet ejector is also connected to the main feedwater line for the steam generator for returning condensate from the steam jet ejector to the steam generator when the pressure in the outlet exceeds a specified limit. 3. The passive emergency dump of claim 1 wherein the second storage means comprises an inverted L-shaped pool, a baffle being disposed within the pool to divide the pool into two compartments which are in fluid communication, the heat exchange means being stored in one of the vertical compartments in the pool. |
claims | 1. The method of protectively coating metallic uranium which comprises dipping the metallic uranium in a molten alloy comprising about 20-75% of copper and about 80-25% of tin, dipping the coated uranium promptly into molten tin, withdrawing it from the molten tin and removing excess molten metal, thereupon dipping it into a molten metal bath comprising aluminum until it is coated with this metal, then promptly withdrawing it from the bath. 2. The process of claim 1 wherein the aluminum bath comprises about 80-95% aluminum and about 20-5% silicon. claim 1 3. The method of protectively coating metallic-uranium which comprises dipping the metallic uranium in a molten alloy comprising about 20-75% of copper and about 80-25% of tin, dipping the coated uranium promptly into molten tin, withdrawing it from the molten tin and removing excess molten metal, thereupon dipping it into a molten alloy, comprising about 80-95% aluminum and about 20-5% silicon at a temperature of about 600xc2x0 C., promptly withdrawing the coated uranium from the aluminum silicon alloy and brazing it to solid aluminum by aluminum-silicon alloy of the aforesaid composition by bringing it while still at about 600xc2x0 C. into assembled relation with the molten alloy and solid aluminum and cooling the assembly to solidify the alloy. 4. The process of claim 3 wherein the metallic uranium is maintained at a temperature of at least about 600xc2x0 C. throughout the process from the copper-tin alloy dip to the assembly. claim 3 5. The method of protectively coating a metallic uranium rod, which comprises dipping the metallic uranium rod in a molten alloy comprising about 20-75% of copper and about 80-25% of tin, dipping the coated uranium rod promptly into molten tin, withdrawing it from the molten tin and removing excess molten metal, thereupon dipping it into a molten metal bath of aluminum-silicon alloy comprising about 80-95% aluminum and about 20-5% silicon until it is coated with aluminum-silicon, then promptly withdrawing it from the aluminum-silicon bath, inserting the hot coated rod into a loosely fitting aluminum can at about 600xc2x0 C. containing sufficient molten aluminum-silicon alloy of the aforesaid composition to fill the space between the rod and the can, and cooling the resulting assembly to solidify the alloy. 6. The process of claim 5 wherein the assembly of rod and can is immediately chilled to solidify the aluminum-silicon alloy. claim 5 7. The process of claim 5 wherein the metallic uranium is maintained at a temperature of at least about 600xc2x0 C. throughout the process from the copper-tin alloy dip to insertion in the aluminum can. claim 5 8. The process of claim 5 wherein the metallic uranium is maintained at a temperature of at least about 600xc2x0 C. throughout the process from the copper-tin dip to insertion in the aluminum can and the can is capped with an aluminum cap so that the clearance between cap and can is filled with the molten aluminum silicon alloy, and the assembly is thereupon chilled at once to cause the aluminum-silicon alloy to harden rapidly, forming a fine-grained alloy bond between the uranium rod and the aluminum can and cap. claim 5 9. The method of protectively coating a metallic uranium rod, which comprises dipping the metallic uranium rod in a molten alloy, comprising about 20-75% of copper and about 80-25% of tin, dipping the coated uranium promptly into molten tin, withdrawing it from the molten tin and removing excess molten metal, thereupon dipping it into a molten metal bath of sodium-modified, degassed aluminum-silicon alloy comprising about 80-95% aluminum and about 20-5% silicon until it is coated with sodium modified aluminum-silicon, then promptly withdrawing it from the sodium-modified aluminum-silicon bath, inserting the hot coated rod into a loosely fitting aluminum can at about 600xc2x0 C. containing sufficient sodium-modified, degassed aluminum-silicon alloy of the aforesaid composition to fill the space between the rod and the can, and cooling the resulting assembly to solidify the alloy. 10. The process of claim 9 wherein the metallic uranium rod is maintained at a temperature of at least about 600xc2x0 C. throughout the process from the copper-tin alloy dip to insertion in the aluminum can. claim 9 |
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claims | 1. A transport/storage cask for a radioactive material, comprising:an inner shell;an outer shell;a circular gamma ray shielding layer placed between said inner shell and said outer shell, said gamma ray shielding layer being formed by aligning a plurality of gamma ray shielding blocks composed of lead or a lead alloy in a block shape in the circumferential direction;a circular neutron shielding layer placed between said inner shell and said outer shell;a first metal member having a higher elasticity limit than the gamma ray shielding blocks and covering at least a part of each of the gamma ray shielding blocks; anda heat transmission fin formed of a thermally conductive material and extending to said outer shell, wherein said heat transmission fin does not contact said plurality of gamma ray shielding blocks. 2. The transport/storage cask for the radioactive material according to claim 1, whereinthe first metal member has a higher thermal conductivity than the gamma ray shielding blocks. 3. The transport/storage cask for the radioactive material according to claim 2, whereinthe first metal member is aluminum, an aluminum alloy, copper or a copper alloy. 4. The transport/storage cask for the radioactive material according to claim 1, whereina plurality of protruding portions for protruding into each of the gamma ray shielding blocks are formed on a cover surface serving as a surface of the first metal member opposing to each of the gamma ray shielding blocks. 5. The transport/storage cask for the radioactive material according to claim 1, whereina plurality of openings are formed in the first metal member, anda plurality of protrusions are formed in each of the gamma ray shielding blocks, at least a part of the protrusions being placed within the openings. 6. The transport/storage cask for the radioactive material according to claim 1, whereinthe first metal member has a section in a U shape. 7. The transport/storage cask for the radioactive material according to claim 6, whereinthe first metal member is arranged so that an opening part of the U shape is adjacent said inner shell. 8. The transport/storage cask for the radioactive material according to claim 1, whereineach of the gamma ray shielding blocks has an overlapping portion overlapping with other circumferentially neighboring gamma ray shielding block in the radial direction. 9. The transport/storage cask for the radioactive material according to claim 1, whereinsaid neutron shielding layer is composed of an organic material including hydrogen, and the organic material is a resin material or a rubber material. 10. The transport/storage cask for the radioactive material according to claim 1, whereinsaid neutron shielding layer is formed by aligning a plurality of neutron shielding blocks in a block shape. 11. The transport/storage cask for the radioactive material according to claim 10, whereinthe neutron shielding blocks are formed in a circular shape and arranged on an outer periphery of a plurality of the gamma ray shielding blocks. 12. The transport/storage cask for the radioactive material according to claim 10, whereinat least a part of each of the neutron shielding blocks is covered with a second metal member having a higher elasticity limit than the neutron shielding blocks. 13. The transport/storage cask for the radioactive material according to claim 12, whereinthe second metal member has a higher thermal conductivity than the neutron shielding blocks. 14. The transport/storage cask for the radioactive material according to claim 13, whereinthe second metal member is aluminum, an aluminum alloy, copper or a copper alloy. 15. The transport/storage cask for the radioactive material according to claim 12, whereinthe second metal member has a section in a U shape. 16. The transport/storage cask for the radioactive material according to claim 1, whereina gel material is coated over at least one of among a contact surface between said inner shell and said gamma ray shielding layer or the neutron shielding layer, a contact surface between said gamma ray shielding layer and said neutron shielding layer, and a contact surface between said outer shell and said gamma ray shielding layer or said neutron shielding layer. 17. The transport/storage cask for the radioactive material according to claim 16, whereinthe gel material is silicon or a silicon material. 18. The transport/storage cask for the radioactive material according to claim 1, whereina reinforcing material having a higher elasticity limit than the gamma ray shielding blocks is buried within each of the gamma ray shielding blocks. 19. The transport/storage cask for the radioactive material according to claim 1, wherein the heat transmission fin is L-shaped and includesa part extending from the inner shell to the outer shell; anda circumferentially extending part abutting the outer shell. 20. The transport/storage cask for the radioactive material according to claim 1, wherein the heat transmission fin contacts the first metal member. |
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046997601 | description | DETAILED DESCRIPTION OF THE INVENTION In the following description, like reference characters designate like or corresponding parts throughout the several views. Also in the following description, it is to be understood that such terms as "forward", "rearward", "left", "right", "upwardly", "downwardly", and the like, are words of convenience and are not to be construed as limiting terms. IN GENERAL Referring now to the drawings, and particularly to FIG. 1, there is shown an elevational view of a reconstitutable nuclear reactor fuel assembly, represented in vertically foreshortened form and being generally designated by the numeral 10. Basically, the fuel assembly 10 includes a lower end structure or bottom nozzle 12 for supporting the assembly on the lower core plate (not shown) in the core region of a reactor (not shown), and a number of longitudinally extending guide tubes or thimbles 14 which project upwardly from the bottom nozzle 12. The assembly 10 further includes a plurality of transverse grids 16 axially spaced along the guide thimbles 14 and an organized array of elongated fuel rods 18 transversely spaced and supported by the grids 16. Also, the assembly 10 has an instrumentation tube 20 located in the center thereof and an upper end structure or top nozzle 22 removably attached to the upper ends of the guide thimbles 14, in a manner fully described below, to form an integral assembly capable of being conventionally handled without damaging the assembly parts. As mentioned above, the fuel rods 18 in the array thereof in the assembly 10 are held in spaced relationship with one another by the grids 16 spaced along the fuel assembly length. Each fuel rod 18 includes nuclear fuel pellets 24 and the opposite ends of the rod are closed by upper and lower end plugs 26,28 to hermetically seal the rod. Commonly, a plenum spring 30 is disposed between the upper end plug 26 and the pellets 24 to maintain the pellets in a tight, stacked relationship within the rod 18. The fuel pellets 24 composed of fissile material are responsible for creating the reactive power of the nuclear reactor. A liquid moderator/coolant such as water, or water containing boron, is pumped upwardly through the fuel assemblies of the core in order to extract heat generated therein for the production of useful work. To control the fission process, a number of control rods 32 are reciprocally movable in the guide thimbles 14 located at predetermined positions in the fuel assembly 10. Specifically, the top nozzle 22 includes a rod cluster control mechanism 34 having an internally threaded cylindrical member 36 with a plurality of radially extending flukes or arms 38. Each arm 38 is interconnected to a control rod 32 such that the control mechanism 34 is operable to move the control rods 32 vertically in the guide thimbles 14 to thereby control the fission process in the fuel assembly 10, all in a well-known manner. TOP NOZZLE ATTACHING STRUCTURE As illustrated in FIGS. 2 and 7, the top nozzle 22 has a lower adapter plate 40 with a plurality of control rod passageways 42 (only one being shown) formed through the adapter plate. The control rod guide thimbles 14 have their uppermost end portions 44 coaxially positioned within the passageways 42 in the adapter plate 40. For gaining access to the fuel rods 18 in reconstitution of the fuel assembly 10, the adapter plate 40 of the top nozzle 22 is removably connected to the upper end portions 44 of the guide thimbles 14 by an attaching structure, generally designated 46, which provides a plurality of structural joints between the top nozzle and the guide thimbles of the fuel assembly skeleton. The attaching structure 46 is generally the same as illustrated and described in the sixth application cross-referenced above, but will be described herein to the extent necessary to facilitate an understanding of the present invention. As best seen in FIGS. 2 through 7, the top nozzle attaching structure 46 of the reconstitutable fuel assembly 10 includes a plurality of outer sockets (only one being shown) defined in the top nozzle adapter plate 40 by the plurality of passageways 42 (also only one being shown) which each contains an annular circumferential groove 48 (only one being shown), a plurality of inner sockets (only one being shown) defined on the upper end portions 44 (only one being shown) of the guide thimbles 14, and a plurality of removable and reusable locking tubes 50 (only one being shown) inserted in the inner sockets to maintain them in locking engagement within the outer sockets. Each inner socket is defined by an annular circumferential bulge 52 on the hollow upper end portion 44 of one guide thimble 14 only a short distance below its upper edge 54. A plurality of elongated axial slots 56 are formed in the upper end portion 44 of each guide thimble 14 to permit inward elastic collapse of the slotted end portion to a compressed position so as to allow the circumferential bulge 52 thereon to be inserted within and removed from the annular groove 48 via the adapter plate passageway 42. The annular bulge 52 seats in the annular groove 48 when the guide thimble end portion 44 is inserted in the adapter plate passageway 42 and has assumed an expanded position. In such manner, the inner socket of each guide thimble 14 is inserted into and withdrawn from locking engagement with one of the outer socket of the adapter plate 40. More particularly, the axially extending passageway 42 in the adapter plate 40 which defines the outer socket is composed of an upper bore 58 and a lower bore 60. The lower bore 60 is of considerably greater axial length than the upper bore 58 and contains the annular groove 48 which is spaced a short distance below a ledge 62 formed at the intersection of the upper and lower bore 58,60. The lower bore 60 has a diameter which is greater than that of the upper bore 58; therefore, the ledge 62 faces in a downward direction. The primary purpose of the ledge 62 is to serve as a stop or an alignment guide for proper axial positioning of the upper end portion 44 in the passageway 42 when the inner socket is inserted into the outer socket. As seen in FIG. 7, the upper edge 54 of the guide thimble 14 abuts the ledge 62. Finally each locking tube 50 is inserted from above the top nozzle 22 into its respective locking position in the hollow upper end portion 44 of one guide thimble 14 forming one inner socket. When the locking tube 50 is inserted in its locking position, as seen in FIG. 7, it retains the bulge 52 of the inner socket in the latter's expanded locking engagement with the annular groove 48 and prevents the inner socket from being moved to its compressed releasing position in which it could be withdrawn from the outer socket. In such manner, each locking tube 50 maintains its respective one inner socket in locking engagement with the outer socket, and thereby retain the structural joint formed by the attachment of the adapter plate 40 of the top nozzle 22 on the upper end portion 44 of each guide thimble 14 in an assembled rigid form. Also, the locking tube 50 includes a slightly outwardly flared (for instance 1-2 degrees) upper peripheral marginal edge portion 64 (FIG. 6), which has an outer diameter slightly larger than the diameter of the upper bore 58, and a pair of small dimples 66 (being shown exaggerated in size) to secure it in its locking position. The flared edge portion 64 provides a tight frictional fit with the adapter plate 40, whereas the dimples 66 extend into the circumferential groove 48 defined in the adapter plate passageway 42. The locking tube 50 will yield as a whole in allowing withdrawal of the dimples 66 from the groove 48 and the tube from its locking position, after which the locking tube 50 will spring back to its original shape. APPARATUS FOR RECTIFYING DAMAGED GUIDE THIMBLE UPPER ENDS Turning now to FIG. 8, for effectuating inspection, removal, replacement and/or rearrangement of fuel rods 18 contained in the reconstitutable fuel assembly 10, the irradiated assembly must be removed from the reactor core and lowered into a work station 68 by means of a standard fuel assembly handling tool (not shown). In the work station 68, the fuel assembly 10 is submerged in coolant and thus maintenance operations are performed by manipulation of remotely-controlled submersible equipment. One component of such equipment is used for removing and replacing the locking tubes and another component is used, after the locking tubes have been removed, for removing and subsequently replacing the top nozzle 22 from and on the guide thimbles 14 of the reconstitutable fuel assembly 10. These components can take the form of the ones described and illustrated in the second through fifth application cross-referenced above. The work station 68 includes a pair of elongated bullet-nose guide members 70 which are mounted on, and project upwardly from, a pair of diagonal corners of a top flange 72 of the work station 52. The guide members 70 assist in alignment of the various components used to remove the locking tubes 50 and remove and replace the top nozzle 22. Also, included in the work station 68 are opposed pairs of movable pads 74 (only one pair being shown) that are mounted on and project through the side walls 76 of the work station at the elevation of the uppermost grid 16 of the fuel assembly 10. The pads 74 are advanced inwardly by cylinders 78, also mounted on the side walls 76, to bear against each side of the grid 16 and take up the clearance between the station 68 and the assembly 10, and thus maintain it in a fixed relation to the work station. In FIG. 8, there is also depicted three general types of damage which, although improbable, could possibly happen to the upper end portions 44 (or insert sleeves) of the guide thimbles 14, and particularly to the segments 80 thereof being separated by axial slots 56, should they be hit by a dropped object or bumped by a piece of equipment during reconstitution of the fuel assembly 10 when the top nozzle 22 is removed from the assembly and the guide thimbles 14 are exposed from above. The first type of segment damage, being represented on guide thimble upper end portion 44a, involves one (or more) of the segments 80a being flared or bent outwardly away from its desired vertical position. In such position of the segment 80a, the damaged upper end portion 44a of the guide thimble 14a will not fit into the corresponding passageway 42 in the top nozzle adapter plate 40 when the top nozzle is replaced on the guide thimbles. The second type of segment damage, being represented on guide thimble upper end portion 44b, involves one (or more) of the segments 80b being bent inwardly away from its desired vertical position. In such position of the segment 80b, the damaged upper end portion 44b of the guide thimble 14b may fit into the corresponding adapter plate passageway 42, but will not seat properly in the groove 48 therein nor will it receive one of the locking tubes 50. However, in view of the invention of the last cross-referenced application, each of the first and second types of segment damage are considered to be directly repairable, as will be explained shortly. Such is not the case with the third type of damage, being represented on the guide thimble upper end portion 44c. It involves one (or more) of the segments 80c being too severly bent outwardly or damaged to be directly repairable, such as by bending it back to its desired vertical position. Nonetheless, the cross-referenced invention makes provision for rectifying the damage by removing the affected section of the upper end portion 44c, as also will be explained shortly. Then, in accordance with the present invention, which will also be described below, a substitute or replacement structure is provided for the removed section. Turning now to FIGS. 9 through 16, as well as to FIG. 8, there is shown a fixture, constituting the preferred embodiment of the cross-referenced invention and being generally designated 82, which can be used during reconstitution of the fuel assembly 10, if the need arises, to rectify (i.e., repair or remove) damage to any of the axially segmented upper end portions 44 of the fuel assembly guide thimbles 14. With the top nozzle 22 removed from the fuel assembly 10, it is possible although highly improbable that, during fuel assembly reconstitution operations, at least one of the guide thimble segments 80 may incur one of the three different types of repairable and irreparable damage which were briefly described above. As shown particularly in FIGS. 9 and 10, the damage rectifying fixture 82 includes a base 84 made up of a generally square plate 86 enclosed by and connected with a frame 88. The plate 86 has a plurality of tool positioning openings 90 defined therein in a pattern matched with that of the guide thimbles 14. The frame 88 has means for mounting the base 84 on the work station 68 in the form of a pair of aligning holes 92 defined in two opposite corners of the frame for receiving the guide members 70 which project upward from the top flange 72 of the work station. When the base 84 is positioned above the station 68, as seen in FIG. 8, and then lowered to just receive the guide members 70 of the station upwardly through its aligning holes 92, as seen in FIG. 18, the tool positioning openings 90 of the base 84 are then disposed in alignment with the exposed upper end portions 44 of the fuel assembly guide thimbles 14. Further, the damage rectifying fixture 82 incorporates a set of three different tools 94,96,98 corresponding to the three different types of damage which can be incurred by the guide thimbles 14, as described above. Each of the tools 94,96,98 is selectively mountable to the base plate 86 at any one of its openings 90 by a threaded nut and bolt connection 100 provided on the upper end of the tool. Basically, therefore, each tool is adapted to operate to rectify a different one of the types of repairable and irreparable damage to the segments 80 on the upper end portions 44 of the guide thimbles 14. (It will be recalled that the portions 44, which commonly take the form of insert sleeves, form the structural joints with the top nozzle adapter plate 40, when the top nozzle 22 is remounted on the guide thimbles 14, as was described earlier). In addition to the damage rectifying tools 94,96,98, the fixture 82 incorporates a group of positioning elements 102 mounted in selected ones of the base openings 90 which will not be occupied by any of the tools. Each of positioning elements 102, as best seen in FIGS. 11 and 12, is a non-compresser socket having a cylindrical body 104 attached to the base plate 86 at one of the openings 90 by a threaded connection 106 and a cylindrical cavity 108 defined in the lower end of the body. The body 104 has outwardly tapered lower guide opening 109 defined therein which leads into the cavity 108. The cavity 108 is of a diameter size slightly larger than the outside diameter size of an undamaged one of the upper end portions 44 which adapts the positioning element 102 to mount upon the undamaged guide thimble upper end portion 44d, as seen in FIG. 19. Thus, selection of the specific positioning holes 90 in the fixture base 84 for attachment of the non-compresser socket elements 102 shall correspond to undamaged guide thimble portions 44d, preferably four in number (only one is shown in the drawings for purposes of clarity), with one in each quadrant of the fuel assembly skeleton. The function of the positioning elements 102 is to precisely locate the fixture 82 relative to the guide thimble upper end portions 44 for ensuring accurate performance of the damage rectifying operations carried out by the tools 94,96,98. Specifically, the purpose of the socket elements 102 is to locate the fixture 82 at a precise distance above and parallel to the normal plane of the top edges 54 of the guide thimbles. In FIGS. 13 and 14, as well as in FIG. 8, the damage rectifying tool 94 is the one designed to repair the first type of damage described above with respect to one (or more) of the segments 80a on the guide thimble upper end portion 44a. The tool 94 takes the form of a compressor socket having a cylindrical body 110 attached to the base plate 86 at one of the openings 90 by the threaded connection 100 and a cylindrical cavity 112 defined in the lower end of the body. The body 110 has an outwardly tapered lower edge 114 defining a guide opening 114 leading into the cavity 112. The cavity 112 is of a diameter size generally slightly less than the outside diameter size of an undamaged one of the upper end portions 44 which adapts the compresser socket tool 94 to operate to rectify repairable damage to the guide thimble upper end portion 44a having an outwardly bent segment 80a by compressing the segment 80a inward to its original position, as seen in FIG. 19. Thus, compresser socket tools 94 would be attached at positioning holes 90 of the base 84 corresponding to those of the guide thimbles having damage of the first type. In FIGS. 15 and 16, as well as in FIG. 8, the damage rectifying tool 96 is the one designed to repair the second type of damage described above with respect to one (or more) of the segments 80b on the guide thimble upper end portion 44b. The tool 96 takes the form of an expander pin having a cylindrical body 118 attached to the base plate 86 at one of the openings 90 by the threaded connection 100 and a reduced diameter lower section 120 with a tapered guide end 122 defined on the lower end of the body. The body 118 has a diameter size generally slightly less than the inside diameter of an undamaged one of the upper end portions 44 which adapts the expander pin tool 96 to operate to rectify repairable damage to the guide thimble upper end portion 44b having an inwardly bent segment 80b by expanding the segment 80b outward to its original position, as seen in FIG. 19. Thus, expander pin tools 96 would be attached at positioning holes 90 of the base 84 corresponding to those of the guide thimbles having damage of the second type. Finally, the tool 98, being the one to rectify the third type of damage, preferably takes the form of a precision internal cutter, such as either one of the two embodiments disclosed and illustrated in U.S. application Ser. No. 649,864 and assigned to the assignee of the present invention, the disclosure of which is incorporated by reference. The cutter tool 98 is operable to rectify irreparable damage to one (or more) of the bent segments 80c of the guide thimble upper end portion 44c by severing and capturing a predetermined section 124 of the guide thimble upper end portion 44c which includes the irreparably damaged segment 80c, as seen in FIGS. 19 and 20. In FIGS. 17 and 18, the damage rectifying fixture 82 is shown positioned in the work station 68 with the top ends of the guide members 70 extending through the aligning holes 92 of the fixture base 84. A long-handled lifting tool (not shown) threaded at its lower end 126 into a threaded central bore 128 in the base plate 86 is used to more the fixture 82 into the starting position shown in FIGS. 17 and 18, and also to more the fixture further toward the guide thimbles 14 to carry out the rectifying operations. Normally, only one type of damage condition would exist at one time, and even then infrequently, in a given fuel assembly and thus each damage rectifying operation using an appropriate one of the tools 94,96,98 would be carried out separately. However, for purposes of explaining the three types of damage and the three tools 94,96,98 which can be used by the fixture 82 of the present invention to rectify the damage, the three different damage conditions are shown as existing at the same time on the fuel assembly guide thimbles 14 in FIG. 18 and will thus be rectified generally at the same time by the three different tools. Therefore, in FIGS. 17 and 18, the fixture 82 is illustrated with one compresser socket tool 94, one expander pin tool 96, one cutter tool 98 and four non-compresser socket elements 102 aligned with corresponding damaged and undamaged guide thimble upper end portions 44a-c and 44d. Note that the four socket elements 102 are mounted in four different quadrants of the base plate 86 corresponding to the four different quadrants of the fuel assembly 10. Downward movement of the long-handled tool (not shown), being connected at its lower end 126 to the base 84, moves the fixture 82 from its starting position in FIG. 18 to its maximum lowered position in FIG. 19 in which the four socket positioning elements 102 are seated on four undamaged guide thimble upper end portions 44d. As the fixture 82 is lowered, the compresser socket tool 94 and expander pin tool 96 move toward the respective damaged upper end portions 44a,44b of the guide thimble 14a,14b into contact therewith. Engagement of the tapered lower edge 114 of the tool 94 and the tapered lower end 122 of the tool 96 with the respective bent segments 80a,80b of the guide thimbles 14a,14b initiates corrective bending of the segments back to their original positions concurrently as the fixture 82 and elements 102 tools 94,96 therewith are lowered. Each of the damaged guide thimble upper end portions 44a,44b are forceably returned and plastically set to an effective outside diameter that permits subsequent entry into the passageways 42 of the top nozzle adapter plate 40. In such manner, the tools 94,96 being adapted to rectify the two types of repairabale damage are operated to rectify the corresponding types of guide thimble segment damage. In addition, the above-described movement of the fixture 82 toward the guide thimbles 14 places the cutter tool 98 within the irreparably damaged upper end portion 44c of the guide thimble 14c with its cutter blades 130 at an operative position to sever the section 124 of the guide thimble which contains the irreparably damaged segment 80c. The cutter tool 98 is connected with and operated by the lower end 132 of a long-handled cutter operating tool (not shown). When the damaged section 124 has been severed, the radial advancement of the cutter blades 130 captures the cut-off section for removal when the fixture 82 is withdrawn from the work station 68, as shown in FIG. 20. TOP NOZZLE/GUIDE THIMBLE NON-STRUCTURAL JOINTS Whereas the first and second types of damage have now been corrected and the repaired upper end portions 44a,44b of the guide thimbles 14a,14b are ready to receive the top nozzle adapter plate 40 to complete structural joints therewith, the severed upper end portion 44c of the guide thimble 14c cannot complete such a joint with the top nozzle. Instead without some additional corrective measure, a open space or gap will be left between the severed upper edge 134 of the guide thimble 14c and the adapter plate 40, allowing disruptive coolant cross flow through the gap and across the open upper end of the severed guide thimble 14c. Turning now to FIGS. 21 through 24, the solution provided by the present invention is to mount an adapter sleeve 136 in each of the passageways in the top nozzle adapter plate 40 that is aligned with a severed guide thimble 14c. The adapter sleeve 136 is of a length longer than that of the damaged section 124 severed from the guide thimble 14c. An upper portion 138 of the adapter sleeve 136 has generally the same length and outside diameter size as that of the severed section 124, whereas a lower portion 140 is much shorter in length but has a larger outside diameter. The lower portion 140 of the adapter sleeve 136 is sized to snugly fit over the severed upper end portion 44c of the severed guide thimble 14c, as seen in FIG. 24, and provide a slip fit joint therewith in which there is no rigid connection between the sleeve 136 and guide thimble 14c when the top nozzle 22 is replaced back on the undamaged and repaired upper end portions 44d,44a of the guide thimbles 14. By way of example, the adapter sleeve 136 can be approximately 1.5 inches long with the same inside and outside diameters as the severed section 124 over its upper portion 138 which is approximately one inch in length. The 0.5 inch lower portion 140 is expanded to a diameter that allows a close-clearance fit with the outside diameter of the remaining upper end portion 44c (the portion of the insert sleeve which remains fastened to the guide thimble 14c). The above solution requires no changes to the manufactured top nozzle, but only the mechanical attachment of the one (or more) adapter sleeve 136 to the adapter plate 40. The sleeve is permanently connected to the adapter plate by a simple roll-formed circumferential bulge 142 formed in the sleeve 136 so as to extend into the adapter plate passageway groove 48. The nature of the modification permits using the original top nozzle 22 removed from the fuel assembly 10 or a new replacement nozzle, if desired. The adapter sleeve 136 does not provide a structural joint between the top nozzle 22 and respective one guide thimble 14c of the fuel assembly skeleton. However, because of conservative load margins for structural joints connecting the top nozzle to the skeleton, some of the structural joints (provided by the attaching structure 46 in FIG. 7) can be eliminated without violating design criteria. Since the adapter sleeve 136 forms part of a non-structural joint, generally designated 144 in FIG. 24, it carries no load. The load of the fuel assembly 10 is carried by the remaining structural joints (attaching structure) 46. These structural joints 46 each support a greater load than they did before the damage occured when the full complement of joints shared the load. However, depending upon the fuel configuration, at least four to six sections 124 can be deleted without exceeding design limits on the remaining structural joints 46. To state it another way, in a fuel assembly having 24 guide thimbles, there should be at least three times as many structural as non-structural joints between the top nozzle and the guide thimbles. The adapter sleeve 136 is made of the same material, stainless steel, as the section 124 of the upper end portion 44c (or insert sleeve). The adapter sleeve 136 requires no locking tube 50, although locking tubes are inserted into the sleeves to maintain the same inside dimensions as normal top nozzle joints. The adapter sleeve 136 can be quickly installed, remotely, under water into irradiated nozzle adapter plates 40 or they can be installed into new, unirradiated replacement nozzles by a "hands-on" procedure. After the number of adapter sleeves 136 required have been mounted in the corresponding adapter plate passageway 42 which are aligned with the severed guide thimble 14c, the top nozzle 22 can be mounted to the skeleton in the normal manner using standard reconstitution fixturing, tooling, and in accordance with the same procedures developed for standard removable top nozzle handling. The aforementioned modification in no way compromises subsequent reconstitutions which may be necessary for that fuel assembly. Therefore, the purpose of the adapter sleeve 136 is to bridge the gap between the top of the severed upper edge 134 of the guide thimble 14c and the adapter plate 40, and thereby provide a continuous protective enclosure for the core component installed in the fuel assembly, i.e., either a control rod, a thimble plug, or a BPRA rod. The adapter sleeve 136 precludes cross flow against the core component and maintains the normal upward flow path of the primary coolant both inside and outside of the guide thimble. In all probability, the fuel assembly skeleton will be provided with a large majority of structural joints 46 and a small minority of non-structural joints 144 between its top nozzle adapter plate 40 and its guide thimbles 14. It is thought that the present invention and many of its attendant advantages will be understood from the foregoing description and it will be apparent that various changes may be made in the form, construction and arrangement thereof without departing from the spirit and scope of the invention or sacrificing all of its material advantages, the form hereinbefore described being merely a preferred or exemplary embodiment thereof. |
061730270 | description | DESCRIPTION OF THE PREFERRED EMBODIMENTS Preferred embodiments of a primary containment vessel according to the present invention will be described hereunder with reference to the accompanying drawings. Further, it is to be noted that the terms or wordings representing directions, positions or the like such as "upper", "lower", "vertical", "horizontal" or the like are used herein in an installed state of a reactor container. First Embodiment (FIG. 1 to FIG. 7) FIG. 1 is a cross sectional view schematically showing an entire structure of a primary containment vessel according to a first embodiment of the present invention. The primary containment vessel of this first embodiment is constructed as follows. As shown in FIG. 1, a primary containment vessel 21 is provided with a reactor pressure vessel (RPV) 24, which is supported on a support skirt portion 23 by means of a RPV pedestal 22 having a substantially hollow cylindrical shape, at the center thereof. An outer peripheral side of the reactor pressure vessel 24 is surrounded with a hollow cylindrical outer peripheral concrete wall 25. Each lower end portion of the outer peripheral concrete wall 25 and the RPV pedestal 22 is supported on a mat concrete wall 26. The outer peripheral concrete wall 25 and the RPV pedestal 22 are joined together by means of a horizontal wall 27 at the substantially central position in a vertical direction of the RPV pedestal 22. In the reactor pressure vessel 24, an upper side from the support skirt portion 23 is surrounded by an upper dry well 28 which is a space defined by the outer peripheral concrete wall 25, the horizontal wall 27 and the RPV pedestal 22. On the other hand, in the reactor pressure vessel 24, a lower side from the support skirt portion 23 is surrounded by a lower dry well 29 which is a space defined by the mat concrete wall 26 and the RPV pedestal 22. The lower dry well 29 includes an air conditioner 30 used only for the lower dry well 29, a reactor water (coolant) recirculation pump (not shown) and a control rod drive mechanism (not shown). Further, when a plant is operated, an atmospheric air is filled in the lower dry well 29, and thereby, it is possible to use the lower dry well 29 as a space where workers can do work. An outer peripheral side of the lower dry well 29 is provided with a wet well 31 which is a space surrounded by the outer peripheral concrete wall 25, the mat concrete wall 26, the horizontal wall 27 and the RPV pedestal 22, at the outer peripheral side thereof. The wet well 31 is provided with a suppression pool 32, in which water is stored, at a half of the lower side thereof. The reactor pressure vessel supporting position is provided with a sealing material 33 which functions as isolating means for air-tightly isolating the upper dry well 28 and the lower dry well 29. Therefore, when the plant is operating, the upper dry well 28, the lower dry well 29 and the wet well 31 are air-tightly isolated. The RPV pedestal 22 is provided with a vertical vent pipe 34, which is opened to the upper dry well 28 side at an upper end portion thereof, as a vent pipe used only for the upper dry well 28. The vertical vent pipe 34 is connected to a plurality of horizontal vent pipes 35 at a lower end portion thereof and the horizontal vent pipes 35 are opened to the water stored in a suppression pool 32. The suppression pool 32 is stored with water capable of safely absorbing a thermal energy radiated from the reactor pressure vessel 24 when an assumed accident such as a main steam pipe breakdown accident happens. Further, the vertical vent pipe 34 used for only upper dry well 28 is provided with a communicating hole 36 which communicates a gas (vapor) phase section of the wet well 31 with the upper dry well 28. The communicating hole 36 is provided with a vacuum breaker 37 which functions as a high pressure gas inflow means. The vacuum breaker 37 selectively allows an inflow of an excessively high pressure gas of the gas phase section of the wet well 31 into the upper dry well 28 in an emergency. The RPV pedestal 22 is provided with a vertical vent pipe 38, which is opened to the lower dry well 29 side at an upper end portion thereof, as a vent pipe used only for the lower dry well 29. The vertical vent pipe 38 is connected to a plurality of horizontal vent pipes 39 at a lower end portion thereof, and the horizontal vent pipes 39 are opened to the water stored in the suppression pool 32. Furthermore, the RPV pedestal 22 is provided with a communicating hole 40 which communicates the gas phase section of the wet well 31 with the lower dry well 29. The communicating hole 40 is provided with a vacuum breaker 41 which functions as a high pressure gas inflow means. The vacuum breaker 41 selectively allows an inflow of an excessively high pressure gas of the gas phase section of the wet well 31 into the lower dry well 29 in an emergency. The primary containment vessel 21 is provided with a passageway which penetrates through the outer peripheral concrete wall 25, the wet well 31 and the RPV pedestal 22 and communicates the outside of the primary containment vessel 21 with the lower dry well 29. The passageway includes: an equipment carrying in and out passageway 42 for carrying in and out equipments included in the lower dry well 29; a personnel passageway for workers 43 for coming in the lower dry well 29 from the outside of the primary containment vessel 21 so that the workers do work in the lower dry well 29; and an equipment passageway 44 for locating a heat exchanger cooling pipe of a reactor water recirculation pump (not shown), an electric cable for a control rod drive mechanism, electric cables of other equipments included in the lower dry well 29, a cooling water pipe to the air conditioner 30 used only for the lower dry well 29, or the like. These equipment carrying in and out passageway 42 and personnel passageway 43 are provided with a scram pipe of the control rod drive mechanism (not shown) from the lower dry well 29. FIG. 2 is a cross sectional view taken along the line II--II of FIG. 1. With reference to FIG. 2, the primary containment vessel 21 is provided with the upper dry well 28 inside the outer peripheral concrete wall 25 having an ring shape in its cross section. An inner peripheral side of the outer peripheral concrete wall 25 is provided with the RPV pedestal 22 which is surrounded by the wet well 31 and has a ring shape in its cross section. The RPV pedestal 22 is provided with a plurality of vertical vent pipes 34 and 38 which have a circular shape in its cross section. The lower dry well 29 is formed inside the RPV pedestal 22. The primary containment vessel 21 is provided with a passageway which penetrates through the outer peripheral concrete wall 25, the wet well 31 and the RPV pedestal 22 and communicates the outside of the primary containment vessel 21 with the lower dry well 29. The passageway includes: an equipment carrying in and out passageway 42 for carrying in and out equipments included in the lower dry well 29; a personnel passageway 43 for workers for coming in the lower dry well 29 from the outside of the primary containment vessel 21 so that the workers do work in the lower dry well 29; and two equipment passageways 44 for locating a heat exchanger cooling pipe of a reactor water recirculation pump (not shown), an electric cable for a control rod drive mechanism, electric cables of other equipments included in the lower dry well 29, a cooling water pipe to the air conditioner 30 used for only lower dry well 29, or the like. These equipment carrying in and out passageway 42 and personnel passageway 43 are provided with a scram pipe of the control rod drive mechanism (not shown) from the lower dry well 29. In this first embodiment, the equipment carrying in and out passageway 42 and the personnel passageway 43 are coaxially located on a horizontal plane, and two equipment passageways are located in a direction perpendicular to the equipment carrying in and out passageway 42 and personnel passageway 43 on the horizontal plane. These equipment passageways 44 may be located at a predetermined angle with respect to the equipment carrying in and out passage 42 and the personnel passageway 43 on the horizontal plane. In other words, the equipments passageways 44 are located so as not to intersect the equipment carrying in and out passage 42 and the personnel passageway 43. In the primary containment vessel 21 constructed as described above in the first embodiment, a flow of gas when an accident happens in the upper dry well 28 and the lower dry well 29 will be described below with reference to FIG. 3. FIG. 3 is a view to explain a flow of gas on the assumption that an accident happens in the upper dry well 28 and the lower dry well 29 of the primary containment vessel according to the first embodiment in the following two cases. (1) Case 1: Accident Happens in Upper Dry Well In the case 1 where an accident such as a main steam pipe breakdown accident happens in the upper dry well 28, a steam pressure of the upper dry well 28 rises up, and then, when the steam pressure becomes a predetermined pressure or more, a steam is jetted into water stored in the suppression pool 32 via the vertical vent pipe 34 used only for the upper dry well 28 and the horizontal vent pipe 35. The jetted high pressure steam is condensed by the water of the suppression pool 32, so that an atmospheric pressure of the upper dry well 28 can be reduced. On the other hand, the steam condensed by the water of the suppression pool 32 increases an atmospheric pressure of the gas phase section of the wet well 31 by a thermal expansion. Then, when the atmospheric pressure of the gas phase section of the wet well 31 becomes a predetermined pressure or more, the vacuum breaker 41 located in the communicating hole 40 communicating with the lower dry well 29 is released into the lower dry well 29 side. In other words, in this case, the lower dry well 29 may be regarded as a gas phase section of the wet well 31. In the conventional primary containment vessel, a pressure analysis has been carried out in a manner that the primary containment vessel is divided into the sum of the upper dry well and the lower dry well, and the sum (V1+V2) of a free space volume V1 of the upper dry well excluding a volume of built-in pipes and equipments and a free space volume V2 of the lower dry well and a free space volume V3 of the wet well are used as one condition of the analysis. In the case of the conventional primary containment vessel of the ABWR in the range of 1350 MWe, an error or the like on the analysis is 15% with respect to a design pressure 3.16 kg/cm.sup.2 g, and therefore, this is a value satisfying the design pressure. In this case, a ratio V3/(V1+V2) of a free space volume V3 of the wet well to the sum (V1+V2) of a free space volume V1 of the upper dry well and a free space volume V2 of the lower dry well was about 0.81. In this first embodiment, if a ratio (V2+V3)/V1 of the sum (V2+V3) of a free space volume V2 of the lower dry well 29 and a free space volume V3 of the wet well 31 to a free space volume V1 of the upper dry well 28 becomes about 0.81 or more, a load acting on an outer wall of the wet well 31 is reduced. Thus, it is possible to reduce the volume of the upper dry well 28, the lower dry well 29 and the gas phase section of the wet well 31, so that the volume of the primary containment vessel 21 can be reduced. Then, the following is a description on the analytic result made when the accident such as the above case 1 happens. In this case, the analysis target accident is a feed water pipe breakdown accident in the upper dry well 28. The analysis time is 50 seconds directly after the accident happens. The reason why the feed water breakdown accident is recited as the analysis target is as follows. More specifically, when a feed water pipe breaks down in the upper dry well 28, a high pressure steam and water in the reactor pressure vessel 24 are discharged from the reactor pressure vessel 24 side, and further, a reactor cooling water is discharged from a turbine (not shown) side until a feed water pump (not shown) is stopped. For this reason, the quantity of water and steam discharged into the upper dry well 28 by the feed water pipe breakdown accident are greater as compared with other pipe breakdown accidents such as a main steam pipe breakdown accident or the like. Therefore, the feed water pipe breakdown accident is the most severe one of accidents happening in the upper dry well 28 in the light of a rise up of the internal pressure of the primary containment vessel 21. FIG. 4 shows a behavior of pressure in the upper dry well 28 when the feed water breakdown accident happens in the primary containment vessel 21 shown in FIG. 1 according to the first embodiment and in the conventional primary containment vessel 1 shown in FIG. 15. In FIG. 4, a solid line 101a shows the present invention and a broken line 101b shows the conventional case. Incidentally, a primary containment vessel of ABWR in the range of 1350 MWe is assumed as the conventional primary containment vessel 1 for the sake of convenience. FIG. 5 shows a behavior of pressure in the gas phase section of the wet well 31 when the feed water breakdown accident happens in two primary containment vessels 21 and 1 likewise FIG. 4. In FIG. 5, a solid line 102a shows the present invention and a broken line 102b shows the conventional case. Further, in FIG. 5, a behavior of pressure in the lower dry well 29 of the primary containment vessel 21 according the first embodiment is additionally shown by a dotted chain line 102c. FIG. 6 shows a behavior of partial pressure of a non-condensable gas in the upper dry well, the lower dry well and the wet well when the feed water breakdown accident happens in the primary containment vessel 21 according to the first embodiment. FIG. 7 shows a behavior of partial pressure of a non-condensable gas in the upper dry well, the lower dry well and the wet well when the feed water breakdown accident happens in the conventional primary containment vessel 1. In FIG. 6 and FIG. 7, solid lines 103a and 104a show a partial pressure of the upper dry well, broken lines 103b and 104b show a partial pressure of the lower dry well, and dotted chain lines 103c and 104c show a partial pressure of the wet well. As shown in FIG. 4, the upper dry well pressure of the conventional primary containment vessel 1 rises up to 270 kPa at 50 seconds after the breakdown accident happens. On the contrary, in the primary containment vessel 21 according to the first embodiment, a peak pressure is about 250 kPa. Therefore, it can be seen that the rise-up quantity from the initial pressure is reduced about 13%. As shown in FIG. 5, this is because the wet well pressure of the reactor container 21 according to the first embodiment has been restrained lower as compared with the wet well pressure of the conventional primary containment vessel 1. The following is an explanation about the reason why the wet well pressure of the primary containment vessel 21 according to the first embodiment has been restrained lower as compared with the wet well pressure of the conventional primary containment vessel 1. That is, when comparing the pressure of the lower dry well of both primary containment vessels 1 and 21, in the case of the present invention shown in FIG. 6 and FIG. 7, the non-condensable gas existing in the upper dry well flows into (is moved to) both the wet well 31 and lower dry well 29. On the other hand, in the case of the conventional case shown in FIG. 7, it can be seen that the non-condensable gas existing in the upper dry well and the lower dry well flows into the wet well. Accordingly, in the primary containment vessels 21 of the first embodiment, the lower dry well performs a function as a wet well space in the light of a dispersion of pressure, and thus, it can be confirmed that the wet well pressure has been restrained as compared with the conventional case. (2) Case 2: Accident Happens in Lower Dry Well Referring again to FIG. 3, another case will be described below. In the case where an accident such as a small-diameter breakdown accident happens in the lower dry well 29, a steam pressure of the lower dry well 29 rises up, and then, when the steam pressure becomes a predetermined pressure or more, a steam (vapor) is jetted into water stored in the suppression pool 32 via the vertical vent pipe 38 used only for the lower dry well 29 and the horizontal vent pipe 39. The jetted high pressure steam is condensed by the water of the suppression pool 32, so that an atmospheric pressure of the lower dry well 29 can be reduced. On the other hand, the steam condensed by the water of the suppression pool 32 increases an atmospheric pressure of the gas phase section of the wet well 31 by a thermal expansion. Then, when the atmospheric pressure of the gas phase section of the wet well 31 becomes a predetermined pressure or more, the vacuum breaker 37 located in the communicating hole 36 communicating with the vertical vent pipe 34 used only for the upper dry well 28 is released into the upper dry well 28 side. In other words, in this case, the upper dry well 28 may be regarded as a gas phase section of the wet well 31. In the conventional primary containment vessel 1, as described above, a ratio V3/(V1+V2) of a free space volume V3 of the wet well to the sum (V1+V2) of a free space volume V1 of the upper dry well and a free space volume V2 of the lower dry well was about 0.81. In this first embodiment, if a ratio (V1+V3)/V2 of the sum (V1+V3) of a free space volume V2 of the upper dry well and a free space volume V3 of the wet well to a free space volume V2 of the lower dry well becomes about 0.81 or more, a load acting on an outer wall of the wet well 31 is reduced. Thus, it is possible to reduce the volume of the upper dry well 28, the lower dry well 29 and the gas phase section of the wet well 31, so that the volume of the primary containment vessel 21 can be reduced. According to the first embodiment, it is possible to reduce the volume of the upper dry well 28, the lower dry well 29 and the wet well 31, so that the volume of the primary containment vessel 21 can be reduced. Therefore, in a nuclear power plant construction, it is possible to reduce a cost spent in material and work, and to shorten a construction period of the plant. Moreover, it is possible for the workers to come in the lower dry well 29 during a plant operation, and to carry out a inspection work of the equipments included in the lower dry well 29, so that the plant can be safely operated, and also, a reliability of the plant can be improved. The primary containment vessel 21 of this first embodiment is applicable to a nuclear power plant having the same electric output as the conventional case. Second Embodiment (FIG. 8) FIG. 8 is a cross sectional view schematically showing the whole construction of a primary containment vessel according to a second embodiment of the present invention. This second embodiment is different from the first embodiment in that an outer peripheral side of a RPV pedestal is provided with a vertical vent pipe used only for the upper dry well and a horizontal vent pipe. Other construction is the same as that of the first embodiment. Therefore, like reference numerals are used to designate the same components as the first embodiment and the details thereof are omitted. In a primary containment vessel 21a of this second embodiment, as shown in FIG. 8, an outer peripheral side of the RPV pedestal 22 is provided with a vertical vent pipe 34a, which is opened to the upper dry well side at the upper end portion thereof, as a vent pipe used only for the upper dry well 28. The vertical vent pipe 34a is connected to a plurality of horizontal vent pipes 35a at a lower end portion thereof, and the horizontal vent pipes 35a are opened to the water stored in a suppression pool 32. The vertical vent pipe 34a used only for the upper dry well 28 is provided with a communicating hole 36a which communicates a gas phase section of the wet well 31 with the upper dry well 28. The communicating hole 36a is provided with a vacuum breaker 37a which functions as a high pressure gas inflow means. The vacuum breaker 37a selectively allows an inflow of an excessively high pressure gas of the gas phase section of the wet well 31 into the upper dry well 28 in an emergency. Moreover, in this second embodiment, the vertical vent pipe 38 used only for the lower dry well 29 and the horizontal vent pipe 39 have been provided in the RPV pedestal 22. The outer peripheral side of the RPV pedestal 22 may be provided with a vent pipe used only for the lower dry well 29. According to this second embodiment, in addition to the same effects as those of the first embodiment, there is no need of providing the vertical vent pipe 34a used only for the upper dry well 28 and the horizontal vent pipe 35a in the RPV pedestal 22, so that the construction of the primary containment vessel can be simplified. Third Embodiment (FIG. 9) FIG. 9 is a cross sectional view schematically showing the whole construction of a primary containment vessel according to a third embodiment of the present invention. This third embodiment is different from the first embodiment in that an outer peripheral side of a RPV pedestal is provided with a vertical vent pipe used only for the upper dry well. Other construction is the same as that of the first embodiment. Therefore, like reference numerals are used to designate the same components as the first embodiment and the details thereof are omitted. In a primary containment vessel 21b of this third embodiment, as shown in FIG. 9, an outer peripheral side of the RPV pedestal is provided with a vertical vent pipe 34b, which is opened to the upper dry well side at the upper end portion thereof, as a vent pipe used only for the upper dry well 28. The vertical vent pipe 34b is opened to the water stored in a suppression pool 32 at a lower end portion thereof. The vertical vent pipe 34b used only for the upper dry well 28 is provided with a communicating hole 36b which communicates a gas phase section of the wet well 31 with the upper dry well 28. The communicating hole 36b is provided with a vacuum breaker 37b which functions as a high pressure gas inflow means. The vacuum breaker 37b selectively allows an inflow of an excessively high pressure gas of the gas phase section of the wet well 31 into the upper dry well 28 in an emergency. According to this third embodiment, in addition to the same effect as the second embodiment, no horizontal vent pipe used for only upper dry well 28 is required, so that the construction of the primary containment vessel can be simplified. Fourth Embodiment (FIG. 10) FIG. 10 is a cross sectional view schematically showing the whole construction of a primary containment vessel according to a fourth embodiment of the present invention. This fourth embodiment is different from the first embodiment in that a passive containment cooling system (hereinafter, referred simply to as PCCS) is additionally provided as a primary containment vessel cooling system. Other construction is the same as that of the first embodiment. Therefore, like reference numerals are used to designate the same components as the first embodiment and the details thereof are omitted. A primary containment vessel 21c of this fourth embodiment is provided with a heat exchanger 45 which is located outside the outer concrete wall 25 and condenses a steam, a pipe 46a which communicates the heat exchanger 45 with the lower dry well 29, and a pipe 46b which communicates the heat exchanger 45 with the wet well 31. The pipe 46a is opened to the lower dry well 29, and on the other hand, the pipe 46b is opened to the water stored in the suppression pool 32 of the wet well 31. The pipe 46a guides a steam generated in the lower dry well 29 to the heat exchanger 45 when an accident happens. The heat exchanger 45 makes a heat exchange with the water stored in a PCCS pool 47 located on the outside of the primary containment vessel 21c, thus the steam generated in the lower dry well 29 being condensed. The pipe 46b guides a condensation water condensed by the heat exchanger 45 into the water stored in the suppression pool 32 of the wet well 31. The PCCS is a passive containment cooling system which does not use dynamic equipments such as a pump or the like and uses only natural force and the PCCS is operable under the condition of a case that it is difficult to use dynamic equipments by the occurrence of a severe accident or the like. Therefore, according to this fourth embodiment, in addition to the same effects as those described in the first embodiment, it is possible to further improve safety of the primary containment vessel even in the case where a severe accident or the like happens. The following is a description on an analysis result of an behavior in the primary containment vessel 21c of the fourth embodiment when a severe accident happens such that the reactor pressure vessel breaks down, and a melted reactor core flows into the lower dry well, and further, a main steam pipe breaks down. In this analysis case, an extremely rare accident of severe accidents is recited as an example, and more specifically, a loss of coolant accident (LOCA) happens in the upper dry well 28, and then, a reactor core is melted and flows into the lower dry well 29. Further, the main steam pipe and the reactor pressure vessel break down, and for this reason, a passage connecting the upper and lower dry wells. As a result, this analysis case is considered as a case of most reducing a dry well division effect according to the present invention. In this case, the analysis time is ten hours after a containment spray is operated (after five hours elapsed just from the occurrence of an accident). Thus, the containment spray is carried out for 30 minutes just after a calculation is started. FIG. 11 shows a behavior of a pressure of the upper dry well 28 when a loss of coolant accident (LOCA) in the primary containment vessel 21c of the fourth embodiment shown in FIG. 10 and the conventional primary containment vessel 1 shown in FIG. 15. FIG. 12 shows a behavior of a pressure of the lower dry well 29 when a loss of coolant accident (LOCA) in the primary containment vessel 21c and the conventional primary containment vessel 1. In these FIG. 11 and FIG. 12, solid lines 105a and 106a show the present invention, and broken lines 105b and 106b show the conventional case. FIG. 13 shows a behavior of a non-condensable gas of the upper dry well, the lower dry well and the wet well of the primary containment vessel 21c according to the fourth embodiment. FIG. 14 shows a behavior of a non-condensable gas of the upper dry well, the lower dry well and the wet well of the conventional primary containment vessel 1. In these FIGS. 13 and 14, solid lines 107a and 108a show a partial pressure of the upper dry well, broken lines 107b and 108b show a partial pressure of the lower dry well, and dotted chain lines 107c and 108c show a partial pressure of the wet well. As seen from FIG. 11, the pressures of the upper dry wells of both the primary containment vessel 1 and 21c are respectively lowered to 360 kPa and 330 kPa by means of the containment spray which has been carried out for 30 minutes just after a calculation is started. However, the pressure of the conventional primary containment vessel 1 rises up to about 500 kPa after ten hours elapsed, and on the contrary, the pressure of the primary containment vessel 21c of this fourth embodiment is about 390 kPa. Therefore, in the primary containment vessel of this fourth embodiment, the pressure rise-up is restrained to about 57% as compared with the conventional case. This is because the wet well pressure of the primary containment vessel 21c of the fourth embodiment is restrained lower than the wet well pressure of the conventional primary containment vessel 1, as seen from FIG. 12. The following is an explanation about the reason why the wet well pressure of the primary containment vessel 21c of the fourth embodiment is restrained lower than the wet well pressure of the conventional primary containment vessel 1. As seen from FIGS. 13 and 14, the pressure of the upper dry well is lowered by means of the containment spray which has been operated for 30 minutes just after a calculation is started. For this reason, a non-condensable gas in the wet well passes through the vacuum breaker, and then, flows into the upper dry well. Thereafter, in the dry well of the fourth embodiment, the non-condensable gas, which has flown into the upper dry well, is left alone in the upper dry well. On the contrary, in the case of the conventional primary containment vessel, the non-condensable gas, which has flown into the upper dry well, is again returned to the wet well, and as a result, the wet well pressure rises up. For this reason, in the conventional primary containment vessel 1, the pressure rise-up after a predetermined time elapsed is great as compared with the primary containment vessel 21c of the fourth embodiment. As is evident from the above explanation, it can be confirmed that the internal pressure of the primary containment vessel 21c of the fourth embodiment is restrained lower than the conventional case when a sever accident happens. As described above in detail, according to the primary containment vessel of the present invention, the upper dry well and the lower dry well are effectively used, and it is therefore possible to meet a request to increase an electric output of a nuclear power plant and to make a primary containment vessel into a compact size by means of a safely and relatively simple structure or arrangement. Furthermore, since the primary containment vessel is made into a compact size, it becomes possible to reduce a cost for constructing the nuclear power plant and to improve economics on a practical operation of the nuclear power plant. It is to be noted that the present invention is not limited to the described embodiments and many other changes and modifications may be made without departing from the scopes of the appended claims. |
abstract | An integral pressurized water reactor that combines all of the components typically associated with a nuclear steam supply system, such as the steam generator, reactor coolant pumps, pressurizer and the reactor, into a single reactor pressure vessel. The reactor pressure vessel is itself enclosed in a containment pressure vessel that also houses a number of safety systems, such as the core make-up tanks, the primary side of residual heat removal heat exchangers, an automatic depressurization system and a recirculation system that enables continuous core cooling through natural circulation over an extended period of time. Actuation of the passive systems is done by single actuation of valves, powered from redundant batteries. |
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claims | 1. A material, comprising:a first metal halide that is operative to function as a scintillator; where the first metal halide excludes cesium iodide, strontium iodide and cesium bromide; anda surface layer comprising a second metal halide that is disposed on a surface of the first metal halide; where the second metal halide has a lower water solubility than the first metal halide, where another layer that comprises a third metal halide is disposed on a surface of the second metal halide; and wherein the third metal halide has a lower water solubility than the second metal halide. 2. The material of claim 1, wherein the first metal halide has a composition that is described by formula (1):M1X1a:Yb (1)where M1 is a metal that is lithium, sodium, potassium, rubidium, cesium, thallium, copper, silver, lead, bismuth, indium, tin, antimony, tantalum, tungsten, strontium, barium, boron, magnesium, calcium, cerium, yttrium, scandium, gadolinium, lanthanum, lutetium, praseodymium, terbium, ytterbium, samarium, europium, holmium, dysprosium, erbium, thulium, or neodymium, X1 is a halogen, where the halogen is chlorine, bromine, iodine, astanine, or a combination thereof and were Y is a codopant and comprises and comprises one or more of thallium, copper, silver, lead, bismuth, indium, tin, antimony, tantalum, tungsten, strontium, barium, boron, magnesium, gadolinium, calcium, potassium, cerium, yttrium, scandium, lanthanum, lutetium, praseodymium, terbium, ytterbium, samarium, europium, holmium, dysprosium, erbium, thulium or neodymium, where M1 and Y are different elements, where a has a value of 1 to 4 and where b has a value of 0 to about 1 and where the second metal halide has a composition that is described by formula (5)M1X2a:Yb (5)where M1, Y, a and b are detailed above in the Formula (1), where X2 in the Formula (5) is a halogen that is fluorine, chlorine, bromine, iodine, or a combination thereof and that has at least one halogen atom having a lower atomic weight than X1 in the Formula (1). 3. The material of claim 2, where b has a value of about 0.001 to about 0.5 and where “a” has a value of 1, 2 or 3. 4. The material of claim 2, where when X2 and X1 both involve a combination of halogen atoms, the combined atomic weight of X2 is less than X1 or alternatively, when X2 involves a combination of halogen atoms, but X1 contains only a single halogen atom, then X2 contains at least one halogen atom that has a lower atomic weight than the single halogen atom contained in X1. 5. The material of claim 1, where the first metal halide has a composition that is described by Formula (2), Formula (3) or Formula (4):M1lM2mX1a:Yb (2)M1lM2mM3nX1a:Yb (3) andM1lM2mM3nM4oX1a:Yb (4),where in the Formulas (2), (3) and (4), where applicable, M1 is lithium, sodium, potassium, calcium, rubidium, gadolinium, cesium, thallium, copper, silver, lead, bismuth, indium, tin, antimony, tantalum, tungsten, strontium, barium, boron, magnesium, calcium, cerium, yttrium, scandium, lanthanum, lutetium, praseodymium, terbium, ytterbium, samarium, europium, holmium, dysprosium, erbium, thulium, or neodymium, M2 is selected from the group consisting of boron, aluminum, gallium, indium, sodium, potassium, calcium, gadolinium, rubidium, cesium, thallium, cerium, yttrium, scandium, lanthanum, lutetium, praseodymium, terbium, ytterbium, samarium, europium, holmium, dysprosium, erbium, thulium, or neodymium, M3 and M4 are different from one another and are one of strontium, calcium, barium, gadolinum, yttrium, scandium, lanthanum, lutetium, praseodymium, terbium, ytterbium, samarium, europium, holmium, dysprosium, erbium, thulium, or neodymium, X1 is a halogen selected from chlorine, bromine, iodine, astinine, or a combination thereof; Y is a codopant and comprises and comprises one or more of thallium, copper, silver, lead, bismuth, indium, tin, antimony, tantalum, tungsten, strontium, barium, boron, magnesium, gadolinium, calcium, potassium, cerium, yttrium, scandium, lanthanum, lutetium, praseodymium, terbium, ytterbium, samarium, europium, holmium, dysprosium, erbium, thulium or neodymium, where l is 0 to 3, m is 0 to 3, n is 0 to 3, and o is 0 to 3, and where the sum of l+m in Formula (2) is not equal to 0, where the sum of l+m, m+n, l+n and l+m+n is not equal to 0 in Formula (3) and where the sum of l+m, m+n, l+n, l+o, m+o, n+o and l+m+n+o is not equal to 0 in the Formula (4); where in the Formulas (2), (3) and (4), “a” is 1 to 8, and “b” is 0 to about 1; where in the Formula (2), M1, M2 and Y are different from each other and only one of M1 or M2 is a rare earth metal; where in the Formula (3), M1, M2, M3 and Y are different from each other and at least one of M1, M2 and M3 is a rare earth metal and where embodiment in the Formula (3), M2, M3 and Y are different from each other and both may or may not be rare earth metals and where in the Formula (4), M1, M2, M3 and M4 and Y are each different from each other and at least one of M1, M2, M3 and M4 and Y are rare earth metals. 6. The material of claim 5, where the second metal halide has a composition that is described by formula (5)M1X2a:Yb (5)where M1 is a metal that is lithium, sodium, potassium, rubidium, cesium, thallium, copper, silver, lead, bismuth, indium, tin, antimony, tantalum, tungsten, strontium, barium, boron, magnesium, calcium, cerium, yttrium, scandium, gadolinium, lanthanum, lutetium, praseodymium, terbium, ytterbium, samarium, europium, holmium, dysprosium, erbium, thulium, or neodymium, X1 is a halogen, where the halogen is chlorine, bromine, iodine, astanine, or a combination thereof and were Y is a codopant and comprises and comprises one or more of thallium, copper, silver, lead, bismuth, indium, tin, antimony, tantalum, tungsten, strontium, barium, boron, magnesium, gadolinium, calcium, potassium, cerium, yttrium, scandium, lanthanum, lutetium, praseodymium, terbium, ytterbium, samarium, europium, holmium, dysprosium, erbium, thulium or neodymium, where M1 and Y are different elements, where a has a value of 1 to 4 and where b has a value 0 to about 1 and where X2 in the Formula (5) is a halogen that is fluorine, chlorine, bromine, iodine, or a combination thereof and that has at least one halogen atom having a lower atomic weight than X1 in the Formulas (2), (3) and (4). 7. The material of claim 5, where the second metal halide has a composition that is described by formulas (6) through (8)M1lM2mX2a:Yb (6)M1lM2mM3nX2a:Yb (7) andM1lM2mM3nM4oX2a:Yb (8),where M1, M2, M3, M4, X, Y, l, m, n, o, a and b from the Formulas (6), (7) and (8) are defined above in Formulas (2) (3) and (4) and where X2 from the Formulas (6), (7) and (8) has a lower atomic weight than X1 in Formulas (2), (3) and (4). 8. The material of claim 6, where when X2 and X1 both involve a combination of halogen atoms, the combined atomic weight of X2 is less than X1 or alternatively, when X2 involves a combination of halogen atoms, but X1 contains only a single halogen atom, then X2 contains at least one halogen atom that has a lower atomic weight than the single halogen atom contained in X1. 9. The material of claim 7, where in formulas (6), (7) and (8) b has a value of about 0.001 to about 0.5 and where “a” has a value of 2 to 7. 10. The material of claim 1, where the first metal halide comprises Nat Tl, KI:Tl, LaCl3:Ce, LaBr3:Ce, LuCl3:Ce, LuBr3:Ce, or a combination thereof. 11. The material of claim 1, where the second metal halide comprises NaF:Tl, CsF:Tl, BaF2, CaF2(Eu), LaF3:Ce, LaBr3:Ce, LuCl3:Ce, LuBr3:Ce, SrF2:Eu, LaFBr2, LaF2Br, or a combination thereof. 12. An article comprising the material of claim 1. 13. The article of claim 12, where the article is a positron emission tomography device, a computed tomography device, or single photon emission computed tomography device. 14. A method comprising:disposing on a surface of a first metal halide a layer comprising a second metal halide;where the first metal halide is a scintillator and excludes cesium iodide, strontium iodide and cesium bromide; and where the second metal halide has a lower water solubility than the first metal halide; anddisposing on a surface of the second metal halide a layer comprising a third metal halide; where the third metal halide has a lower water solubility than the second metal halide. 15. The method of claim 14, wherein the first metal halide has a composition that is described by formula (1):M1X4a:Yb (1)where M1 is a metal that is lithium, sodium, potassium, rubidium, cesium, thallium, copper, silver, lead, bismuth, indium, tin, antimony, tantalum, tungsten, strontium, barium, boron, magnesium, calcium, cerium, yttrium, scandium, gadolinium, lanthanum, lutetium, praseodymium, terbium, ytterbium, samarium, europium, holmium, dysprosium, erbium, thulium, or neodymium, X1 is a halogen, where the halogen is chlorine, bromine, iodine, astanine, or a combination thereof and were Y is a codopant and comprises and comprises one or more of thallium, copper, silver, lead, bismuth, indium, tin, antimony, tantalum, tungsten, strontium, barium, boron, magnesium, gadolinium, calcium, potassium, cerium, yttrium, scandium, lanthanum, lutetium, praseodymium, terbium, ytterbium, samarium, europium, holmium, dysprosium, erbium, thulium or neodymium, where M1 and Y are different elements, where a has a value of 1 to 4 and where b has a value of 0 to about 1 and where the second metal halide has a composition that is described by formula (5)M1X2a:Yb (5)where M1, Y, a and b are detailed above in the Formula (1), where X2 in the Formula (5) is a halogen that is fluorine, chlorine, bromine, iodine, or a combination thereof and that has at least one halogen atom having a lower atomic weight than X1 in the Formula (1). 16. The method of claim 15, where b has a value of about 0.001 to about 0.5 and where “a” has a value of 1, 2 or 3. 17. The method of claim 15, where when X2 and X1 both involve a combination of halogen atoms, the combined atomic weight of X2 is less than X1 or alternatively, when X2 involves a combination of halogen atoms, but X1 contains only a single halogen atom, then X2 contains at least one halogen atom that has a lower atomic weight than the single halogen atom contained in X1. 18. The method of claim 14, where disposing on the second metal halide is produced by chlorinating or fluorinating the surface of a first metal halide. |
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claims | 1. A method of irradiating a target in a subject, comprising the steps of:positioning a subject on a supporting device;positioning a delivery device adapted to deliver charged particles; anddelivering charged particles to a target in the subject, wherein the delivery device is in motion and rotates around the target while at least some of the charged particles are delivered, and one or more parameters of the charged particles are modulated when the delivery device is in motion, said parameters including the energy, the intensity, and the beam direction of the charged particles. 2. The method of claim 1 wherein two or more parameters including the energy, the intensity, the beam direction and the beam shape of the charged particles are concurrently modulated when the delivery device is in motion. 3. The method of claim 1 wherein in the delivering step a multi-leaf collimator is used to shape and modulate the energy of the charged particles concurrently. 4. The method of claim 1 wherein said delivery device is mounted to a gantry capable of rotating in 360 degrees or more. 5. The method of claim 1 wherein in the delivering step the supporting device is concurrently moved. 6. The method of claim 1 wherein the charged particles are protons. 7. A radiation method comprising:positioning a subject on a supporting device;positioning a delivery device adapted to deliver charged particles; anddelivering charged particles to a target in the subject, wherein the delivery device is in motion while at least some of the charged particles are delivered, and wherein all or substantially all of charged particles for a treatment fraction are delivered to the target during a single rotation of the delivery device in about 360 degrees or less. 8. A method of irradiating a target in a subject, comprising the steps of:positioning a subject on a supporting device;positioning a delivery device adapted to deliver charged particles; anddelivering charged particles to a target in the subject, wherein the energy of the charged particles is modulated such that the Bragg peaks of the charged particles are deposited approximately on the distal periphery of the target, and the delivery device is in motion while at least some of the charged particles are delivered. 9. The method of claim 8 wherein in the delivering step the delivery device is stationary during delivery of at least a portion of the charged particles. 10. The method of claim 8 wherein the charged particles are in the form of a pencil beam. 11. The method of claim 8 wherein the charged particles are protons. 12. The method of claim 8 wherein in the delivering step the energy and the intensity of the charged particles are concurrently modulated. 13. The method of claim 8 wherein in the delivering step all or substantially all charged particles for a treatment fraction are delivered to the target during a single rotation of the delivery device in about 360 degrees or less. 14. The method of claim 8 wherein in the delivering step the supporting device is concurrently moved. 15. The method of claim 8 further comprising the step of gating the delivery of charged particles in response to abnormal or normal movement of the subject. 16. A charged particle therapy system comprising:a particle accelerator;a particle delivery device; anda beam path for transporting charged particles generated by the particle accelerator to the delivery device;wherein said delivery device is configured to rotate around a target and be in motion while at least some of the charged particles are delivered in operation, and said delivery device comprises a multi-leaf collimator configured to shape and modulate the energy of the particles concurrently. 17. The system of claim 16 wherein the delivery device is coupled to a gantry rotatable in 360 degrees or more. 18. The system of claim 16 wherein said multi-leaf collimator is configured to shape and scatter the charged particles concurrently. 19. The system of claim 16 wherein said multi-leaf collimator is a 3-dimensional multi-leaf collimator. 20. A method of irradiating a target in a subject, comprising the steps of:positioning a subject on a supporting device;positioning a delivery device adapted to deliver charged particles; anddelivering charged particles for a treatment fraction to a target in the subject using the delivery device with more than one rotations, wherein in a first rotation, the Bragg peaks of substantially all charged particles are deposited approximately on the distal periphery of the target, and in a second rotation, the Bragg peaks of substantially all charged particles are deposited approximately in the interior of the target. 21. The method of claim 20 wherein the delivery device is in motion while at least some of the charged particles are delivered. 22. The method of claim 20 wherein one or more parameters of the charged particles are modulated when the delivery device is in motion, said parameters including the energy, the intensity, the beam direction and the beam shape of the charged particles. 23. The method of claim 20 wherein two or more parameters including the energy, the intensity, the beam direction and the beam shape of the charged particles are concurrently modulated when the delivery device is in motion. 24. The method of claim 20 wherein said first or second rotation is a complete rotation in about 360 degrees or a partial rotation less than 360 degrees. 25. The method of claim 20 wherein the charged particles are delivered to the target at selected angles of rotations. 26. The method of claim 20 further comprising the step of gating the delivery of charged particles in response to abnormal or normal movement of the subject. 27. A radiation method comprising:positioning a subject on a supporting device;positioning a delivery device adapted to deliver charged particles; anddelivering charged particles to a target in the subject, wherein the delivery device is in motion and rotates around the target while at least some of the charged particles are delivered; andgating the delivery of charged particles in response to abnormal or normal movement of the subject. 28. A charged particle therapy system comprising:a particle accelerator configured to generate charged particles;a delivery device, said delivery device is rotatable and configured to be in motion while at least some of the charged particles are delivered in operationa beam path for transporting charged particles from the particle accelerator to the delivery device; anda control system configured to gate the delivery of charged particles in response to abnormal or normal movement of the target. 29. A radiation method comprising:positioning a subject on a supporting device;positioning a delivery device adapted to deliver protons or heavy ions; anddelivering protons or heavy ions to a target in the subject, wherein the delivery device is in motion and rotates around the target while at least some protons or heavy ions are delivered. 30. A charged particle therapy system comprising:a particle accelerator configured to generate protons or heavy ions;a delivery device; anda beam path for transporting protons or heavy ions generated by the particle accelerator to the delivery device;wherein said delivery device is rotatable and configured to be in motion while at least some protons or heavy ions are delivered in operation. |
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abstract | A nuclear reactor system for generating electricity includes a nuclear reactor that uses water as a coolant and a moderator, and generates thermal energy through nuclear fission. The reactor includes a reactor vessel and a reactor core. The reactor core comprises a plurality of fuel assemblies and one or more core baffles. In an embodiment the reactor core is square-shaped with a hollow center portion, and is concentric with the reactor vessel. The reactor permits decay-heat removal from the fuel assemblies in the absence of the coolant. A power conversion system is arranged to indirectly receive thermal energy generated by the reactor core and to generate electricity. A refueling water storage tank is useable when the reactor core is refueled with nuclear fuel. A containment surrounds the reactor vessel and the refueling water storage tank. |
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claims | 1. A structure comprising:a first scintillating screen that converts an absorbed portion of incident radiation directed at the structure into light photons;a photosensor array;a second scintillating screen,a fiber optic plate between the photosensor array and the second scintillating screen, the photosensor array being between the first scintillating screen and the fiber optic plate, the second scintillating screen converts an absorbed portion of the incident radiation transmitted through the first scintillating screen, the photosensor array and the fiber optic plate, into light photons, where a surface of the first scintillating screen faces the photosensor array and a surface of the second scintillating screen faces the fiber optic plate,wherein the photosensor array is operable to capture at least a portion of the light photons from the first scintillating screen and the second scintillating screen and convert the captured light photons into electrical signals,wherein the fiber optic plate is a substrate for the photosensor array. 2. The structure of claim 1, wherein the photosensor array comprises a plurality of bidirectionally photosensitive storage elements for capturing the at least a portion of the light photons from the first scintillating screen and the second scintillating screen, switching elements where one switching element of the plurality of switching elements corresponds to one of the plurality of photosensitive storage elements, respectively, a transparent metal bias layer and a transparent 2D patterned metal layer, where the transparent 2D patterned metal layer faces the fiber optic plate. 3. The structure of claim 1, wherein the first scintillating screen comprises a scintillating structure having a first thickness, and the second scintillating screen comprises a scintillating structure having a second thickness, where the second thickness is greater than the first thickness. 4. The structure of claim 1, wherein the second scintillating screen further comprises a backing, the backing contacting a surface of the second scintillating screen opposite of a surface facing the fiber optic plate. 5. The structure of claim 1, wherein the first scintillating screen further comprises a backing, the backing of the first scintillating screen facing an incoming x-ray beam energy. 6. The structure of claim 1, wherein the first scintillating screen and the second scintillating screen are formed of a different type, the type being granular or columnar. 7. The structure of claim 1, wherein the fiber optic plate has a thickness between 1 mm and 3 mm, inclusive. 8. The structure of claim 2, wherein at least one of the transparent 2D patterned metal layer and the transparent metal bias layer comprises an optical filter to balance gain from the first scintillating screen and the second scintillating screen. 9. The structure of claim 8, wherein the optical filter comprising a layer of absorbing material. 10. The structure of claim 3, wherein a ratio of the first thickness to a combination of the first thickness and the second thickness is based on at least one of an incoming x-ray beam energy, a target spatial resolution performance or a target detective quantum efficiency. 11. An imaging system comprising:a processor configured to be in communication with a structure comprising:a first scintillating screen that converts an absorbed portion of incident radiation directed at the structure into light photons;a photosensor array; anda second scintillating screen;a fiber optic plate between the photosensor array and the second scintillating screen, the photosensor array being between the first scintillating screen and the fiber optic plate, the second scintillating screen converts an absorbed portion of the incident radiation transmitted through the first scintillating screen, the photosensor array and the fiber optic plate into light photons, where a surface of the first scintillating screen faces the photosensor array and a surface of the second scintillating screen faces the fiber optic plate,wherein the photosensor array is operable to capture at least a portion of the light photons from the first scintillating screen and the second scintillating screen and convert the captured light photons into electrical signals,wherein the fiber optic plate is the substrate for the photosensor arraythe processor is configured to:receive the electrical signals from the structure; andproduce the image having a plurality of pixels using the electrical signals. 12. The imaging system of claim 11, wherein the photosensor array comprises a plurality of bidirectionally photosensitive storage elements for capturing the at least a portion of the light photons from the first scintillating screen and the second scintillating screen, switching elements where one switching element of the plurality of switching elements corresponds to one of the plurality of photosensitive storage elements, respectively and a first metal layer and a second metal layer,wherein the first metal layer is a transparent metal bias layer and the second metal layer is a transparent 2D patterned metal layer, where the transparent 2D patterned metal layer faces the fiber optic plate, wherein the processor controls each row of switching elements using a scanning control unit, thereby connecting the corresponding photosensitive storage elements to amplifiers, whose outputs are digitized to pixel values for each row of the image. 13. A radiation detector comprising:a first radiation converter;a second radiation converter;a photosensor array;a fiber optic plate between the second radiation converter and the photosensor array, the photosensor array between the first radiation converter and the fiber optic plate, the fiber optic plate being a substrate for the photosensor array where a surface of the first radiation converter faces the photosensor array and a surface of the second radiation converter faces the fiber optic plate,the first radiation converter being configured to:receive and partially absorb incident penetrating radiation directed towards the radiation detector; andconvert the absorbed incident radiation into a burst of a plurality of light photons, a number of which reach the photosensor array and are detected;the second radiation converter being configured to:receive and partially absorb the portion of the incident radiation transmitted through the first radiation converter, the photosensor array and the fiber optic plate; andconvert the absorbed radiation into a burst of a plurality of light photons, a number of which reach the photosensor array and are detected,the photosensor array being configured to:respond a spatial pattern of the light photons from the first radiation converter and the second radiation converter by converting the light photons into an electrical signal pattern representative of a sum of the spatial pattern of the light photons from the first radiation converter and the spatial pattern of the light photons from the second radiation converter. 14. The radiation detector of claim 13, wherein the photosensor array comprises a first metal layer and a second metal layer, the first metal layer being directly in contact with the first radiation converter or being directly attached to the first radiation converter using an optical adhesive and the second metal layer being directly in contact with the fiber optic plate or being directly attached to the fiber optic plate using an optical adhesive. |
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053612825 | claims | 1. A method for manufacturing a dimensionally stable and corrosion-resistant fuel channel, comprising the steps of: forming first and second strips of material comprising metal alloy in a hexagonal close-packed crystallographic phase having a uniform texture; subjecting said first and second strips to heat treatment by heating to a first temperature which initiates transformation from said hexagonal close-packed crystallographic phase to a body-centered cubic crystallographic phase and then quenching to a second temperature at a rate which initiates transformation to a hexagonal close-packed crystallographic phase having a texture factor f.sub.L =0.28-0.38; forming each of said heat-treated first and second strips into first and second fuel channel components respectively; seam welding said first and second fuel channel components to assemble a fuel channel; and subjecting said fuel channel to thermal sizing by heating to a third temperature less than said second temperature and sufficient to anneal said material and then cooling. forming first and second strips of material comprising metal alloy in a hexagonal close-packed crystallographic phase having a uniform texture; subjecting said first and second strips to heat treatment by heating to a first temperature in the temperature range of 1800.degree. F. to 2050.degree. F., holding at a temperature no less than said first temperature for no less than 0.25 sec and then quenching to a second temperature not greater than 1500.degree. F.; warm forming each of said heat-treated first and second strips into first and second fuel channel components respectively by heating said first and second strips to a third temperature sufficient to increase the ductility of said strip material prior to forming; seam welding said first and second fuel channel components to assemble a fuel channel; and subjecting said fuel channel to thermal sizing by heating to a fourth temperature in a temperature range of 1000.degree. F. to 1250.degree. F., holding at a temperature no less than said fourth temperature for no less than 15 minutes and then cooling, wherein said third temperature is less than said fourth temperature. forming first and second strips of material comprising metal alloy in a hexagonal close-packed crystallographic phase having a uniform texture; subjecting said first and second strips to heat treatment by heating to a first temperature which initiates transformation from said hexagonal close-packed crystallographic phase to a body-centered cubic crystallographic phase and then quenching to a second temperature at a rate which initiates transformation to a hexagonal close-packed crystallographic phase having a texture factor f.sub.L =0.28-0.38; forming each of said heat-treated first and second strips into first and second fuel channel components respectively; seam welding said first and second fuel channel components to assemble a fuel channel; and subjecting said fuel channel to thermal sizing by heating to a third temperature less than said second temperature and sufficient to anneal said material and then cooling. 2. The method as defined in claim 1, wherein said first temperature lies in the temperature range of 1800.degree. F. to 2050.degree. F., said first and second strips are held at a temperature in said temperature range for at least 0.25 sec and said second temperature is no more than 1500.degree. F. 3. The method as defined in claim 1, wherein at least one of said heat treatment and thermal sizing steps is carried out in an inert atmosphere or in a vacuum. 4. The method as defined in claim 1, wherein said quenching is performed using an inert gas as the quenching medium. 5. The method as defined in claim 4, wherein said inert gas is helium. 6. The method as defined in claim 1, wherein said quenching is carried out at a rate of 10.degree. F./sec to 400.degree. F./sec. 7. The method as defined in claim 1, wherein said third temperature is in a temperature range of 1000.degree. F. to 1250.degree. F. and said fuel channel is held at a temperature in said temperature range for at least 15 minutes. 8. The method as defined in claim 1, wherein after said heat treatment and before said first and second fuel channel components are formed, said first and second strips are heated to a fourth temperature sufficient to increase the ductility of said strip material, wherein said fourth temperature is less than said third temperature. 9. The method as defined in claim 8, wherein said fourth temperature lies in a temperature range of 50.degree. F. to 800.degree. F. 10. A method for manufacturing a dimensionally stable and corrosion-resistant fuel channel, comprising the steps of: 11. The method as defined in claim 10, wherein said quenching is performed using an inert gas as the quenching medium. 12. The method as defined in claim 10, wherein said quenching is carried out at a rate of 10.degree. F./sec to 400.degree. F./sec. 13. The method as defined in claim 10, wherein said third temperature lies in a temperature range of 50.degree. F. to 800.degree. F. 14. A dimensionally stable and corrosion-resistant fuel channel made of metal alloy having a hexagonal close-packed crystallographic phase which is stable in a first temperature range and a body-centered cubic crystallographic phase which is stable in a second temperature range different than said first temperature range, formed by the steps of: 15. The fuel channel as defined in claim 14, wherein said metal alloy is zirconium-based alloy. 16. The fuel channel as defined in claim 14, wherein said first temperature lies in the temperature range of 1800.degree. F. to 2050.degree. F., said first and second strips are held at a temperature in said temperature range for at least 0.25 sec and said second temperature is no more than 1500.degree. F. 17. The fuel channel as defined in claim 14, wherein said quenching is carried out at a rate of 10.degree. F./sec to 400.degree. F./sec. 18. The fuel channel as defined in claim 14, wherein said third temperature is in a temperature range of 1000.degree. F. to 1250.degree. F. and said fuel channel is held at a temperature in said temperature range for at least 15 minutes. 19. The fuel channel as defined in claim 14, wherein after said heat treatment and before said first and second fuel channel components are formed, said first and second strips are heated to a fourth temperature sufficient to increase the ductility of said strip material, wherein said fourth temperature is less than said third temperature. 20. The fuel channel as defined in claim 19, wherein said fourth temperature lies in a temperature range of 50.degree. F. to 800.degree. F. |
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047643056 | summary | BACKGROUND OF THE INVENTION The present invention relates to a process for conditioning radioactive or toxic waste in epoxy resins. Over the past few years processes have been developed for the conditioning of radioactive or toxic waste in thermosetting resins consisting of introducing the waste into a polymerizable mixture, e.g. constituted by an epoxy resin and a hardener and then allowing the resin to polymerize to obtain a solid block within which are confined the radioactive or toxic wastes. More specifically, the present invention relates to polymerizable mixtures of this type based on epoxy resins, more particularly usable for the treatment of certain waste materials, such as large solid waste and organic liquids. Thus, the conditioning of large solid radioactive waste materials causes problems due to the phenomenon of the epoxy resin shrinking during hardening. These problems have hitherto been solved by adding an inert filler, such as e.g. sand to the polymerizable mixture and as described in French patent No. 2 361 725. However, this has led to difficulties in performing the process. In the case of contaminated organic liquids, it is possible to condition these in thermosetting resins, as described in French Pat. No. 2 230 041, but the treatment of such liquids leads to problems relating to obtaining a homogeneous mixture. SUMMARY OF THE INVENTION The present invention relates to a polymerizable mixture of an epoxy resin and a hardener, which also contains a special additive making it possible to solve the problems resulting from the conditioning of waste constituted by large size objects or organic liquids. The inventive toxic or radioactive waste conditioning process is characterized in that it comprises incorporating said waste into a polymerizable mixture containing at least one epoxy resin, pitch and at least one hardener for the epoxy resin and allowing the mixture to harden to obtain a solid block. According to a preferred embodiment, the mixture consists of 30 to 45% by weight epoxy resin, 30 to 50% by weight pitch and 20 to 25% by weight hardener. Generally the mixture contains at the most 50% by weight pitch. The addition of pitch to polymerizable mixtures based on epoxy resin used in the prior art in particular makes it possible to increase the fluidity of the polymerizable composition, which makes the same more suitable for the coating of large solid waste, such as filter cartridges, tools, metal filings placed in a basket and the like. The presence of pitch also makes it possible to improve the compatibility of the epoxy resins with organic liquids, such as drainage oils, distillation residues, scintillation liquids and solvents such as tributyl phosphate or xylene, which are generally immiscible with thermosetting resins. Thus, the polymerizable mixture according to the invention is of considerable interest for the treatment of large solid waste and organic liquids. Moreover, in view of the fact that the presence of pitch is not prejudicial to the quality of the finally obtained solid block, the polymerizable mixtures according to the invention can be used for conditioning other waste, which makes it possible to reduce the treatment or conditioning costs, because the price of the pitch is lower than that of the epoxy resins. The pitches used in the invention can be those obtained from the distillation of carbonization tars, particularly tars having a low content of insoluble products. These pitches must be compatible with the epoxy resin and hardener used and are also chosen as a function of the type of waste treated. For example, it is possible to use the pitch marketed under reference 730/30, which is a liquid coal pitch without volatile fractions and obtained from the treatment of tars having low insoluble product contents. This tar has the following characteristics: water content : traces density at +20.degree. C. : 1.184 viscosity at 30.degree. C. : 30 poises fractional distillation: fraction before 200.degree. C. : <0.5% PA1 fraction from 200.degree. to 250.degree. C. : 6% PA1 fraction from 250.degree. to 300.degree. C. : 18% PA1 fraction from 300.degree. to 350.degree. C. : 16% distillation residue : 60% insoluble in benzene : 10%. It is also possible to use the pitch marketed under reference 710/25 and which differs from pitch 730/30 by its viscosity of 25 poises and different distillation fractions. These pitches are compatible with epoxy resins in substantially all proportions. They are inexpensive and when added to epoxy resins in a quantity of no more than 50% by weight of the polymerizable mixture, it is possible to maintain the interesting characteristics of the epoxy resins, such as infusibility, high mechanical properties and long life. These pitch--epoxy resin mixtures can also harden at ambient temperature, in the same way as the polymerizable mixture based on epoxy resin. The viscosity of the epoxy resin--pitch mixtures is lower than that of mixtures of epoxy resin and sand used in the prior art for the treatment of large solid radioactive waste. Thus, such waste can be conditioned under better conditions, particularly as regards the preparation and transfer of the coating mixture. During the hardening of such mixtures, chemical bonds are established between the pitch and the epoxy reagents. Thus, this participation of the pitch in the polymerization gives the matrix a higher stability than in the case where use is made of mixtures of epoxy resin and sand, optionally also containing solvents and/or plasticizers. The addition of pitch to the polymerizable epoxy resin mixture gives the resin a thermoplastic character, which increases with the pitch content of the mixture. Thus, the hardness of the mixture decreases when the pitch content increases. In the same way the hardness of the mixture decreases when the temperature increases and this effect increases with the pitch content of the mixture. Moreover, to maintain adequate hardness characteristics in the solid blocks obtained, namely a Shore hardness of at least 50D, it is preferable for the added pitch quantity to represent no more than 40% by weight of the polymerizable mixture. Thus, due to the chemical interactions between the pitch and the epoxy resin, a pitch proportion, e.g. up to 30% by weight in the mixture has no influence on the hardness. However, above a 30% by weight proportion, the hardness rapidly decreases, because the excess pitch acts as an inert plasticizer. In certain cases, the pitch can also act as a solvent, particularly with respect to certain liquid waste materials which can only be incorporated with difficulty into the epoxy resins. Generally for putting into effect the process according to the invention, use is made of a first liquid constituent consisting of pitch and hardener, as well as a second liquid constituent based on the epoxy resin. The two liquid constituents are mixed with the liquid waste, so as to obtain a homogeneous mixture. The mixture is then allowed to harden to obtain a solid block. In the case of solid waste, the mixture is poured into the container and its low viscosity makes it possible to penetrate the interstices of the solid waste. It is possible to perform these operations at ambient temperature, as in the case of polymerizable mixtures based on epoxy resin and hardener of the prior art. Thus, the reaction is exothermic and the evolution of the temperature in the reaction medium is the same as that obtained when using mixtures containing solely an epoxy resin and a hardener. In order to obtain a 200.iota. block, a maximum temperature of approximately 100.degree. C. is obtained in 7h when the mixture contains 40% by weight of pitch. However, the presence of the pitch considerably slows down the final crosslinking of the epoxy resins and three weeks to a month may be necessary to obtain the desired hardness. However, it is possible to activate this phenomenon by carrying out the hardening of the mixture at an initial temperature exceeding 20.degree. C. Nevertheless, for a given volume, the maximum temperature must not exceed the boiling temperatures of the constituents. For putting into effect the process according to the invention, it is also possible to directly mix the pitch, the epoxy resin and the hardener with the waste to be treated at the time of use, whilst then leaving the mixture to harden so as to obtain a solid block. In the latter case, the hardening can be carried out at a temperature of 20.degree. to 60.degree. C. It is preferable not to mix the pitch and epoxy resin beforehand, because these can react and lead to an increase of the viscosity during the storage of the mixture due to slow polymerization. However, it is possible to add the pitch to the hardener beforehand, so that at the time of coating the waste there is a mixture of two liquid constituents, which facilitates the actual coating process. Thus, the invention also relates to a polymerizable mixture with two liquid constituents usable for the conditioning of toxic or radioactive waste comprising a first liquid constituent incorporating at least one epoxy resin and a second liquid constituent incorporating a mixture of pitch and at least one hardener of the epoxy resin, the pitch proportion in the second liquid constituent being 53 to 73% by weight. It is preferably fixed at 64% by weight, which corresponds to a polymerizable mixture containing 40% by weight pitch. The use of such a mixture with two liquid constituents is particularly advantageous for the treatment of radioactive waste constituted by large objects, because it is easier to use than the mixture of three constituents based on resin, sand and hardener used in the prior art, which had to be prepared in a continuous mixer at the time of the coating operation. When using the mixture of two liquid constituents according to the invention, a static mixer can be used, i.e. a simpler and less complicated device. Generally, the pitch is introduced into the polymerizable mixture or into the second liquid constituent in the form of a solution in an appropriate solvent. Thus, pitch is a solid compound and for introducing it in the form of a liquid constituent, a preferably slightly volatile solvent is added thereto, such as chrysanthemum or anthracene oil and the viscosity of the solution is adjusted by regulating the added solvent quantity. The epoxy resins used in the polymerizable mixture according to the invention can e.g. be bis-phenol A diglycidyl ethers and their viscosity can be regulated by adding a diluent reactive at low vapour tension, such as neopentyl diglycidyl ether. The hardeners used with the resins of this type can be constituted by compounds having at least one NH.sub.2 group, e.g. by cycloaliphatic or aromatic amines, aromatic or cycloaliphatic polyamines and derivatives of propylene amine. It is also possible to use polyamino amides. Preferably compounds are chosen having a mixture of cycloaliphatic and aromatic amines, which permits polymerization of the resin to take place at any pH. The hardener can be constitued by a compound of this type in the pure state or dissolved in an appropriate diluent, such as benzyl alcohol. Generally preference is given to the use of hardeners constituted by adducts, which are the product of the reaction of a small amount of epoxy resin with the compound having at least one NH.sub.2 group, to which is optionally added a cycloaliphatic polyamine, a non-reactive diluent and/or a hardening accelerator. The non-reactive diluent can be benzyl alcohol and the hardening accelerator the product of the reaction of acrylic acid benzoic acid, salicylic acid or a phenol, such as resorcinol with diaminodiphenyl methane. Preferably, the hardener is an adduct of diaminodiphenyl methane and epoxy resin, which may optionally also contain a cycloaliphatic polyamine. The hardener quantity present in the polymerizable mixture for obtaining the polymerization and crosslinking of the epoxy resin is dependent on the epoxy resin used and in particular its epoxy equivalent, i.e. the resin mass containing an epoxy function. In general, to obtain the hardening and crosslinking of the eopxy resin, use is made of a hardener quantity such that there is at least one NH.sub.2 amine function per epoxy resin equivalent, the hardener quantity being such that the hardener:epoxy resin weight ratio exceeds 0.5 and is e.g. equal to 0.6. The hardener--epoxy resin mixture described hereinbefore must have a low exothermicity in order to permit mass polymerization, so that in all cases the maximum temperature remains below the boiling temperature of the constituents, e.g. 90.degree. C. for a 200 liter barrel. The waste which can be conditioned by the process of the invention can be in various forms. Thus, they can be constituted by solid waste, such as powders of evaporation concentrates, liquid effluents, pulverulent products, ash from the incineration of fuel waste and large objects such as filter cartridges, tools or metal filings in a basket. The process according to the invention can also be used for treating organic liquid waste, such as aliphatic or aromatic hydrocarbons, chlorinated solvents, extraction solvents such as tributyl phosphate and trilauryl amine, drainage oils and scintillation liquids used for beta counting. However, in view of the hydrophobic nature of pitch, it is difficult to use the process of the invention for treating waste having a high water content and aqueous liquid waste. However, filter cartridges having a 100% water content can be treated by the inventive process, because this essentially corresponds to the water saturation of the filters. The quantity of waste incorporated into the polymerizable mixture according to the invention is of the same order as that which can be incorporated in the prior art polymerizable mixtures. In the case of most waste, the incorporated quantity can represent 40 to 60% of the total formed by the waste and the polymerizable mixture. However, in the case of certain liquid waste, it is necessary not to exceed certain contents, because otherwise there can be a decanting of the liquids during the mixing operation and/or a sweating phenomenon during the polymerization operation. Moreover, the quantities which can be incorporated also depend on the pitch content of the mixture and the nature of the epoxy resin, the hardener and the pitch used in the polymerizable mixture. According to a variant of the process according to the invention, around the solid block constituted by the waste incorporated into the hardened mixture of epoxy resin, pitch and hardener is formed a protective barrier produced from a polymerizable mixture containing at least one epoxy resin, pitch and at least one epoxy resin hardener. This makes it possible to form a barrier with respect to the diffusion of active or toxic elements, particularly when the solid waste is close to the walls of the hardened block. Furthermore, this barrier is very effective with respect to the diffusion of tritium and tritium-added water. In this variant, it is possible to firstly form a hardened block by incorporating the waste into the polymerizable mixture, followed by the inclusion of this hardened block in a hollow cylindrical barrel produced from a polymerizable mixture incorporating an epoxy resin, pitch and a hardener. In this case, the waste and the polymerizable mixture incorporating at least one epoxy resin, pitch and at least one hardener for the epoxy resin are introduced into a cylindrical barrel obtained by hardening a polymerizable mixture incorporating at least one epoxy resin, the pitch and at least one epoxy resin hardener and said polymerizable mixture is allowed to harden to obtain a solid block within said barrel. The latter can be prepared by conventional methods, e.g. by casting in a mould, within which is placed an inner core, or alternatively by centrifuging. This protective barrier can also be formed during the conditioning of radioactive or toxic waste by placing the solid waste in a basket, such as a metal basket located in a barrel, so as to provide an adequate thickness between the basket and the barrel. In this case, the waste is firstly introduced into a basket, the basket containing the waste is placed in a barrel so as to leave a space between the inner barrel wall and the outer basket wall, the barrel and basket are filled with the polymerizable mixture containing at least one epoxy resin, pitch and at least one epoxy resin hardener and said mixture is allowed to harden to obtain in said barrel a solid block incorporating an outer layer formed solely from the polymerizable mixture based on epoxy resin, pitch and hardener. Thus, on introducing the polymerizable fluid mixture into the basket, this passes through the perforations of the basket so as to fill the space between the barrel and the basket and then form by hardening the outer layer constituting the protective barrier. |
claims | 1. An apparatus for preparing a small amount of a radioactive substance combination, comprising:a one-part body,a mixing device integrated in the body and adapted to receive a small amount of chemical substances, andat least one receptacle integrated in the body and connected to the mixing device and adapted to hold a small amount of a chemical substance. 2. The apparatus as claimed in claim 1, wherein one or more of the at least one receptacle has a volume of less than 100 μl. 3. The apparatus as claimed in claim 1, further comprising: one or more conduits integrated in the apparatus and having a volume of less than 100 μl. 4. The apparatus as claimed in claim 3, wherein each conduit has a height of less than 100 μm and a width of less than 500 μm. 5. The apparatus as claimed in claim 2, wherein the mixing device and each of the at least one receptacle are interconnected by a corresponding at least one conduit. 6. The apparatus as claimed in claim 1, wherein the mixing device is a conduit. 7. The apparatus as claimed in claim 1, wherein the mixing device is selected from the group including cascade mixers, diffusion mixers, lamination mixers, mixers operating according to the split-recombine principle, mixers operating with the assistance of alternating electrical fields or with sonic or vibratory support. 8. The apparatus as claimed in claim 5, wherein the mixing device has a holding capacity of less than 100 μl. 9. The apparatus as claimed in claim 1, wherein the apparatus is adapted to be closed and sealed towards the outside. 10. The apparatus as claimed in claim 1, further comprising: at least one access for a sensor, and/or one mechanical interface with the outside. 11. The apparatus as claimed in claim 1, further comprising:at least one conveying and dosing means, respectively, selected from the group including: means or a part of means operating with the use of centrifugal force, electrical force acting on a fluid, pressure or volume variation, or a conveying method functioning with sonic or vibratory support. 12. The apparatus as claimed in claim 1, further comprising:at least one measuring or sensor means for determining a physical magnitude from among the group including: the type and power of radioactive radiation, pH, and temperature; a means for carrying out chromatography or electrophoresis; and/or a means for detecting refraction of light and/or detecting at least one property of a substance from among the group including the presence or absence, quantity, color, and index of refractivity, an ion exchange column, a size exclusion column or part thereof. 13. The apparatus as claimed in claim 1, wherein the apparatus is controllable from outside and further comprises means for receiving control and/or power supply signals. 14. The apparatus as claimed in claim 1, wherein at least parts of the apparatus are manufactured as a monolith via micro process engineering, including micro injection molding or micro embossing. 15. The apparatus as claimed in claim 1, wherein, at the inside, the apparatus comprises coated surfaces to prevent substances from adhering or for catalyzing chemical reactions. 16. The apparatus as claimed in claim 1, further comprising: a heating and/or cooling device or part thereof 17. The apparatus as claimed in claim 1, wherein the apparatus is made of plastics, polyethylene, polypropylene, polymethylmethacrylate, cyclo-olefin-copolymer (COC), polytetrafluoroethylene, polycarbonate, silicon, metal, or glass. 18. A system, comprising:an apparatus for preparing a small amount of a radioactive substance combination, wherein the apparatus comprises:a one-part body;a mixing device integrated in the body and adapted to receive a small amount of chemical substances; andat least one receptacle integrated in the body and connected to the mixing device and adapted to hold a small amount of a chemical substance; anda control unit adapted to be coupled to the apparatus for control thereof. 19. The system as claimed in claim 18, further comprising:an isotope source adapted to be coupled to the apparatus, andanother source of chemical substances not integrated in the apparatus. 20. The system as claimed in claim 18, further comprising:a conveying and dosing means, respectively, adapted to be coupled to the apparatus;measuring or sensor means;a means for carrying out chromatography or electrophoresis; and/ora means for detecting refraction of light and/or detecting at least one property of a substance from among the group including the presence or absence, quantity, color, and index of refractivity, an ion exchange column, a size exclusion column or part thereof. 21. A method of preparing a small amount of a radioactive substance combination comprising:providing an apparatus adapted for preparing a small amount of a chemical substance combination, wherein the apparatus comprises:a one-part body,a mixing device integrated in the body and adapted to receive a small amount of chemical substances, andat least one receptacle integrated in the body and connected to the mixing device and adapted to hold a small amount of a chemical substance;supplying a small amount of at least one substance into the mixing device of the apparatus;supplying a small amount of at least one radioactive substance into the mixing device;mixing the at least one substance with the at least one radioactive substance; andwithdrawing the resulting substance combination. 22. The method as claimed in claim 21, wherein supplying of a small amount of the at least one substance comprises supplying an amount of less than 1 ml. 23. The method as claimed in claim 21, further comprising: supplying a substance from a receptacle which is integrated in the apparatus. 24. The method as claimed in claim 21, wherein at least one substance is introduced into a receptacle of the apparatus when the apparatus is manufactured. 25. The method as claimed in claim 21, wherein supplying a small amount of at least one radioactive substance comprises supplying the radioactive substance from outside the apparatus. 26. The method as claimed in claim 21, further comprising: cooling or heating of substances. 27. The method as claimed in claim 21, further comprising: performing quality control of one or more substances or substance combinations in the apparatus. 28. The method as claimed in claim 21, further comprising: performing quality control of the resulting radiochemical substance combination prior to withdrawal. 29. The method as claimed in claim 21, wherein performing quality control comprises performing size exclusion chromatography, ion exchange chromatography, and/or thin film chromatography. 30. The method as claimed in claim 21, wherein mixing the at least one substance with the at least one radioactive substance comprises radioactive labeling of biomolecules with isotopes. 31. The method as claimed in claim 21, wherein mixing the at least one substance with the at least one radioactive substance results in a chemical bond between the at least one radioactive substance and the at least one substance. 32. The method as claimed in claim 21, wherein the at least one radioactive substance is selected from the group including Me2+, Me3+, MeO4−, and halogens. 33. The method as claimed in claim 21, wherein the at least one radioactive substance is selected from the following group: cobalt-57, cobalt-58, selenium-75, gallium-67, gallium-68, iodine-123, iodine-124, iodine-125, iodine-131, astatine-211, actinium-225, bismuth-212, bismuth-213, lead-212, technetium-99m, rhenium-186, rhenium-188, silver-111, indium-111, platinum-197, palladium-109, copper-67, phosphorus-32, phosphorus-33, yttrium-90, scandium-47, samarium-153, ytterbium-169, lutetium-177, rhodium-105, praseodymium-142, praseodymium-143, terbium-161, holmium-166, thallium-201, or gold-199. 34. The method as claimed in claim 21, further comprising: supplying a buffer solution selected from the group including acetate, citrate, phosphanate, carbonate, HEPES, MES, or other acceptable buffer solutions. 35. The method as claimed in claim 21, further comprising: controlling and monitoring the method sequence by means of a control unit adapted to be coupled to the apparatus. |
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051924935 | abstract | A system for improving the performance of nuclear power plant feedwater control systems and simplifying steam generator low water level reactor protection logic includes a plurality of water level channels for measuring steam generator water level and generating a plurality of signals representative thereof. The median steam generator water level signal is selected from among the plurality of steam generator water level signals. The median steam generator water level signal is then communicated to the feedwater control system through an output interface. |
claims | 1. A shielded X-ray radiation apparatus for analysis of materials via gamma activation analysis, the apparatus comprising:an X-ray conversion target, wherein, when an electron accelerator directs an electron beam having an electron beam direction at the X-ray conversion target, the X-ray conversion target generates X-rays;an X-ray attenuation shield including an elongate cavity to house the X-ray conversion target and incorporating a region to accommodate a sample;a neutron attenuation shield;a gamma attenuation shield; anda support casing housed within the X-ray attenuation shield and configured to support the X-ray conversion target and the sample,wherein the neutron attenuation shield is adjacent to and substantially surrounds the X-ray attenuation shield; andwherein the gamma attenuation shield is adjacent to and substantially surrounds the neutron attenuation shield. 2. The shielded X-ray radiation apparatus according to claim 1, wherein the X-ray conversion target is configured to generate the X-rays when the electron beam has an energy between 7 MeV and 15 MeV. 3. The shielded X-ray radiation apparatus according to claim 1, wherein the thickness of the X-ray attenuation shield at an angle of 90° from the electron beam direction measured relative to the location of the conversion target is within the range of 60-80% of the thickness in the forward direction, and the thickness at an angle of 180° from the electron beam direction measured relative to the location of the conversion target is within the range of 25-50% of the thickness in the forward direction. 4. The shielded X-ray radiation apparatus according to claim 3, wherein the thickness of the X-ray attenuation shield at an angle of 90° from the electron beam direction measured relative to the location of the conversion target is approximately 75% of the thickness in the forward direction, and the thickness at an angle of 180° from the electron beam direction measured relative to the location of the conversion target is approximately 50% of the thickness in the forward direction. 5. The shielded X-ray radiation apparatus according to claim 1, wherein a first thickness (tXR) of the X-ray attenuation shield thickness in the forward direction is estimated by the equation: tXR=TVL×log10[(R×60×106)/(r d2)], where d is the distance from the X-ray conversion target, R is the X-ray dose rate at 1 m from the X-ray conversion target produced by the X-ray conversion target, r is the shielded X-ray dose rate at the closest personnel-accessible point, and TVL is a predefined tenth value layer thickness for the X-ray attenuation shield material. 6. The shielded X-ray radiation apparatus according to claim 1, wherein the neutron attenuation shield has a thickness (tnt) in the forward direction which is estimated by the equation: tnt=TVLn log10 (f) where TVLn is a predefined tenth-value layer thickness for the attenuation of low energy neutrons in the neutron attenuation shield and f is a ratio of an unshielded neutron dose rate to a desired shielded neutron dose rate. 7. The shielded X-ray radiation apparatus according to claim 1, wherein the thickness of the neutron attenuation shield at an angle of 180° from the electron beam direction measured relative to the location of the conversion target in a rearward direction is 50% to 100% of the thickness (tnt) in the forward direction. 8. The shielded X-ray radiation apparatus according to claim 1, wherein the gamma attenuation shield has a thickness which is proportional to the thickness of the neutron attenuation shield. 9. The shielded X-ray radiation apparatus according to claim 1, wherein the X-ray attenuation shield is constructed from lead, tungsten, or layers of lead and tungsten. 10. The shielded X-ray radiation apparatus according to claim 1, wherein the neutron attenuation shield is constructed from a polymer material having a hydrogen density of approximately 0.1 g/cm3. 11. The shielded X-ray radiation apparatus according to claim 10, wherein the polymer material comprises boron or lithium. 12. The shielded X-ray radiation apparatus according to claim 1, wherein the gamma attenuation shield is constructed from lead, steel, or a composite of lead and steel. 13. The shielded X-ray radiation apparatus according to claim 1, wherein:the X-ray attenuation shield has a thickness in the electron beam direction calculated using a tabulated tenth value layer thickness for X-rays, the X-ray dose-rate output of the X-ray conversion target, and a desired X-ray dose rate outside of the shield, wherein the thickness of the X-ray attenuation shield decreases with increasing angle from the electron beam direction measured relative to the location of the conversion target; andthe neutron attenuation shield has a thickness in the electron beam direction calculated using a tabulated tenth value layer thickness for neutrons, the neutron dose-rate output of the X-ray conversion target, and a desired neutron dose rate outside of the shield, wherein the thickness of the neutron attenuation shield decreases with increasing distance from the X-ray conversion target. 14. The shielded X-ray radiation apparatus according to claim 1, further comprising the electron accelerator, wherein the electron accelerator is housed within the X-ray attenuation shield, the neutron attenuation shield, and the gamma attenuation shield. 15. The shielded X-ray radiation apparatus according to claim 14, wherein each of the X-ray attenuation shield, the neutron attenuation shield, and the gamma attenuation shield comprises:a relatively wide head portion housing the X-ray conversion target and the sample; anda relatively narrow tail portion housing at least part of the electron accelerator. 16. The shielded X-ray radiation apparatus according to claim 1, further comprising the electron accelerator, wherein at least a forward portion of the electron accelerator is housed within the support casing, the X-ray attenuation shield, the neutron attenuation shield, and the gamma attenuation shield. 17. The shielded X-ray radiation apparatus according to claim 1, wherein the support casing supports the sample and the X-ray conversion target in correct relative positions. 18. A shielded X-ray radiation apparatus for analysis of materials via gamma activation analysis, the apparatus comprising:an X-ray conversion target, wherein, when an electron accelerator directs an electron beam having an electron beam direction at the X-ray conversion target, the X-ray conversion target generates X-rays;an X-ray attenuation shield including an elongate cavity to house the X-ray conversion target and incorporating a region to accommodate a sample;a neutron attenuation shield;a gamma attenuation shield; anda removable plug configured to insert the sample into the elongate cavity;wherein the removable plug comprises adjacent blocks of material, each block having a thickness and a composition which substantially matches thickness and composition of a corresponding one of the X-ray attenuation, neutron attenuation, and gamma-ray attenuation shields;wherein the neutron attenuation shield is adjacent to and substantially surrounds the X-ray attenuation shield; andwherein the gamma attenuation shield is adjacent to and substantially surrounds the neutron attenuation shield. 19. The shielded X-ray radiation apparatus according to claim 18, wherein the removable plug further comprises a stage member on which the sample to be irradiated is locatable; and wherein the adjacent blocks of materials comprise a first block adjacent the stage member, a second block abutting the first block, and a third block abutting the second block; wherein:the first block comprises a material to substantially attenuate X-rays and having a thickness which is the same or substantially the same as the X-ray attenuation shield,the second block comprises a material to substantially attenuate neutrons and having a thickness which is the same or substantially the same as the neutron attenuation shield, andthe third block comprises a material to substantially attenuate gamma-rays with a thickness which is the same or substantially the same as the gamma attenuation shield; andwherein the shielded X-ray radiation apparatus comprises a sleeve through which the stage member is able to traverse. 20. The shielded X-ray radiation apparatus according to claim 19, wherein the sleeve of the apparatus and the removable plug have a clearance tolerance of less than 2.00 mm. 21. The shielded X-ray radiation apparatus according to claim 19, wherein the stage member of the removable plug is constructed from a substance that is free from elements that undergo significant activation from X-rays or neutrons. 22. The shielded X-ray radiation apparatus according to claim 19, wherein an outer profile of the removable plug is stepped, with at least one of the adjacent blocks increasing in width, in a direction perpendicular to a direction of travel of the removable plug with increasing distance from the stage member. 23. The shielded X-ray radiation apparatus according to claim 22, wherein the width of the or each step is in the range of 5 to 15 mm. 24. The shielded X-ray radiation apparatus according to claim 22, wherein the first block comprises at least two steps, such that the width of the first blocks increase in a stepwise fashion from the innermost to outermost steps. |
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summary | ||
053348435 | claims | 1. A scintillator screen for an X-ray system comprising: a substrate of low-Z, non-scintillating material; and bodies associated with the substrate including a high-Z, scintillating material adapted to scintillate upon exposure to photoelectrons produced by X-rays passing through the high-Z material of the bodies, the bodies being of such size and being spatially separated from one another by the low-Z material of the substrate to provide the screen with a predetermined sensitivity to X-rays within a preselected energy range. a second substrate positioned adjacent said first substrate, said second substrate including low-Z material; and second bodies associated with the second substrate including high-Z material, at least one of said second substrate and said second bodies being adapted to scintillate upon exposure to photoelectrons provided by X-rays passing through the high-Z material of said second bodies, said second bodies being of such a size and being spatially separated one from another by the low-Z material of said second substrate to provide a pre-selected sensitivity to X-rays within an alternative energy range. a substrate of low-Z material; and bodies comprised of fibers associated with the substrate including a high-Z material, at least one of the materials of the substrate and bodies adapted to scintillate upon exposure to photoelectrons produced by X-rays passing through the high-Z material of the bodies, the bodies being of such size and being spatially separated from one another by the low-Z material of the substrate to provide the screen with a predetermined sensitivity to X-rays within a preselected energy range. a substrate including lithium fluoride; and bodies associated with the substrate including barium fluoride, the bodies being of such size and being spatially separated from one another by the lithium fluoride of the substrate to provide the screen with a predetermined sensitivity to X-rays within a preselected energy range. a substrate including anthracene; and bodies associated with the substrate including heavy flint glass, the bodies being of such size and being spatially separated from one another by the anthracene of the substrate to provide the screen with a predetermined sensitivity to X-rays within a preselected energy range. a substrate of low-Z material; and bodies associated with the substrate including a quantity of leaded plastic coated with a low-Z material, at least one of the materials of the substrate and bodies adapted to scintillate upon exposure to photoelectrons produced by X-rays passing through the bodies, and the bodies are of such size and are spatially separated from one another by the low-Z material of the substrate to provide the screen with a predetermined sensitivity to X-rays within a preselected energy range. a substrate including lithium fluoride and bodies associated with the substrate including barium fluoride, the bodies having a thickness of about 10 microns as measured in a direction therethrough which corresponds generally to the direction of travel of an X-ray into the body, the bodies being uniformly dispersed throughout the substrate and occupying about 25% of the total volume of the screen, and the bodies being of such size and being spatially separated from one another to provide the screen with a predetermined sensitivity to X-rays within a preselected energy range. 2. A screen as defined in claim 1 wherein the bodies are dispersed uniformly throughout the substrate. 3. A screen as defined in claim 1 wherein the materials of the substrate and the bodies have indices of refraction which are within ten percent of each other. 4. A screen as defined in claim 1 wherein each of the bodies has a thickness between about 1.times.10.sup.-3 mm and 1.times.10.sup.-2 mm as measured in a direction therethrough which corresponds generally to the direction of travel of an X-ray into the bodies. 5. The screen as defined in claim 1 wherein the bodies are provided by grains embedded within the substrate. 6. A screen as defined in claim 5 wherein the high-Z material of the bodies comprises about 25% of the total volume of the screen. 7. The screen as defined in claim 1 wherein at least one of the bodies is provided by a layer of high-Z material juxtaposed with the substrate material. 8. A screen as defined in claim 7 wherein the high-Z material of the bodies comprises about 25% of the total volume of the screen. 9. A screen as defined in claim 1 wherein the high-Z material of the bodies comprises about 25% of the total volume of the screen. 10. A screen as defined in claim 1 wherein the bodies include gadolinium oxysulphide and the substrate includes plastic. 11. The screen of claim 1 further comprising: 12. A scintillator screen for an X-ray system comprising: 13. A screen as defined in claim 12 wherein the high-Z material of the bodies comprises about 25% of the total volume of the screen. 14. A scintillator screen for an X-ray system comprising: 15. A scintillator screen for an X-ray system comprising: 16. A scintillator screen for an X-ray system comprising: 17. A scintillator screen for an X-ray system comprising: |
abstract | A monochromator is adapted to select at least one band of wavelengths from diverging incident radiation. The apparatus includes a first crystal and a second crystal. A band of emitted wavelengths of the first crystal is adapted to the at least one band of wavelengths. A surface curvature of the first crystal is adapted to focus emitted radiation in a first plane. A band of emitted wavelengths of the second crystal also is adapted to the at least one band of wavelengths. Parallel faces of a lattice structure of the second crystal are oriented at a first predetermined angle from a surface of the second crystal. In another embodiment, an apparatus is adapted to select at least one band of wavelengths from diverging incident synchrotron radiation in a given range of wavelengths with an energy resolution finer than about five parts in 10000 and optical efficiency greater than about 50 percent. |
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description | This application claims priority from PCT Application No. PCT/US2005/030796, filed Aug. 30, 2005, which claims priority from provisional U.S. Application No. 60/605,481, filed Aug. 30, 2004. The invention is directed to improved containers for pharmaceuticals and the tubing and tubing connectors associated therewith, particularly containers for pharmaceuticals which are heated, irradiated or otherwise subjected to increased pressure. In a preferred embodiment, the invention is directed to an improved container for use in a radioisotope generator. Specifically, the designs and materials of the column container and its closure and associated tubing and tubing connectors have been improved. The invention includes improved pharmaceutical containers, particularly improved containers for pharmaceuticals that are subjected to increased pressure (such as by heating or other means) and/or are subjected to radioactivity. In a preferred embodiment, the invention is directed to an improved container, also called a column, for use in a radioisotope generator. In an especially preferred embodiment, the improved column is for use with rubidium-82 generator such as those disclosed in U.S. Pat. Nos. 3,953,567; 4,400,358; 4,406,877; 4,562,829; 4,585,009; 4,585,941; and 5,497,951, incorporated herein by reference in their entirety. In a particularly preferred embodiment, the improved column is used in a rubidium-82 generator such as that sold under the trade name CardioGen®. The improved pharmaceutical container of the invention includes an improved seal and crimping process, as well as changes to the design of the stopper and the container to prevent blockages and improve consistency in packing and closing the container, which improves flow rate and elution from the column. Further improvements include constructing the container and stopper out of radiation resistant or tolerant materials. In addition, flexible tubing used with the container is made of a radiation resistant or tolerant material, and the Luer locks used to fasten the flexible tubing to the container is made of a radiation resistant or tolerant material and is further improved to insure a tight, secure lock which will not inadvertently loosen or disconnect. Specifically, the improved container has a new, stronger seal which is used to crimp the stopper in a pharmaceutical container and particularly, which is used to seal a radioisotope generator column/stopper assembly system, such as the CardioGen® system. This improved seal prevents leakage, even at increased pressure, and reduces ballooning of the rubber stopper material. The seal has a configuration similar to one of those shown in FIG. 5B through FIG. 5F and FIG. 6 and is made of any suitably strong material including metal or plastic. A pneumatically operated automatic or semi-automatic crimper, set at optimized pressure, is preferably used to crimp the seal during assembly of a pharmaceutical container such as a radioisotope generator column/stopper assembly system. The invention includes identification of optimized crimping pressure(s) for crimping the seal (regardless of material) to a pharmaceutical container such as a glass or plastic vial or column and thus securing in place a rubber closure(s) when using an automatic crimping system and/or manual crimping. The stopper which is crimped into place is also improved. Specifically, it is made of a material which is radiation resistant or tolerant, is resistant to ballooning and can withstand the pressure at which the container operates. Additionally, the configuration and placement of the stopper at the bottom of the column reduces the “dead volume”—space where non-radioactive, decayed eluate could mix with (and dilute) fresh, radioactive eluate, reducing the efficacy of the eluent. The improved pharmaceutical container also includes improvements to the design which improve its packing/assembly and thus ensure specified flow of eluent through the container. These improvements are illustrated in the context of a radioisotope generator column container. Flow rate of the eluent through the column could be partially or completely blocked if the stopper blocks the outlet arm of the column. As shown in FIG. 1, the outlet arm of the container of the invention has been repositioned slightly and a small piece of plastic removed from the inside edge of the column to create a recess or notch where the outlet arm enters the column lumen to prevent a stopper from blocking flow. See FIG. 4. A small reinforcement piece of resin is added to the outside of the column between the outlet arm and column body to provide additional strength. Another improvement in the containers of the invention addresses consistency of assembly and packing of the containers. In prior columns for a radioisotope generator, a plastic basket or spacer was supplied separately and was placed on the top of the column packing before the seal was inserted and the seal crimped into place. In these prior columns, placement of the baskets or spacers, which hold the column packing in place, could vary significantly, potentially creating some problems with consistency in packing. In the improved columns, two small orientation knobs have been added to the outside of the top basket/spacer and the orientation knobs are positioned 180° apart. These knobs fit into two small slots cut into the wall of the column. This combination eliminates the potential variability of manual alignment and depth placement of the basket/spacer into the column and ensures a consistent fit every time. Critical to the function of the column is the alignment of the basket/spacer openings with the column inlet in the top arm. This prevents misalignment and consequent restricted flow and possible back pressure and also ensures consistent and timely out put of eluent to the patient. Another improvement is to make the column assembly out of a radiation resistant or tolerant material, such as radiation resistant polypropylene. Likewise, the flexible tubing and Luer connector are made of radiation resistant or tolerant materials, such as radiation resistant polyvinylchloride. Furthermore, the Luer connector on the flexible tube and its counterpart Luer connector on the column assembly are configured to provide for a tight lock which will not leak and which will not loosen or inadvertently disconnect during use. The invention was designed to solve a number of technical problems experienced with prior art pharmaceutical containers. 1. Leakage From the Stopper/Column Interface Leakage from the flange (or other area) of the seal of prior pharmaceutical containers such as column/stopper assembly systems was found to occur when the system was exposed to increasing pressure. The new seal, consisting of a stronger material crimped at optimized crimping pressure, prevents leakage at the flange seal area even at increasing pressure. 2. Ballooning Ballooning and/or burst of rubber materials (both before and after irradiation) through the center hole of current aluminum seals has been observed when they are subject to repeated pulsations of pressure cycling. The seals of the invention, which are stronger and are crimped at optimized pressure, reduce the likelihood of this problem. However, in a preferred embodiment the seal used in the improved container of the invention has a center hole of reduced size. For example, a seal with the configuration of those in FIG. 5B, FIG. 5C, FIG. 5E or FIG. 6 may preferably be used. Due to the small center hole and strength of these seals, and crimping at optimized pressure, ballooning and/or burst of rubber materials is prevented. Consequently, pharmaceutical containers of the invention, and particularly column/stopper systems of the invention, can be exposed to much higher pressures during use of the system in the field. In addition, the larger surface area of the crimp resulting from the reduction of the diameter of the center hole serves as additional support for the rubber closure and inhibits possible rupture as it is weakened over time due to the cumulative effect of exposure to radiation from the column or container content. Also, the stopper is made of a radiation resistant or tolerant material. This also helps prevent ballooning and bursting. 3. Leakage Through Puncture Points Leakage through puncture points has been observed in prior art pharmaceutical containers. Such leakage may be eliminated in containers of the invention through a combination of the stronger seal material, preferably a smaller center hole, and crimping at optimized pressure. 4. Splitting of the Seal Splitting or tearing of current aluminum seals has been observed at pressures intended for use with a pharmaceutical container system (or pressures to which the system can potentially be exposed during intended usage in the field). Due to the strength of the new seal material, no splitting or rupture of seal material is observed at pressures intended for use. For example, the seals on the columns of the invention do not split or rupture when used in, for example, a rubidium generator at intended pressures. 5. Inconsistent Manual Crimping Procedure The manual crimping procedure commonly used with many prior container systems, including radioisotope column systems, is not always consistent and thus does not result in reproducible crimping pressures. Over-pressuring results in buckling and collapse of the skirt of the seal material. Under-pressuring results in a loose overseal. Use of the automatic or semi-automatic crimping procedure of the invention with compressed or pressurized air results in consistent/reproducible crimping pressures, and enables selection of optimized crimping pressures when crimping various seal materials. 6. Maintenance of Consistent Flow/Reduction of Back Pressure In some prior pharmaceutical columns, flow rate of the eluent through the column could be partially or completely blocked because the stopper blocked the outlet arm of the column. The outlet arm of the container of the invention has been repositioned slightly and a small piece of plastic removed from the inside edge of the column to create a recess or notch where the outlet arm enters the column lumen to prevent a stopper from blocking flow. A small reinforcement piece of resin is added to the outside of the column between the outlet arm and column body to provide additional strength. The recessed outlet arm and notch near the bottom of the column body greatly reduces the chance of back pressure due to a stopper blocking the outlet arm. 7. Inconsistent Positioning within Column In a column for a radioisotope generator, a plastic basket or spacer is supplied separately and is placed on the top of the packed column before the seal or closure is inserted and the seal crimped into place. In prior columns, the baskets/spacers, which hold the column packing in place, were not easily positioned consistently both in terms of depth and orientation. In the improved columns of the invention, two small orientation knobs have been added to the outside of the top basket/spacer and these orientation knobs are positioned 1800 apart. These knobs fit into two small slots cut into the wall of the column. This combination eliminates the potential variability of manual placement of the basket into the column, ensuring a consistent fit from generator to generator and reducing the variability in packing density associated with this manual process. 8. Degradation due to Radiation Many materials degrade when exposed to radiation. Degradation includes possible changes in color, loss of flexibility, increased brittleness and the leaching out of various substances from the materials. To avoid these potential problems, the column assembly, stopper, flexible tubing and Luer connectors are made out of radiation resistant or tolerant materials. Frequently, when a material is said to be radiation resistant or tolerant, that means the material can withstand the amount of radiation used for sterilization, which is typically about 25 kGy. For the purposes of the present invention, however, a material is radiation resistant or tolerant when it can be exposed to about 145 kGy radiation and not degrade to the point where the functioning of the column assembly will be adversely affected. 9. Properly Closed Luer Locks Luer locks are known in the art. However, it can be difficult to determine when a Luer lock has been sufficiently tightened to form a tight, non-leaking lock. Thus, one improvement is to provide for one or more tabs on each Luer connector. When the tabs achieve a certain orientation with respect to each other, for example when the tabs line up, such orientation means that the Luer lock has been sufficiently tightened. Another potential difficulty with Luer locks is that they can come loose, i.e. disconnect, during use, which has the potential of causing a leak. To overcome this potential difficulty, the Luer connectors screw together and are each provided with one or more tabs. As the Luer connectors approach their fully tightened position, the tabs overlap. Further tightening causes the overlapping tabs to pass by each other, which can cause a clicking sound or sensation. When this occurs, the Luer lock is sufficiently tightened. Also, the Luer locks cannot become loose, e.g. unscrew, because the overlapping tabs will inhibit this action. Referring now to FIG. 1, FIG. 1A shows a side view and FIG. 1B shows a bottom view of the inventive container (e.g., column assembly) of one embodiment of the invention. FIG. 1C is another side view of the inventive column assembly, cut along line A-A of FIG. 1B. FIG. 1D is detail B from FIG. 1C, at a scale of 3:1 compared to FIG. 1C. FIG. 1E is a top view of the inventive column assembly, cut along line E-E of FIG. 1A. FIG. 1F is another side view of the inventive column assembly, cut along line C-C of FIG. 1B. FIG. 1G is detail D of FIG. 1F, at a scale of 2:1 compared to FIG. 1F. FIG. 1A has an inlet arm 1 which has an inlet arm female Luer cap 2 at its distal end. The proximal end of the inlet arm 1 attaches to the upper portion of a column 3. There is also an inlet arm support means 4 to support the inlet arm 1. The support means is preferably material which is added to support the inlet arm 1. Preferably, this material is the same material used to construct the column assembly. As shown, the inlet arm support means 4 is a triangular shaped member attached to the inlet arm 1 and the column 3, although the shape of the support is not limited to a triangle. It can be square, a bar passing from the inlet arm 1 to the column 3, or any other suitable shape. The column 3 has a top portion 5 and a bottom portion 6. The top portion 5 comprises a first top portion 7 and a second top portion 8. The first top portion 7 is on top of and has a diameter greater then the second top portion 8, which is on top of and has a greater diameter than the column 3. The bottom portion 6 of the column 3 has a similar configuration. It has a first bottom portion 9 and a second bottom portion 10. The first bottom portion 9 sits below and has a greater diameter than the second bottom portion 10, which sits below and has a greater diameter than the column 3. Also shown is a bottom stopper 11. An outlet arm 12 is attached to the bottom portion of the column 3. The distal end of the outlet arm 12 terminates in an outlet arm female Luer cap 13. There is also an outlet arm support means 14 to support the outlet arm 12. The support means is preferably material which is added to support the outlet arm 12. Preferably, this material is the same material used to construct the column assembly. As shown, the outlet arm support means 14 is a triangular shaped member which attaches to the column and the outlet arm 12, although the shape of the support is not limited to a triangle. It can be a square, a bar passing from the outlet arm 12 to the column 3, or any other suitable shape. FIG. 1C shows a cross section of the inventive column assembly, cut through line A-A of FIG. 1B. As shown, the inlet arm 1, column 3 and outlet arm 12 are hollow. Turning to the hollow interior or lumen of the column 3, it first defines a top stopper receptacle area 15. Below that and in communication with it is a top basket receptacle area 16. As shown in FIG. 1C, the top basket receptacle area 16 contains a top basket or spacer 17. Following that is a packing material containing area 18. Underneath the packing material containing area 18 is a bottom screen 19, followed by a bottom open area 20. Underneath the bottom open area 20 is a bottom stopper receptacle area 21. FIG. 1C shows the bottom stopper 11 inserted into the bottom stopper receptacle area 21 of the column 3. Note that the bottom stopper 11 consumes most of the bottom stopper receptacle area 21. This minimizes the dead volume in the bottom stopper receptacle area 21. Minimization of the dead volume minimizes mixing of fresh, radioactive eluent with non-radioactive or decayed eluent, which could dilute the fresh eluent, thereby maintaining a narrow rubidium-82 bolus profile. The inlet arm 1 and outlet arm 12 are each hollow, the hollow portions being 22 and 23 respectively, and are in communication with the hollow portion of the column 3. As shown in FIG. 1C, the hollow portion 22 of the inlet arm 1 is in communication with the top basket receptacle area 16. The intersection of the column 3 and the outflow arm 12 is shown in more detail in FIG. 1D. As shown therein, no portion of the outflow arm 12 extends into the hollow portion of the column 3, as was the case with certain prior art column assemblies. Also, the hollow portion 23 of the outflow arm 12 intersects the hollow portion of column 3 at the top of the bottom stopper receptacle area 21 or at about the place the bottom stopper receptacle area 21 and the bottom open area 20 intersect. This configuration, not found in prior art column assemblies, prevents the bottom stopper 11 from blocking the outflow arm 12. In a preferred embodiment, an outflow notch 25 is formed where the hollow portion 23 of the outflow arm 12 intersects the hollow interior of the column 3, thus further preventing any blockage of the outflow arm 12 by the bottom stopper 11. This embodiment is shown in more detail in FIG. 4. FIG. 1E is a top view of the inventive column assembly. Visible from this perspective are, for example, the top basket or spacer 17 and the top basket receptacle area 16. Also shown are notches 24a and 24b. The notches 24a and 24b are made in the wall of the top basket receptacle area 16. As shown in FIG. 1E, they are 180 degrees opposed to each other. They are configured to cooperate with a pair of protrusions which appear on a top basket (discussed below with respect to FIG. 3) such that the protrusions fit into notches 24a and 24b. This configuration insures proper placement of the top basket into the top basket receptacle area 16 so that the top basket is straight and at the correct depth. In prior art column assemblies, which lacked these notches and protrusions, it was possible to insert the top basket in such a manner that it was not straight and/or at the wrong depth, which adversely affected the function of the column assembly. FIG. 1E shows two notches 24a and 24b 180° opposed to each other. It is understood that the present invention is not limited to this configuration. Rather, there can be 1, 3, 4, 5, 6 or more notches present in the wall of the top basket receptacle area 16 in any configuration, so long as these notches cooperate with protrusions on the top basket to insure its proper fit. FIG. 1F shows a side view of the inventive column assembly, cut along line C-C of FIG. 1B. FIG. 1G is detail D of FIG. 1E, showing an alternative embodiment for the first top portion 7a. As shown in FIG. 1G, this first top portion 7a slopes downwardly from its top, whereas the first top portion 7 of FIG. 1F is squared off, i.e., non-sloping. FIG. 2 shows an alternative embodiment of the inventive column assembly. As shown in FIG. 2D, which is detail B from FIG. 2C at a scale of 3:1, the bottom stopper 11a is configured to fit into substantially all of the space of the bottom stopper receptacle area 21. This insures a better fit between the outer wall of the bottom stopper 11a and the inner wall of the bottom stopper receptacle area 21, thus further insuring against any leaks. In addition, the stopper 11a reduces the dead volume in the bottom stopper receptacle area 21. Minimization of the dead volume minimizes mixing with non-radioactive or decayed eluent, which could dilute the fresh eluent, thereby maintaining a narrow rubidium-82 bolus profile. The bottom stopper 11a further comprises a bottom stopper hollow space 11b. This bottom stopper hollow space 11b helps prevent the bottom stopper 11a from blocking the outflow arm 12. The column assembly is preferably made of polypropylene. Prior art column assemblies were made with H5820 polypropylene. While that product can still be used, in a preferred embodiment the polyproplylene random copolymers PP P5M4R-034 or PP 13R9A (Huntsman Polymers (The Woodlands, Tex.)) can be used because they are more resistant to radiation than the prior art H5820 polypropylene. See the Prospector X5 data sheets with ATSM and ISO properties for PP P5M4R-034 and PP 13R9A, which are incorporated herein by reference in their entirety. Of the two Huntsman polypropylenes, PP 13R9A is the more preferred, based upon UV profile, Instron stress testing and appearance after gamma-irradiation. The manufacturing process for the inventive column assembly has also been improved. A new automatic mold has been designed which increases the quality and appearance of the column assembly, and which increases the efficiency of the manufacturing process. Manufacturing is presently done by Duerr Molding (Union, N.J.). For example, pins are used to form the hollow portions of the inlet arm 22 and outflow arm 23. In the prior art molding process, these pins were not fixed, so they floated. As a result, the side wall thickness of the inlet arm 1 and outlet arm 12 varied. In the present process, the pins are fixed. Therefore, the thickness of the side walls is more uniform. Also, as described above, the position of the outflow arm 12 has been moved, the outflow arm no longer protrudes into the hollow interior or lumen of the column 3, and the outflow arm resides in a recess or notch. This prevents the outflow arm from being blocked. Furthermore, support means 4, 14 are provided to strengthen the inlet arm 1 and the outflow arm 12. In addition, notches 24a and 24b are provided for the proper placement of the top basket. In the inventive column assembly shown in, for example, FIG. 1A and FIG. 2A. The inlet arm 1 and the outlet arm 12 are straight. That is because this is the way the column assembly looks at the end of the molding process. In use, the inlet arm 1 and the outlet arm 12 are curved upward, in much the same configuration as the prior art CardioGen® generator is used. Further improvement to the manufacturing process and column assembly are described throughout the instant specification. The packing material area 18 of the column 3 is designed to receive packing material. The type of packing material used depends upon the intended use of the column arrangement. When used as, for example, a rubidium-82 generator, such as CardioGen®, the packing material is one which will adhere strontium-82 but will allow for the elution of rubidium-82. Strontium(II)-82 decays into rubidium(I)-82. Elution of strontium-82 is not desired because it binds to bone and exposes the patient to unnecessary radiation exposure. Presently, stannic oxide is he preferred packing material. The packing material is loaded into the column 3 in a conventional manner. The column 3 is then loaded with strontium-82 in a conventional manner. A liquid containing the strontium-82 is slowly added to the top of the packed column and allowed to flow through it by the force of gravity. If necessary, a small vacuum can be used. Also, the packing material is preferably wetted before the strontium-82 is added. Slow addition of the strontium-82 is preferred because it will result in the strontium-82 being absorbed as close to the top of the column as possible. Filters, preferably fiberglass filters, can also be used in this conventional loading procedure. For example, two fiberglass filters are first placed in the column 3, then a portion of the packing material is added, followed by a single fiberglass filter, then the remainder of the packing material, then two more fiberglass filters. Once filled, the top basket or spacer 17 is inserted into the top basket receptacle area 16. The top basket 17 acts as a retainer to hold the packing material in place. FIG. 3 shows schematics of the spacer or top basket 26 of the inventive column assembly. The spacer or top basket 26 is cylindrical in shape with an open top portion 27 and a screen 28 at the bottom portion 29. Another top basket or spacer 17 of similar configuration is shown in FIG. 1, placed in the top basket receptacle area 16. As shown in the embodiment of FIGS. 3B and 3D, the top basket 26 actually has three cylindrical areas, a top cylindrical area 30, a middle cylindrical area 31 and a lower cylindrical area 32. The top 30 and bottom 32 cylindrical areas have diameters about equal to each other, and their diameters are greater than the diameter of the middle cylindrical area 31. The top basket 26 also contains protrusions 33a, 33b which are designed to cooperate with notches 24a, 24b in the top basket receptacle area 16. In operation, the protrusions 33a, 33b fit into the notches 24a, 24b to insure proper alignment of the top basket 26 in the top basket receptacle area 16. When so positioned, the top basket 26 acts as a retainer to hold the packing material in place. As shown in FIGS. 3A and 3C, the two protrusions 33a, 33b are 180° opposed to each other. They are located at the top cylindrical area 30. As was the case with the notches 24a, 24b, the present invention is not limited to this configuration. Rather, there can be 1, 3, 4, 5, 6 or more protrusions, in any orientation, so long as they cooperate with the notches to help insure a proper fit for the top basket 26. The top basket 26 also contains a side opening 34. As shown in FIGS. 3B and 3D, the side opening is in the middle cylindrical area 31 of the top basket 26. The purpose of the side opening is to line up with the inlet arm 1 when the top basket 26 is placed in the top basket receptacle area 16. In this arrangement, when a liquid is introduced into the inlet arm 1, it will pass through the side opening 34 into the top basket 26. The top basket 26 can be made of any suitable material, such as polypropylene. Preferably, the material will be radiation resistant, i.e. resistant to degradation in the presence of a radioactive material. More preferably, the top basket 26 is made of the same material used to construct the column assembly. In a preferred embodiment, that material is PP P5M4-R-034 or PP 13R9A polypropylene (Huntsman Polymers (The Woodlands, Tex.). Even more preferably, the material is the PP 13R9A polypropylene. In a yet further preferred embodiment, the top basket 26 is molded at the same time the rest of the column assembly is molded. As discussed above, FIG. 4 shows a detailed view of the bottom 6 portion of the column 3. FIG. 4 shows the outflow notch 25 where the hollow portion 23 of the outflow arm 12 intersects the hollow interior of the column 3. The outlet notch 25 prevents blockage of the hollow portion 23 of the outflow arm 12 by the bottom stopper 11 (not shown in FIG. 4). FIG. 5 shows various types of crimp seals to use with the present invention. FIG. 5A shows the current, prior art crimp seal. FIGS. 5B-5F show various alternate embodiments of the crimp seal. The function of the crimp seal is to form a tight, crimped seal between the stoppers (described below) and the pharmaceutical container to prevent leakage. Also, a central hole is provided in the crimp seal to allow for the insertion of a needle or similar device. In one preferred embodiment the pharmaceutical container is a column, or column assembly, such as one used in a rubidium generator. The crimp seal can be made of any material, such as plastic or metal. The material should preferably be radiation resistant, and of sufficient strength to withstand pressures of at least 90 psi and preferably up to 160 psi. More preferably, the material should be metal. Preferred metals comprise aluminum, steel and tin, or suitable alloys or mixtures thereof. The metal can be optionally coated. For example, tin coated steel can be used. The diameter of the crimp seal will vary according to use, for example, vary according to the diameter of the pharmaceutical container which is to be crimped. With respect to a column assembly to be used as a rubidium-82 generator, such as CardioGen®, the diameter of the crimp seal is preferably about 20 mm across its top. FIG. 5A shows a conventional prior art crimp seal 35. It is made out of aluminum which is about 0.20 mm thick, has a flat top portion 36 with a diameter of about 20 mm with central hole 37 of about 9.5 mm in diameter and a skirt 38 about 7.5 mm high. There are several potential problems with this prior art crimp seal. First, because aluminum with a thickness of only 0.20 mm is used, the crimp seal might not be strong enough to insure a strong, leakproof seal. Second, the central hole 37 is large, and therefore the stopper might not be properly supported. Also, the larger central hole 37 may allow for ballooning of the stopper. Third, this crimp seal is manually crimped to the column 3. Manual crimping can result in undesirable variability of crimping pressure and, accordingly, can affect how well the crimp seal 35 seals the column 3 to prevent leakage. FIG. 5B shows one type of useful crimp seal 39. This crimp seal 39 comprises two parts, a top crimp member 40 and a bottom washer 41. Both the top crimp member 40 and the bottom washer 41 are made of aluminum (vendor—West). The thickness of the aluminum for each part can vary depending upon the intended use, but the aluminum used for each member is generally about 0.20 mm thick. The top crimp member 40 has a central hole 42 and a skirt 43. The size of each, and the diameter of the crimp seal, can vary depending upon use. As shown in FIG. 5B, the central hole 42 has a diameter of about 6.4 mm and the skirt 43 is about 7.6 mm high. The diameter of the top crimp member 40 is about 20 mm. The top crimp member 40 also has a cover 44, which covers the central hole 42 when not in use but can be pulled or pealed back when in use. Also, while none of FIGS. 5C through 5F or FIG. 6 show a cover, it is understood that each of these embodiments can employ a cover if desired. FIG. 5B also employs a bottom washer 41. The bottom washer 41 contains a central hole 45. The bottom washer central hole 45 can have a diameter greater than, the same as or smaller than the diameter of the central hole 42 in the top crimp member 40. As shown in FIG. 5B, both central holes 45, 42 have about the same diameter, i.e. about 6.4 mm. The bottom washer 41 does not have a skirt. The diameter of the bottom washer 41 is about 20 mm. When used, the bottom washer 41 is placed below the top crimp member 40 and both are crimped into place. Crimping is preferably performed via an automatic or semi-automatic crimper, which is discussed in more detail below. In the alternative, other processes which control the crimping pressure applied can be used. FIG. 5C shows another embodiment of the inventive crimp seals. This crimp seal 46 comprises a single member. It is made out of steel (vendor—Microliter). The thickness of the steel can vary according to the intended use, but is generally about 0.20 mm thick. This crimp seal 46 is about 20 mm in diameter, contains a central hole 47 of about 5.0 mm in diameter and has a skirt 48 about 7.2 mm high. The crimp seal 46 is preferably crimped into place using an automatic or semi-automatic crimper, although other processes which control the pressure applied can be used. FIG. 5D shows yet another embodiment of the inventive crimp seals. This crimp seal 49 comprises a single member. It is made out of steel (vendor—Microliter). The thickness of the steel can vary according to the intended use, but is generally about 0.20 mm thick. This crimp seal 49 has a diameter of about 20 mm, contains a central hole 50 of about 8.0 mm in diameter and a skirt 51 about 7.2 mm high. The crimp seal 49 is preferably crimped into place using a semi-automatic crimper, although other processes which control the pressure applied can be used. FIG. 5E is yet still another embodiment of the inventive crimp seals. This embodiment comprises two parts, a top crimp member 52 and a bottom washer 53. Both the top crimp member 52 and the bottom washer 53 are made of aluminum (vendor—Microliter). The thickness of the aluminum can vary depending upon the intended use, but the aluminum used for each member is generally about 0.20 mm thick. The top crimp member 52 has a central hole 54 and a skirt 55. The central hole 54 has a diameter of about 9.6 mm and the skirt 55 is about 7.6 mm high. The top crimp member 52 has a diameter of about 20 mm. The top crimp member 52 also contains an insert 56, which is seated in or under the central hole 54. The insert 56 can be made of any suitable substance, but is preferable made of metal, such as steel, aluminum or tin, or plastic. The insert 56 also contains an insert central hole 57, which has a diameter of about 5 mm. The bottom washer 53 also has a central hole 58, which has a diameter of about 5 mm. The bottom washer 53 is about 20 mm in diameter and it does not have a skirt. When used, the bottom washer 53 is placed below the top crimp member 52 and the insert 56 and then all are crimped into place. Crimping is preferably performed using an automatic or semi-automatic crimper, although other processes which control the pressure applied can be used. FIG. 5F shows yet another embodiment of the inventive crimp seals. Like FIG. 5E, FIG. 5F employs two members, a top crimp member 59 and a bottom washer 60. Both members are made of aluminum (vendor-Microliter). While the thickness of the aluminum can vary with the intended use, generally each member is about 0.20 mm thick. The top crimp member 59 contains a central hole 61 and a skirt 62. The central hole 61 has a diameter of about 9.6 mm and the skirt 62 is about 7.6 mm high. The top crimp member 59 has a diameter of about 20 mm. The bottom washer 60 also has a central hole 63. The bottom washer central hole 63 has a diameter of about 11.4 mm. The diameter of the entire bottom washer 60 is about 20 mm. The bottom washer 60 does not have a skirt. When used, the bottom washer 60 is placed below the top crimp member 59. Both are then crimped into place. Preferably, an automatic crimper is employed, although other processes which control the pressure applied can be used. FIG. 6 is an alternate and preferred embodiment of the inventive crimp seals. This crimp seal 64 comprises a single member. It is made out of steel (vendor—Microliter), code #20-000 M. See the Microliter Product Catalog, which is incorporated herein by reference in its entirety. The thickness of the steel is about 0.20 mm. The crimp seal 64 contains a central hole 65 and a skirt 66. The central hole 65 is about 5.00 mm±0.25 mm in diameter and the skirt 66 is about 7.00 mm±0.25 mm high. The entire crimp seal 64 has a diameter of about 20.75 mm±0.25 mm. The crimp seal 64 is preferably crimped into place using an automatic or semi-automatic crimper. FIG. 7 shows an improved stopper 67 to be used with the inventive column assembly. The stopper 67 is preferably made from a material which will form a tight seal with the column assembly. In a preferred embodiment the stopper 67 is made of a material which is also resistant to radiation. Prior art stoppers were made of materials such as Itran-Tompkins PT-29 green neoprene rubber. This material had two potential disadvantages. First, it could degrade when exposed to radiation. Second, it contained latex, which could cause allergic reactions. Various materials were compared to the PT-29 green neoprene used in the prior art. These materials included neoprene, isoprene, bromobutyl, chlorobutyl, nitrile, isoprene/chlorobutyl, EPDM (ethylene propylene diene monomer) and Viton. These materials were coated, uncoated, siliconized and non-siliconized. These materials were made into column assembly stoppers and were irradiated simulating the exposure from a 100mCi generator over a time period of 45 days (about 145 kGy). Irradiated stoppers were compared to non-irradiated controls by integrity (pressure) testing of the column/stopper assemblies. Assemblies were pressurized to determine load pressure required to cause ballooning of rubber materials or leaks/burst at the seal closure (up to about 200 psi). In addition, for the purpose of determining potential rubber extractables and/or leechables, additional column/stopper assemblies were irradiated in the presence of 0.9% saline solution. The saline solution was then scanned at 250 mm for UV absorbing extractables. Three compositions were identified as suitable to use in stoppers: West Pharmaceutical Services (Lionville, Pa.) 4588/40 isoprene/chlorobutyl; American Stelmi (Princeton, N.J.) 6720 bromobutyl; and Helvoet-Pharma (Pennsauken, N.J.) Helvoet FM 140/0 chlorobutyl. Of these materials, the most preferred product to use is the West 4588/40 isoprene/chlorobutyl. The stopper 67 should be configured so that it forms a tight seal with the column assembly and minimizes the dead volume (mixing), thus maintaining a narrow rubidium-82 bolus profile and maximizing efficiency. One preferred structure for the stopper is shown in FIG. 7. Referring to FIG. 7B, the stopper 67 comprises a generally cylindrical top section 68 and a generally cylindrical bottom section 69. The diameter of the stopper bottom section 69 is about the same as or slightly larger than the inside diameter of the first top portion 7 and first bottom portion 9 of the cylinder 3, assuming both of these portions 7, 9 have the same diameter. If these portions have different diameters, then the cylindrical bottom section 69 of the stopper 67 will have about the same or slightly larger inside diameter as the portion 7, 9 it is intended to be inserted into. The reason for this configuration is to insure a tight fit between the stopper 67 and the first top 7 and first bottom 9 portions of the cylinder 3. A tight cylinder 3/stopper 67 interface helps prevent leakage. The stopper top section 68 has a greater diameter than the stopper bottom section 69 to prevent the stopper 67 from being inserted too far into the cylinder 3. In addition, optionally the stopper top section 68 can have a curved upper edge 70. The stopper bottom section 69, in one preferred embodiment, contains a U-shaped groove 71 in its base. See FIG. 7A. The U-shaped groove 71 traverses greater than half the length of the stopper bottom section 69, and it terminates in a semi-circular section 72. Preferably, the center point 73 of the semicircular section 72 should be about at the center point of the stopper bottom section 69. The stopper top section 68 contains a central circular indentation 74 in its top surface. See FIG. 7C. Preferably, the diameter of the central circular indentation 74 has a diameter about equal to the width of the U-shape groove 71. As shown in FIGS. 7B and 7D, the central circular indentation 74 and the U-shaped groove 71 should preferably line up with each other when the stopper is viewed through its cross-section. The central circular indentation 74 and U-shaped groove 71 allow for easy insertion of a needle or similar device into the stopper 67. The surface of the stopper top section 68 also contains three spherical dots 75a, 75b, 75c and an indicia, such as a spherical lug 76. They are spaced equidistant from each other around the central circular indentation 74. Also, the spherical lug 76 is placed so that it is above the U-shaped grove 71. In this configuration, when the stopper 67 is inserted into the first top portion 7 of the column 3, the spherical lug 76 can be lined up with the inlet arm 1. Thus, the open end of the U-shaped groove 71 will face the inlet arm 1, thus preventing its blockage. The same holds true for the first bottom portion 9 of the column 3. When the stopper 67 (stopper 11 shown in FIG. 1 and stopper 11b in FIG. 2 can have the same or different configurations from stopper 67) is inserted therein, the spherical lug 76 is lined up with the outlet arm 12. The open end of the U-shaped groove 71 will then face the outlet arm 12 and prevent its blockage. It is understood that the present invention is not limited to a U-shaped groove 71. Any other configuration, such as a notch, can be used so long as any potential blockage is avoided. In fact, if there is no potential for blockage, the U-shaped groove 71 or alternative structure can be eliminated. The stopper 67 is affixed to the column 3 via crimping, using the crimping seals described above in FIGS. 5 and 6. In the prior art, crimping was performed manually. The disadvantage of manual crimping is that it is not always uniform. One problem this can cause is leakage. To overcome this potential problem, the present invention preferably uses automatic or semi-automatic crimping. Any automatic or semi-automatic crimper can be used for the present invention, so long as it can consistently crimp seals at a specified, controlled pressure. One preferred type of automatic crimper is a pneumatic crimper, which is powered by gas. One example of a pneumatic crimper suitable for the present invention as an AP/CP2000 Lightweight Air Crimper/Decapper (Laboratory Precision Limited, UK). See Laboratory Precision Limited brochure copyrighted Apr. 4, 2001, which is incorporated herein by reference in its entirety. In the crimping process, a stopper 67 is inserted into the top portion 5 or bottom portion 6 of the column 3, so that it is seated in the first top portion 7 or first bottom portion 9, respectively. A crimp seal or a crimp seal and washer (see FIGS. 5 and 6) is/are placed over the stopper 67. The crimp seal or crimp seal and washer are then crimped into place, either manually or, preferably, automatically or semi-automatically. While the crimping pressure used is optimized based upon the configuration and material of the crimp seal and stopper, generally about 117±3 psi pressure is used. The resulting crimped crimp seal/stopper configuration can withstand the operative pressures of the system, i.e. at least 90 psi and preferably up to 200 psi. When in operation, connector tubes (not shown) are connected to the column assembly. Referring to FIG. 1A, both the inlet arm 1 and the outlet arm 12 have a female Luer cap 2, 13 at their distal ends. These female Luer caps 2, 13 engage male Luer caps at the proximal ends of the connector tubes. Prior art connector tubes can discolor from clear to brown and harden upon prolonged exposure to radiation. Also, the Luer connector can discolor and become brittle. In addition, the Luer connectors can loosen or become unintentionally disconnected during use. Accordingly, the present invention includes constructing connector tubing out of radiation resistant materials. Preferably, the tubing is made from a flexible radiation resistant polyvinyl chloride (PVC) and the Luer connector is made from a rigid radiation resistant PVC. For example, a preferred material for constructing the tubing is AlphaGary PVC 2232 A/R-78S Clear 030X. See AlphaGary Test Result Certificate, Report Date Aug. 20, 1999; Technical Data, Date of Origin 8/99; and Material Safety Data Sheet printed Apr. 5, 2000; which are incorporated herein by reference in their entirety. A preferred material for constructing the Luer connector is AlphaGary PVC 2212 RHT/1-118 Clear 080×. See AlphaGary Data Sheet, Revision Date 4/02, which is incorporated herein by reference in its entirety. Also, using this AlphaGary rigid PVC for the Luer connector allows the heat bonding of tubing to the Luer connector. In an alternative embodiment of the present invention, the distal end of the connector tube attached to the outlet arm 12 of the column assembly as shown in FIG. 1A has a check valve (not shown) attached to it. In a preferred embodiment, the check valve is included in the patient tube 103, shown in FIG. 9, either before or after the patient sterilization filter 104. The check valve prevents a back flow of fluids from entering the connector tube when connected to or disconnected from a patient. In another alternative embodiment, sometimes the generator is placed so far away from a patient that the patient tube cannot reach all the way to the patient. In this instance, one or more extension tubes can be added, the length of which is sufficient to reach the patient. Preferably, a single extension tube is used and in a preferred embodiment, it is made of the same materials as the connector tubes discussed above to provide for, e.g., flexibility and radiation resistance. The present invention further includes an improved Luer lock. The improvements are described below. An embodiment of this improved Luer lock is set forth in FIG. 8. These improved Luer locks can be used with the pharmaceutical containers of the present invention, or in any other indication where it is desirable to have a connection that will not loosen or inadvertently disconnect. In the embodiment of FIG. 8, FIG. 8A show a side view of the inventive column assembly with the inlet arm 1 projecting forward. Also shown is the female Luer cap 2 at the distal end of the inlet arm 1. As shown in FIG. 8C, the female Luer cap 2 terminates in a flange 77. The flange 77 can be flat or, as shown, contain a groove 78. Other configurations, known in the art, can also be used. The flange 77 is configured to engage and mate with threads 78 in a male Luer cap 79. When the two caps 2, 79 are screwed together, they form a tight Luer lock which will be leak resistant. This configuration is shown in FIG. 8D. One difficulty with a Luer lock is to know when the male and female caps 79, 2 have been connected sufficiently to form a tight lock. To overcome this problem, one or more tabs are provided on each of the male 79 and female Luer caps 2. As shown for example in FIGS. 8C and 8D, two tabs are provided on each cap 80a, 80b, 81a and 81b, although it is understood that the invention is not limited to this configuration only. For example, each of the Luer caps can also contain 1, 3, 4, 5, 6 or more tabs. In one embodiment, the female Luer cap tabs 80a, 80b and the male Luer cap tabs 81a, 81b are so positioned that when the Luer locks is sufficiently tight, the tabs line up with each other. This way, a user knows when tightening is completed. The present invention, however, is not limited to this one configuration, so long as the tab or tabs on each of the Luer connectors 79, 2 are arranged in a desired configuration to demonstrate that the Luer connectors 79, 2 are sufficiently tightened. In another preferred embodiment, as shown in FIG. 8D, the male Luer cap tabs 81a, 81b overlap with the female Luer cap tabs 80a, 80b. The tabs are so positioned that this overlap occurs when the tightening is complete. At the point of desired tightening, the tabs 80a, 80b, 81a, 81b pass by or click past each other. That way, the Luer locks cannot be over- or under-tightened. Also, loosening or disconnection of the Luer lock during use is prevented by the overlapping of the tabs, preventing the Luer connectors 79, 2 from turning in a loosening direction. Although the inventive Luer locks are shown only as part of the generator as shown in FIGS. 8A and 8B, the inventive Luer locks can be used in place of conventional Luer locks at any place in the inventive generator system. Moreover, the inventive generator system can contain a combination of conventional Luer locks and the inventive Luer locks. Finally, the inventive Luer locks are not solely intended for use with the inventive generator system. Rather, they can be used in place of conventional Luer locks wherever those conventional Luer locks are used. When the inventive column assembly is used as, for example, a rubidium-82 generator, it is pre-packaged with strontium-82 in the factory. That is, the product shipped to the customer is radioactive. Therefore, the radioactive column assembly is shipped in a shielded (e.g. lead) container. Nevertheless, leakage is still a concern upon shipping. Thus, to improve safety when the radioactive column assembly is shipped, an inventive improvement is to ship the product with a liquid absorbent pad. Preferably, the shipping pad is a GP100 absorbent pad (Shell Packaging Corporation, Springfield, N.J.). GP100 is a 100% polypropylene non-woven mat of randomly oriented micro-fibers (2-10 micron diameters). See SPC General Product Specifications for GP100 dated May 26, 2003, which is incorporated herein by reference in its entirety. This type of shipping pad is useful in absorbing any leaks which may occur. Improved Seal The new seal, which is used to crimp the rubber stopper in place in a pharmaceutical container and particularly, which is used to seal a radioisotope generator column/stopper assembly system, such as CardioGen®, is preferably made of a sufficiently strong material to eliminate the problems discussed above. FIGS. 5B through 5F and FIG. 6 illustrate various method of reinforcing the top portion of the seal by use of a second layer (washer) or use of a stronger material such as steel/tin in addition to reducing the size of the center hole. The material may include metal or plastic, but is preferably metal. The metal may include heavy gauge aluminum, steel or tin, but is preferably steel or tin. The seal generally has the configuration shown in FIGS. 5B through 5F and FIG. 6 and may have a small or large central hole, a shorter or longer skirt and optionally, a cover (e.g., plastic or aluminum over the central hole). The dimensions of the seal will vary, and one skilled in the art will understand that they should be appropriate to the container which is being sealed. Approximate dimensions for seals for a radioisotope generator column are shown in the various examples in FIG. 5 and in FIG. 6. These dimensions are approximate and are not intended to be limiting. The central hole of the seals of the invention may vary in size. In a preferred embodiment the seal has a smaller central hole such as, for example, those proportional to the central holes shown in FIG. 5B, FIG. 5C, FIG. 5E and FIG. 6. In one embodiment, seals of FIG. 5B through FIG. 5F and FIG. 6 are used to seal a radioisotope generator column. These seals are available from the vendors West Pharmaceutical Services (Lionville, Pa.) and Microliter Analytical Supplies Inc. (Suwannee, Ga.). In a particularly preferred embodiment, the central hole of the seal is reduced in size such as in the seals in FIG. 5B, FIG. 5C, FIG. 5E and FIG. 6. The preferred configuration for this application is a 1-piece steel/tin crimp with a center hole of approximately 4-5 mm diameter and a skirt length of approximately 7.2 to 7.5 mm as shown in FIG. 6. The combination of using a stronger material such as steel/tin or heavier gauge aluminum and reduction of the center hole results in optimum performance in maintaining a secure leakage free seal under high pressure and particularly repeated exposure (pulsing or cycling) to high pressure as occurs with the use of the rubidium-82 generator as the enlarged surface area of the crimp limits excessive expansion of the rubber closure under pressure. The use of a stronger material such as steel/tin or heavy gauge aluminum further improves the performance of the crimp by reducing the likelihood of failure due to relaxation or fatigue of the seal flange which is formed at the point where the crimp skirt is folded under the column or container flange when exposed to high or pulsating pressures. It is understood that the skirt length can be varied to provide a proper fit with the container/rubber seal combination to which it is applied. Automatic Crimper and Improved Crimping Process In a preferred embodiment, an automatic or semi-automatic crimper is used to crimp the seals of the invention. The automatic or semi-automatic crimper is set at an optimized pressure and is able to crimp seals of any material during assembly of a pharmaceutical container such as a radioisotope generator column/stopper assembly system. Suitable automatic crimpers include pressurized and/or compressed air crimpers such as those available from Laboratory Precision Limited under the trade name/model number AP/CP2000. Use of the automatic or semi-automatic crimping procedure of the invention with compressed or pressurized air results in consistent/reproducible crimping pressures, and enables selection of optimized crimping pressures when crimping various seal materials. Use of optimized pressures improves the performance of the seals of the invention and also improves performance of seals of only moderate strength, such as lighter gauge aluminum and some plastics. The automatic or semi-automatic, pneumatically powered crimper used to apply the seal is preferably operated at an optimized pressure of between 60-140 psi. However, although automatic or semi-automatic crimpers are preferred, it should be noted that application of the seal is not limited to automated equipment, and systems ranging from manual to fully automatic may be used, provided their operation can be optimized to produce repeatable and consistent predetermined pressures in applying the seals. Column Design Improvements Manufacturing Process: To create the new column design, a new automatic mold has been designed. The mold and the new columns produced therein exhibit improved column quality and appearance. The new mold also increases the efficiency of the manufacturing process. The increased speed of the new automated mold enables one operator to run the process efficiently. Column Design: The improved pharmaceutical container also includes improvements to the design which ensure specified flow of eluent through the container and improve its packing and consistency. In one embodiment the improved container comprises a column used in a radioisotope generator. The improved column includes a repositioned outlet arm, and the column outlet resides in a recess or notch in the inside ledge of the column where the outlet arm enters the column lumen, to prevent a stopper from blocking the flow. These improvements further include introducing small reinforcement pieces of resin to the outside of the column between the outlet arm and column body and between the inlet arm and column body to provide additional strength. Additionally, the seam of the inlet and outlet arms has been eliminated by changing the mold runners. This change has improved the consistency of the inlet and outlet arm diameters and made the arms stronger. Furthermore, to address consistency of packing of the containers, two small alignment slots have been cut into the wall of the column to receive the orientation knobs on the baskets that properly align and seat the basket in the column and limit the insertion depth into the column. This improves the consistency of packing density and eliminates potential blockage of the inlet arm. Additionally, in one embodiment, the improved column has stopper flanges and Luer flanges with much smoother surfaces with sharper edges to improve the sealing ability of the crimp. These attributes improve stopper and Luer contact to the column and greatly reduce the chance of leakage. Also, the flashing on the column is reduced greatly to enhance the appearance of the part. Finally, the column assembly is made from a radiation resistant or tolerant material. The most preferred material is Huntsman PP 13R9A polypropylene. Luer Lock and Connector Tube Improvements The Luer locks and connector tubes used with the column have also been improved. First, the connector tubes are made from a radiation resistant or tolerant material. Preferably, this material is AlphaGary PVC 2232 A/R-78S clear 030X. Second, the terminal end of the connector tube which attaches to the column contains a male Luer cap. This male Luer cap is made of a radiation resistant material, preferably AlphaGary PVC 2212RHT/1-118 clear 080X. Third, the male and female Luer caps screw together and each contains tabs, preferably two tabs each. When the tabs line up with each other in one embodiment or overlap with each other in another embodiment, that indicates that the two Luer caps are sufficiently tightened or screwed together to form a tight seal or lock. Also, in a preferred embodiment the overlapping tabs prevent the Luer caps from becoming loose, ie unscrewing. FIG. 9 is a diagram of the entire radionucleotide generator system. In this system, a saline supply 83 is connected to a saline supply tube 84. The saline tube 84 passes through a first check valve 85 and a second check valve 86. The check valves 85, 86 are used to insure that the saline solution only flows in the direction of the rubidium generator column 3. Interspersed between the check valves 85, 86 is a syringe pump 87. The syringe pump 87 connects to saline supply tube 84 at a T-junction 88 via a syringe pump luer connection 89. After the second check valve 86, a pressure transducer 90 is connected to the saline supply tube 84 via a pressure transducer luer connection 91. The saline supply tube 84 terminates at a first sterilization filter 92 and is connected to it via a first sterilization filter luer connection 93. The sterilization filter 92 is connected to a column connector tube 94 via a column connection tube luer connector 93. The column connector tube 94 passes through a generator shield 95 and connects to the female luer cap 2 of the inlet arm 1 via a male luer cap as shown in FIG. 8D. The generator shield 95 prevents exposure to radiation from the column 3 which can contain radioactive materials, such as strontium and rubidium-82. The inlet arm 1 is connected to the column 3 which is connected to the outlet arm 12 as shown in, for example, FIGS. 1 and 2. The female luer cap 13 of the outflow arm 12 connects to the male luer cap (not shown) of outlet connecting tube 96. The outflow connecting tube 96 passes through the generator shield 95 and connects via an outflow connecting tube luer connector 97 to a divergence valve tube 98. The divergence valve tube 98 passes through a positron (beta) detector 99, which is used to insure that the liquid to be injected into a patient has the correct level of radioactivity. Recall that at this point the liquid, which is usually a saline solution and starts at the saline supply 83, has now passed through the column 3 and thus, will contain rubidium-82. After the positron (beta) detector 99, the divergence valve tube 98 passes to a divergence valve 100. The divergence valve 100 will divert the liquid to either the diversion outlet tubing 101 or a waste connection tube 102. The diversion outlet tubing 101 connects via the patient tube-luer connection 102 to a patient tube 103, which terminates at a patient sterilization filter 104 which is solvent bonded at the time of manufacture to the patient tube 103. A needle may be attached to the patient sterilization filter 104. The patient tube 103 can pass directly to a patient (via the patient sterilization filter 104). In an alternative embodiment, the patient tube 103 can include a check valve prior to the patient sterilization filter 104. The check valve may be solvent bonded at the time of manufacture of the assembly (not shown). The check valve can be connected to the patient tube 103 by a check valve luer connection (not shown) which may be solvent bonded at the time of manufacture of the patient line. In yet another alternative embodiment, the check valve can be connected after the patient sterilization filter 104, optionally via a luer connection. Also, as described above, if the distance to the patient is too great, one or more additional connector tubes (also called extension tubes) (not shown) can be added to the assembly to bridge the distance to the patient. For example, one or more extension tubes may be connected with a luer fitting between the patient tube luer connection 102 and the patient tubing 103. The waste connector tube 109 passes through a waste sterilization filter 105 to a waste bottle 106, and these can be connected to each other via a waste luer connection 107. The waste bottle 106 is surrounded by a waste shield 108 to prevent exposure to radiation. The system shown in FIG. 9 and discussed above contains a number of luer connections. Some or all of these luer connections can be the inventive luer connections described above. Conversely, some or all of the luer connections can be of the conventional type, or do not even have to be luer connections at all, but rather can be any type of connectors, and can be jointly referred to as “connecting means”. Preferably, some or all of the connecting means are of the inventive type while the remainder are conventional luer connections. In addition, the tubes and connecting means are preferably made of radiation resistant materials. Preferably, they are made of the materials discussed above. This is especially true of those tubes and connecting means which are exposed to radiation. Shipping Improvements The columns can be shipped pre-loaded with, for example, strontium-82. Therefore, the columns are shipped in sealed containers containing GP-100 absorbent material to absorb any leakage. The above description is to be taken as illustrative and not in the limiting sense. Many modifications can be made to the design without deviating from the scope thereof. |
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058898301 | summary | CROSS-REFERENCE TO RELATED APPLICATION This application is a Continuation of International Application Serial No. PCT/DE95/01823 published as WO96/20485 Jul. 4, 1996. BACKGROUND OF THE INVENTION FIELD OF THE INVENTION The invention relates to a cooling system having a cooling pipe for cooling a containment chamber that serves to receive core melt of a reactor core of a nuclear power plant. In order to provide safe operation, nuclear power plants have numerous diverse and redundant safety systems, including cooling systems, through the use of which operating conditions that deviate from normal operating conditions can be detected early and counteracted. As a result, such safety-critical states as reactor core meltdown are practically precluded. In order to control that kind of accident, which is considered hypothetical, German Published, Non-Prosecuted Patent Application DE 40 41 295 A1, corresponding to U.S. Pat. No. 5,343,506, describes a core retainer and a method for emergency cooling of a nuclear power plant. The core retainer has a catch basin, which is disposed immediately below the reactor pressure vessel that encloses the reactor core. Both the catch basin and the reactor pressure vessel are disposed inside a reactor cavern, which is a concrete structure. Cooling channels extend along the floor and the walls of the catch basin between the catch basin and the concrete structure and coolant water can be carried through the cooling channels. The cooling channels on the floor communicate with a water supply and discharge into a cooling pipe that protrudes in siphonlike fashion into the water supply. The siphonlike cooling pipe includes one part shaped as an inverted U. The apex of the U is located above an operative level of the water supply, and although the cooling pipe does dip into the water supply, in the vicinity of its apex it protrudes out of the water supply. As a result, as long as the level is at the operative level, no coolant water enters the cooling channels. It is not until the water supply is flooded to a level higher than the apex of the U that coolant water enters the cooling channels, resulting in cooling of the outside of the catch basin. Cooling of the interior of the catch basin is carried out through a flood pipe, which is passed from the water supply through the concrete structure into the catch basin. The flood pipe is closed in the catch basin by a meltable stopper that does not melt open until at a high ambient temperature, thus allowing coolant water to flow into the interior of the catch basin. Coolant water is present in the flood pipe even during normal operation of the nuclear power plant, and as a result the meltable stopper is continuously cooled. SUMMARY OF THE INVENTION It is accordingly an object of the invention to provide a cooling system for cooling a retention or containment chamber constructed for receiving a core melt, which overcomes the hereinafore-mentioned disadvantages of the heretofore-known devices of this general type and which initiates cooling of a catch basin by passive measures and therefore inherently safely. With the objects of the invention in view, there is also provided a cooling system for cooling a containment chamber for receiving core melt of a reactor core of a nuclear power plant, comprising a flooding container to be filled with coolant fluid; a cooling pipe leading from the flooding container to the containment chamber; and a passively opening closure element closing the cooling pipe in the flooding container and opening as a function of a level of the coolant fluid. Through the use of a closure element that opens as a function of the level of the coolant fluid, it is assured that cooling of the containment chamber will not ensue by feeding of coolant fluid into the flooding container until a safety-critical state exists. The coolant fluid that additionally flows into the flooding container is preferably primary coolant water, which emerges from the primary coolant loop of the reactor core during the safety-critical state. Coolant fluid can optionally be fed into the flooding container through a separate coolant fluid reservoir. The cooling pipe is closed until the closure element opens and therefore is free of water. As a result, during normal operation of the nuclear power plant, coolant fluid and particularly coolant water is kept away from the containment chamber, thus averting such problematic factors as corrosion from coolant water or unintended cooling of a temperature-dependent closure element that closes the cooling pipe. Moreover, the inherent safety of the nuclear power plant is improved by the passively opening closure element, and human error in initiating cooling of the containment chamber is precluded. In accordance with another feature of the invention, the closure element is a float that closes off the cooling pipe. At an operative level of the coolant water, this float has such buoyancy that it sealingly closes the cooling pipe, for instance through a ball seat. The float is preferably movable along a primary axis in a guide, so that unintended slippage of the float from its sealing seat is avoided even upon the occurrence of jarring of the kind that can be caused by earthquakes, for instance. In accordance with a further feature of the invention, the float has an interior that can be filled with coolant fluid. A filler pipe passing into this interior has an inlet opening for coolant fluid, through which the coolant fluid flows in if a flooding level occurs that rises above an operative level. The inlet opening may be located geodetically above the operative level or geodetically below this operative level. In the latter case, the filler pipe is extended from the inlet opening in a U above the operative level, so that an apex of the inverted U is located above the operative level. In this latter case as well, coolant fluid does not flow into the interior of the float until the operative level has been exceeded by a predeterminable amount. Coolant fluid flowing into the interior lessens the buoyancy of the float, so that beyond a certain fill level of the interior, the float leaves its sealing seat, thereby opening the cooling pipe. Cooling of the cooling pipe thus ensues in a passive way. In accordance with an added feature of the invention, the float has a condensed water suction removal device, by which condensed water that occurs can be removed by suction during a normal operative state of the nuclear power plant. As a result, lowering of the float from the occurrence of condensed water and an attendant unintentional initiation of cooling of the containment chamber are reliably avoided. In accordance with an additional feature of the invention, the containment chamber communicates with the flooding container through a return for coolant fluid that extends geodetically above the cooling pipe, and in particular above the operative level. This return is closed in the flooding container by a further closure element that opens as a function of the level. Internal cooling of the containment chamber by a coolant fluid loop is attained through the use of the return. Coolant fluid flowing from the flooding container to the containment chamber flows in natural circulation. This assures that during a safety-critical state of the nuclear power plant, sufficient coolant fluid is returned to the flooding container, and cooling of the containment chamber and in particular of the core melt received in the containment chamber occurs. In accordance with yet another feature of the invention, the closure element that closes the return to the flooding container and which may also be a float, has a ball valve. This ball valve may have a floatable ball, which is held in a sealing position by a guide path. The ball valve protects the closure element from a pressure wave which can occur, for instance, from a temperature increase inside the containment chamber. If coolant water flows out of the containment chamber into the return, the ball of the ball valve floats upward and thereby opens the return to the flooding container. In accordance with yet a further feature of the invention, the cooling pipe is a flood pipe, which discharges into the containment chamber and thereby assures direct cooling particularly of the surface of any core melt that has flowed into the containment chamber. The flood pipe preferably extends horizontally and can be both installed and removed by working from the flooding container. Installing the flood pipe in the containment chamber from the flooding container has the advantage of permitting the mounting to be provided outside the containment chamber, which may be poorly accessible and might be affected by radiation. This is especially favorable in the case of a containment chamber that surrounds the reactor core, since this installation can be carried out after the containment chamber is lined in the usual way with a crucible-like guard and collection layer. In accordance with yet an added feature of the invention, during normal operation of the nuclear power plant, the flood pipe is closed in the containment chamber with a closure element that opens as a function of temperature. During normal operation of the nuclear power plant, it is filled with air, and as a result the closure element that opens as a function of temperature is thermally insulated from the coolant water of the flooding container, and coolant water does not enter the flood pipe until during a safety-critical state of the nuclear power plant, so that the effects of corrosion are reliably avoided. As a result of the thermal insulation of the closure element that opens as a function of temperature, reliable opening, and in particular melting open, in the event of major heat development inside the containment chamber, are assured. The closure element that opens as a function of temperature can therefore be constructed in such a way that it opens the flood pipe only at high temperatures as compared with a closure element that is in direct contact with coolant water. The closure element that opens as a function of temperature is preferably resistant to neutron radiation, which occurs during normal operation of the nuclear power plant in the immediately vicinity of the reactor core and particularly in the reactor cavern that receives the reactor pressure vessel. Moreover, it has the advantage of using only a single melting element (melting screw, melting strip), so that canting and thus belated opening of the closure element as would occur if there were a plurality of elements melting at different times, is averted. The closure element is furthermore adapted to the cross section of the flood pipe, so that installation of the flood pipe with the closure element already assembled is assured. In accordance with yet an additional feature of the invention, the closure element that opens as a function of temperature has a material that melts open at a high temperature, for instance above 900.degree. C. This material may be corrosion-resistant and radiation-resistant and in particular may be silver. The closure element that opens as a function of temperature may have a bale closure with a silver tightening screw. The bale closure presses a cap sealingly into the flood pipe, so that this flood pipe is closed with certainty during normal operation of the nuclear power plant. In accordance with again another feature of the invention, the closure element that opens as a function of temperature can alternatively have a closure cap that is sealingly soldered to the flood pipe. Silver can also be used as the soldering substance. In accordance with again a further feature of the invention, the containment chamber has an external cooling device for externally cooling at least a floor and/or one wall of the containment chamber. The cooling pipe is a supply line connecting the external cooling device to the flooding container. During normal operation of the nuclear power plant, the supply line is closed by a float. The external cooling device preferably has a drain line for the coolant fluid, which returns to the flooding container. As a result, coolant fluid, in particular primary coolant water that has flowed into the flooding container, returns to the flooding container again, so that a coolant loop is provided for the external cooling of the containment chamber. In accordance with again an added feature of the invention, the containment chamber is a crucible-like catch basin disposed below the reactor core. Cooling of the catch basin, which ensues passively by a float disposed in the flooding container, takes place on the outside of the catch basin by the external cooling device and/or in the interior of the catch basin through the use of a flood pipe. Preferably, a flood pipe is extended thermally elastically from the flooding container to the catch basin, discharging into the latter. The flood pipe has a compensator outside the catch basin, in particular between the wall of the catch basin and a concrete structure that forms a reactor cavern. The compensator, which in particular is welded on and has a welded-on spherical flange, seals off the catch basin that has an interior with a temperature of approximately 300.degree. C., for instance, from the external cooling of the catch basin, which has a temperature of 20.degree. C. to 30.degree. C. The compensator serves to compensate for thermal expansions of the catch basin and additionally assures sealing off of the flood pipe from a coolant fluid flow for cooling the outer wall of the catch basin. In accordance with a concomitant feature of the invention, the cooling system is also suitable for cooling a propagation chamber located laterally below the reactor core. The interior of the propagation chamber may be cooled by a flood pipe, which extends from a flooding container into the propagation chamber. External cooling of the propagation chamber by suitably extended cooling channels, which are flooded with coolant fluid through a passively opening closure element, such as a float, inside the flooding container, is also possible. Other features which are considered as characteristic for the invention are set forth in the appended claims. Although the invention is illustrated and described herein as embodied in a cooling system for cooling a containment chamber constructed for receiving a core melt, it is nevertheless not intended to be limited to the details shown, since various modifications and structural changes may be made therein without departing from the spirit of the invention and within the scope and range of equivalents of the claims. The construction and method of operation of the invention, however, together with additional objects and advantages thereof will be best understood from the following description of specific embodiments when read in connection with the accompanying drawings. |
description | The present invention relates to a monitoring target selecting device, a monitoring target selecting method, and a program. Priority is claimed on Japanese Patent Application No. 2017-210663, filed Oct. 31, 2017, the content of which is incorporated herein by reference. In a nuclear power generation plant, an abnormality sign monitoring system acquires measurement information from a plurality of measurement devices provided in a control target, and acquires the acquired measurement information as plant operation data. In addition, the abnormality sign monitoring system detects an abnormality sign of the nuclear power generation plant on the basis of a correlation representing a mutual relationship between a plurality of acquired measurement parameters. Note that an abnormality diagnosis system includes a database, an abnormality sign monitoring system, and the like. The abnormality diagnosis system identifies an abnormal event and the like (abnormal event, facility with an abnormality sign, and abnormality countermeasure processing) of the nuclear power generation plant on the basis of an abnormality sign detection result and information of the database output from the abnormality sign monitoring system (for example, refer to Patent Document 1). [Patent Document] Japanese Unexamined Patent application, First Publication No. 2017-62730 However, in the technology described in Patent Document 1, since the correlation breaks down when the measurement information changes discontinuously, such as when the plurality of measurement devices have different operation cycles, it may be difficult to detect an abnormality sign in an operation cycle accordingly. Moreover, abnormality detection in an operation cycle to be monitored may become inefficient due to abnormality detection (erroneous detection) caused by different operation cycles in the technology described in Patent Document 1. The present invention provides a monitoring target selecting device, a monitoring target selecting method, and a program which can reduce erroneous detection and identify an abnormal event and the like effectively in an abnormality diagnosis system. According to a first aspect of the present invention, a monitoring target selecting device is a monitoring target selecting device which is configured to output a measurement parameter to an abnormality diagnosis device that is configured to diagnose an abnormal event of a plant based on a correlation value representing a mutual correlation between measurement parameters output from the monitoring target selecting device, and includes a classification unit configured to acquire a plurality of measurement parameters measured in the plant, classify a change behavior of a measured value over a time for each of the plurality of measurement parameters in a first period, and classify a change behavior of a measured value over a time for each of the plurality of measurement parameters in each of the first period and a second period, and a selection unit configured to select a measurement parameter as a measurement parameter to be output to the abnormality diagnosis device on the basis of a result of comparing a behavior of the measurement parameter in the first period and a behavior of the measurement parameter in the second period. In addition, according to a second aspect of the present invention, in the monitoring target selecting device, when the behavior of the measurement parameter in the first period is the same as a behavior of the measurement parameter in the second period, the classification unit, as a result of the comparison, may select this measurement parameter as the measurement parameter to be output to the abnormality diagnosis device, and, when the behavior of the measurement parameter in the first period is different from the behavior of the measurement parameter in the second period and the measurement parameter in the second period is normal, may select this measurement parameter as the measurement parameter to be output to the abnormality diagnosis device. In addition, according to a third aspect of the present invention, in the monitoring target selecting device, the classification unit may classify the measurement parameter, on the basis of an inclination value of a measured value over a time, into a first behavior whose inclination value is equal to or greater than a classification threshold value, a second behavior whose absolute value of the inclination value is less than the classification threshold value, and a third behavior whose absolute value of the inclination value is equal to or greater than the classification threshold value when the inclination value is a negative value. In addition, according to a fourth aspect of the present invention, in the monitoring target selecting device, when the classification results of the behavior of the measurement parameter in the first period and the behavior of the measurement parameter in the second period are the same, the classification unit may select the measurement parameter to be output to the abnormality diagnosis device. In addition, according to a fifth aspect of the present invention, in the monitoring target selecting device, when a ratio or difference between mean values of the measurement parameter in each period is within a range of a first predetermined value in case that the behavior of the measurement parameter in the first period and the behavior of the measurement parameter in the second period are the second behavior, the classification unit may select the measurement parameter to be output to the abnormality diagnosis device. In addition, according to a sixth aspect of the present invention, in the monitoring target selecting device, when a ratio or difference between respective inclination values is within a range of a second predetermined value in case that the behavior of the measurement parameter in the first period and the behavior of the measurement parameter in the second period are the first behavior, the classification unit may select the measurement parameter to be output to the abnormality diagnosis device. In addition, according to a seventh aspect of the present invention, in the monitoring target selecting device, when respective inclination values are negative values and a ratio or difference between respective inclination values is within a range of a second predetermined value in case that the behavior of the measurement parameter in the first period and the behavior of the measurement parameter in the second period are the third behavior, the classification unit may select the measurement parameter to be output to the abnormality diagnosis device. In addition, according to an eighth aspect of the present invention, the monitoring target selecting device further includes a storage unit configured to store the behavior of the measurement parameter in the first period and the behavior of the measurement parameter in the second period for each measurement parameter, wherein the selection unit refers to information stored in the storage unit, and selects the measurement parameter to be output to the abnormality diagnosis device when the behavior of the measurement parameter in the second period is normal: when the behavior of the measurement parameter in the first period is different from the behavior of the measurement parameter in the second period; when a ratio or difference between mean values of the measurement parameter in each period is outside the range of a first predetermined value in case that the behavior of the measurement parameter in the first period and the behavior of the measurement parameter in the second period are the second behavior; when a ratio between respective inclination values is outside a predetermined range in case that the behavior of the measurement parameter in the first period and the behavior of the measurement parameter in the second period are the first behavior; or when respective inclination values are negative values and a ratio or difference between respective inclination values is outside the range of a second predetermined value in case that the behavior of the measurement parameter in the first period and the behavior of the measurement parameter in the second period are the third behavior. In addition, according to a ninth aspect of the present invention, a monitoring target selecting method of a monitoring target selecting device which is configured to output a measurement parameter to an abnormality diagnosis device that is configured to diagnose an abnormal event of a plant based on a correlation value representing a mutual correlation of measurement parameters output by the monitoring target selection device includes a step of acquiring, by a classification unit, a plurality of measurement parameters measured in the plant and classifying a change behavior of a measured value over a time for each of the plurality of measurement parameters in a first period, a step of classifying, by the classification unit, a change behavior of a measured value over a time for each of the plurality of measurement parameters in each of the first period and a second period, and a step of selecting, by a selection unit, the measurement parameter to be output to the abnormality diagnosis device on the basis of a result of comparing a behavior of the measurement parameter in the first period and a behavior of the measurement parameter in the second period. In addition, according to a tenth aspect of the present invention, a program causes a computer of a monitoring target selecting device which is configured to output a measurement parameter to an abnormality diagnosis device that is configured to diagnose an abnormal event of a plant based on a correlation value representing a mutual correlation of measurement parameters output by the monitoring target selection device to execute a procedure of acquiring a plurality of measurement parameters measured in the plant and classifying a change behavior of a measured value over a time for each of the plurality of measurement parameters in a first period, a procedure of classifying a change behavior of a measured value over a time for each of the plurality of measurement parameters in each of the first period and a second period, and a procedure of selecting the measurement parameter to be output to the abnormality diagnosis device on the basis of a result of comparing a behavior of the measurement parameter in the first period and a behavior of the measurement parameter in the second period. According to at least one of the aspects described above, the monitoring target selecting device can reduce erroneous detection and identify an abnormal event and the like effectively in the abnormality diagnosis system. Hereinafter, embodiments of the present invention will be described with reference to the drawings. FIG. 1 is a diagram which shows a configuration example of an abnormality diagnosis system 1 according to the present embodiment. As shown in FIG. 1, the abnormality diagnosis system 1 includes a monitoring target selecting device 3, an abnormality sign monitoring device 4, and an abnormality diagnosis device 5. Note that an example in which the abnormality diagnosis system 1 is applied to a nuclear power generation plant will be described in the present embodiment. In addition, the abnormality diagnosis device 5 includes an acquisition unit 51, a storage unit 52, and an abnormality diagnosis control unit 53. The abnormality diagnosis system 1 acquires measurement parameters output from each of a plurality of measurement devices 21, 22, 23, . . . , and so forth provided in the nuclear power generation plant as plant operation data. The abnormality diagnosis system 1 diagnoses abnormality of a nuclear power generation plant having a nuclear reactor on the basis of the acquired measurement parameters. Here, the measurement devices 21, 22, 23, . . . , and so forth are devices that measure, for example, a pump pressure, an injected flow rate, an output flow rate, a water level, a pump bearing temperature, and the like. The monitoring target selecting device 3 acquires measurement parameters from each of the measurement devices 21, 22, 23, . . . , and so forth. The monitoring target selecting device 3 determines whether each of the acquired measurement parameters is an output target. The monitoring target selecting device 3 selects a measurement parameter determined as an output target and outputs the selected measurement parameter to the abnormality sign monitoring device 4. Note that a configuration, a determination method, and the like of the monitoring target selecting device 3 will be described below. The abnormality sign monitoring device 4 detects, for example, an abnormality sign of the nuclear power generation plant on the basis of a correlation value representing a correlation between the plurality of acquired measurement parameters. The abnormality sign monitoring device 4 outputs a detection result of the detected abnormality sign to the abnormality diagnosis device 5. The detection result of the abnormality sign includes, for example, a parameter transition, a parameter contribution degree, a position of a measurement parameter on a system, and the like. Note that a configuration, a determination method, and the like of the abnormality sign monitoring device 4 will be described below. The abnormality diagnosis device 5 identifies an abnormal event of the nuclear power generation plant on the basis of the detection result of the abnormality sign output from the abnormality sign monitoring device 4 and an operation history of the nuclear power generation plant in the past stored in its own device. The acquisition unit 51 acquires the detection result of the abnormality sign output from the abnormality sign monitoring device 4. The storage unit 52 stores various types of information generated on the basis of the operation history of the nuclear power generation plant in the past. For example, the storage unit 52 stores an abnormal event of the nuclear power generation plant, an abnormality sign facility associated with the abnormal event, and abnormality countermeasure processing associated with the abnormal event. In addition, the storage unit 52 stores a parameter transition for determination associated with the abnormal event and a parameter contribution degree for determination associated with the abnormal event. Note that the parameter transition for determination and the parameter contribution degree for determination are generated on the basis of the operation history of the nuclear power generation plant in the past. The abnormality diagnosis control unit 53 identifies an abnormal event and the like of the nuclear power generation plant by comparing and collating (that is, performing matching determination) the acquired detection result of the abnormality sign with various types of information stored by the storage unit 52. For example, the abnormality diagnosis control unit 53 compares the parameter transition, the parameter contribution degree, and the position of a measurement parameter on the system included in the acquired detection result of the abnormality sign with a parameter transition, a parameter contribution degree, and a position of a measurement parameter on the system stored by the storage unit 52. As a result of comparing these, if there is a parameter transition, a parameter contribution degree, and a position of a measurement parameter on the system matching in the detection result of the abnormality sign, the abnormality diagnosis control unit 53 identifies an abnormal event associated with the matched parameter transition, parameter contribution degree, and position of a measurement parameter on the system. Furthermore, the abnormality diagnosis control unit 53 identifies an abnormality sign facility and abnormality countermeasure processing associated with the identified abnormal event. Note that an abnormality diagnosis method of the abnormality diagnosis device 5 described above is an example, and is not limited thereto. Next, a configuration example and an operation example of the monitoring target selecting device 3 will be described. FIG. 2 is a diagram which shows a configuration example of the monitoring target selecting device 3 according to the present embodiment. As shown in FIG. 2, the monitoring target selecting device 3 includes an acquisition unit 31, a storage unit 32, a target selection unit 33, an output unit 34, and a trend monitoring unit 35. In addition, the target selection unit 33 includes a classification unit 331 and a selection unit 332. See also FIG. 12. The acquisition unit 31 acquires measurement parameters from each of the measurement devices 21, 22, 23, . . . , and so forth. The target selection unit 33 selects a measurement parameter that is an output target for each of the measurement parameters acquired by the acquisition unit 31. The classification unit 331 classifies a chronological change (behavior) of an N cycle (a first period) for each acquired measurement parameter. In addition, the classification unit 331 classifies a chronological change (behavior) of an N+1 cycle (a second period) for the acquired measurement parameter. Here, the N cycle is, for example, thirteen months, and the N+1 cycle is, for example, one week. In this manner, the N+1 cycle (the second period) may be a period shorter than the N cycle (the first period). The classification unit 331 obtains an inclination by, for example, linearly approximating a change of a measurement parameter over a time. The classification unit 331 classifies the behavior on the basis of a size of the inclination. A type of the behavior includes, for example, “rising” in which a measurement value increases as time elapses, “constant” in which a measurement value is within a range of a predetermined value as time elapses, and “falling” in which a measurement value decreases as time elapses. The selection unit 332 determines whether a classification result (hereinafter, referred to as a classification result of the N cycle) of the behavior in the N cycle classified by the classification unit 331 matches or is different from the classification result (hereinafter, referred to as a classification result of the N+1 cycle) of the behavior in the N+1 cycle. The selection unit 332 determines that the measurement parameter is an evaluation target when the classification result of the N cycle is different from the classification result of the N+1 cycle. When the classification result of the N cycle matches the classification result of the N+1 cycle, the selection unit 332 obtains a first mean value of a measurement parameter in the N cycle and a second mean value of the measurement parameter in the N+1 cycle, and determines whether the obtained first mean value and second mean value are the same as or different from each other. Note that the selection unit 332 determines that the first mean value and the second mean value are the same, for example, when a ratio or difference between the first mean value and the second mean value is within a range of a predetermined value. The selection unit 332 determines that the measurement parameter is an evaluation target when the first mean value and the second mean value are different. The selection unit 332 obtains a first inclination of a measurement parameter over a time in the N cycle and a second inclination of the measurement parameter over a time in the N+1 cycle when the classification result of the N cycle matches the classification result of the N+1 cycle, and determines whether the obtained first inclination and second inclination are the same as or different from each other. When the first inclination and the second inclination are different, the selection unit 332 determines that the measurement parameter is an evaluation target. When the classification result of the N cycle matches the classification result of the N+1 cycle, when the first mean value matches the second mean value, and when the first inclination matches the second inclination, the selection unit 332 determines that the measurement parameter is an output target (out-of-evaluation target) and outputs the acquired measurement parameter to the abnormality sign monitoring device 4 via the output unit 34. Note that the classification unit 331 may classify the behavior after standardizing the measured value using maximum and minimum values of the plurality of measurement parameters at the time of classifying the plurality of measurement parameters. Furthermore, with respect to the measurement parameter determined as an evaluation target, the selection unit 332 determines whether it is normal or abnormal by comparing the information stored in the storage unit 32 that the classification result of the N cycle and the classification result of the N+1 cycle are different from each other, that the first mean value and the second mean value are different from each other, or that the first inclination and the second inclination are different from each other. When it is determined to be normal and the monitoring target selecting device 3 acquires a measurement parameter that does not require monitoring, the selection unit 332 outputs the acquired measurement parameter to the abnormality sign monitoring device 4 via the output unit 34. When it is determined to be abnormal or the monitoring target selecting device 3 acquires a measurement parameter that requires monitoring, the selection unit 332 outputs the acquired measurement parameter to the trend monitoring unit 35. The storage unit 32 stores a classification threshold value for classifying measurement data. The storage unit 32 stores a classification result of the behavior of the N cycle and a classification result of the behavior of the N+1 cycle during a normal operation for each measurement parameter. The storage unit 32 stores an operation state of the N cycle and an operation state of the N+1 cycle. Note that the operation states include states before an inspection, after the inspection, before a replacement of the device, and after the replacement of the device, a state in which work has been performed, a state in which work has not been performed, and the like. The storage unit 32 stores the classification result of the behavior in the N cycle and the classification result of the behavior in the N+1 cycle for each measurement parameter. The output unit 34 outputs the measurement parameter selected by the target selection unit 33 to the abnormality sign monitoring device 4. The trend monitoring unit 35 performs measurement parameter monitoring (trend monitoring) and the like on the measurement parameter output by the selection unit 332. The trend monitoring unit 35 may output information indicating a result of the monitoring and the monitored measurement parameter to the abnormality sign monitoring device 4 or the abnormality diagnosis device 5 as a broken line of FIG. 1. Alternatively, the trend monitoring unit 35 may be displayed on a displayer (not shown) included therein. Next, an example of classifying a measurement parameter will be described. FIG. 3 is a diagram which shows an example of classifying a measurement parameter according to the present embodiment. In FIG. 3, the horizontal axis represents time and the vertical axis represents a measured value. A reference numeral g1 is an example in which a state (behavior) of a chronological change increases. The classification unit 331 determines that the classification result of the behavior is “rising” (a first behavior) when the size of the inclination obtained by linear approximation is equal to or greater than the classification threshold value. A reference numeral g2 is an example in which the state (behavior) of a chronological change is constant. The classification unit 331 determines that the classification result of the behavior is “constant” (a second behavior) when an absolute value of the size of the inclination obtained by linear approximation is less than the classification threshold value. A reference numeral g3 is an example in which the state (behavior) of a chronological change decreases. The classification unit 331 determines that the classification result of the behavior is “falling” (a third behavior) when the inclination obtained by linear approximation is a negative value and the absolute value of the inclination is equal to or greater than the classification threshold value. Next, the example of classifying a measurement parameter will be described. FIG. 4 is a diagram which shows an example of changing the behavior in the N cycle and the N+1 cycle according to the present embodiment. In the example shown in FIG. 4, a behavior of the N cycle is “rising” and a behavior of the N+1 cycle is “constant.” Note that the behavior in the N cycle is a behavior when normal. Then, the behavior in the N+1 cycle is an analysis target of the monitoring target selecting device 3. As described above, in the same measurement parameter, when the classification result of the behavior of the N cycle is different from the classification result of the behavior of the N+1 cycle, the selection unit 332 confirms a factor in which the classification results of the behaviors are different, and determines whether the difference is normal or abnormal. Here, the factor is, for example, when a measurement parameter of the N cycle is before work inspection and a measurement parameter of the N+1 cycle is after inspection, when the measurement parameter of the N cycle is before the replacement of the device, and the measurement parameter of the N+1 cycle is after the replacement of the device, or the like. For example, it may be normal that the behavior of the N+1 cycle in a period (for example, one week) is constant after restarting of the device. In addition, it may be normal that the behavior of the N+1 cycle is constant after the replacement of the device. Next, an information example stored by the storage unit 32 will be described. FIG. 5 is a diagram which shows an information example stored by the storage unit 32 according to the present embodiment. As shown in FIG. 5, the storage unit 32 stores classification results of the behaviors of the N cycle and the N+1 cycle classified by the classification unit 331. For example, the storage unit 32 stores a measurement parameter A in association with the “rising” of the classification result of the behavior of the N cycle and “rising” of the classification result of the behavior of the N+1 cycle. The storage unit 32 stores a measurement parameter C in association with “constant” of the classification result of the behavior of the N cycle and “constant” of the classification result of the behavior of the N+1 cycle. Next, an example of a processing procedure performed by the monitoring target selecting device 3 will be described. FIG. 6 is a flowchart which shows an example of a processing procedure performed by the monitoring target selecting device 3 according to the present embodiment. Note that the monitoring target selecting device 3 performs the following processing on each measurement parameter. (Step S1) The acquisition unit 31 acquires measurement parameters from each of the measurement devices 21, 22, 23, . . . , and so forth. (Step S2) The classification unit 331 classifies the chronological change (behavior) of the N cycle. In addition, the classification unit 331 classifies the chronological change (behavior) of the N+1 cycle. (Step S3) The selection unit 332 determines whether a classification result of the behavior of the N cycle is different from a classification result of the behavior of the N+1 cycle. The selection unit 332 proceeds to processing of step S7 when it is determined that the classification result of the behavior of the N cycle is different from the classification result of the behavior of the N+1 cycle (YES in step S3). The selection unit 332 proceeds to processing of step S4 when it is determined that the classification result of the behavior of the N cycle is the same as the classification result of the behavior of the N+1 cycle (NO in step S3). (Step S4) The selection unit 332 obtains a mean value of the measurement parameter in the N cycle and obtains a mean value of the measurement parameter in the N+1 cycle when both the classification result of the behavior of the N cycle and the classification result of the behavior of the N+1 cycle are “constant.” Subsequently, the selection unit 332 determines whether the mean value of the measurement parameter in the N cycle is different from the mean value of the measurement parameter in the N+1 cycle. For example, the selection unit 332 determines that the mean values are the same when a ratio or difference between the mean value of a measurement parameter in the N cycle and the mean value of the measurement parameter in the N+1 cycle is within a range of a first predetermined value. In addition, the selection unit 332 determines that the mean values are different when the ratio or difference between the mean value of a measurement parameter in the N cycle and the mean value of the measurement parameter in the N+1 cycle is outside the range of the first predetermined value. When the selection unit 332 has determined that the mean value of a measurement parameter in the N cycle and the mean value of the measurement parameter in the N+1 cycle are different from each other (YES in step S4), the procedure proceeds to the processing of step S7. When the selection unit 332 has determined that the mean value of a measurement parameter in the N cycle and the mean value of the measurement parameter in the N+1 cycle are the same as each other (NO in step S4), the procedure proceeds to processing of step S5. (Step S5) The selection unit 332 obtains an inclination of the measurement parameter over a time in the N cycle using, for example, linear approximation, and obtains an inclination of the measurement parameter over a time in the N+1 cycle using, for example, linear approximation, when both the classification result of the behavior of the N cycle and the classification result of the behavior of the N+1 cycle are “rising” or “falling.” Subsequently, the selection unit 332 determines whether the inclination in the N cycle is different from the inclination in the N+1 cycle. For example, the selection unit 332 determines that the inclinations are the same when a ratio or difference between the inclination in the N cycle and the inclination in the N+1 cycle is within a range of a second predetermined value. In addition, the selection unit 332 determines that the inclinations are different when the ratio or difference between the inclination in the N cycle and the inclination in the N+1 cycle is outside the range of the second predetermined value. Note that the second predetermined values may be the same value or different values for “rising” and “falling.” When the selection unit 332 has determined that the inclination in the N cycle and the inclination in the N+1 cycle are different (YES in step S5), the procedure proceeds to the processing of step S7. When the selection unit 332 has determined that the inclination in the N cycle and the inclination in the N+1 cycle are the same (NO in step S5), the procedure proceeds to processing of step S6. (Step S6) When the classification result of the behavior of the N cycle is the same as the classification result of the behavior of the N+1 cycle, when the mean value of a measurement parameter in the N cycle is the same as the mean value of the measurement parameter in the N+1 cycle, or when the inclination of a measurement parameter in the N cycle is the same as the inclination of the measurement parameter in the N+1 cycle, the selection unit 332 selects this measurement parameter as an out-of-evaluation target. After the selection, the selection unit 332 proceeds to processing of step S8. (Step S7) When the classification result of the behavior of the N cycle is different from the classification result of the behavior of the N+1 cycle, when the mean value of a measurement parameter in the N cycle is different from the mean value of the measurement parameter in the N+1 cycle, or when the inclination of a measurement parameter in the N cycle is different from the inclination of the measurement parameter in the N+1 cycle, the selection unit 332 selects this measurement parameter as an evaluation target. After the selection, the selection unit 332 proceeds to processing of step S9. (Step S8) The selection unit 332 outputs the selected measurement parameter to the abnormality sign monitoring device 4. The monitoring target selecting device 3 ends the processing. (Step S9) The selection unit 332 determines whether it is normal or abnormal that the behavior of the N cycle and the behavior of the N+1 cycle of the measurement parameter selected as an evaluation target in step S7 are different from each other on the basis of the information stored in the storage unit 32. Alternatively, the selection unit 332 determines whether it is normal or abnormal that the mean value of the measurement parameter in the N cycle is different from the mean value of the measurement parameter in the N+L cycle, which is set as an evaluation target in step S7, on the basis of the information stored in the storage unit 32. Alternatively, the selection unit 332 determines whether it is normal or abnormal that the inclination in the N cycle and the inclination in the N+1 cycle of the measurement parameter that is selected as an evaluation target in step S7 are different from each other on the basis of the information stored in the storage unit 32. The selection unit 332 proceeds to processing of step S10 when it is determined to be normal (normal in step S9), and proceeds to processing of step 11 when it is determined to be abnormal (abnormal in step S9). (Step S10) The selection unit 332 determines whether measurement parameter monitoring determined to be normal is necessary on the basis of the information stored in the storage unit 32. The selection unit 332 proceeds to the processing of step S8 when it is determined that the monitoring is not necessary (step S10; sign monitoring is possible). The selection unit 332 proceeds to processing of step S11 when it is determined that the monitoring is necessary (step S10; sign monitoring is not possible). This means that the selection unit 332 determines whether the difference in classification result of the behavior in the N+1 cycle is temporary or continuous. If the classification result of the behavior is continuously different, the selection unit 332 determines that it is an out-of-target because monitoring is not necessary. (Step S11) The selection unit 332 outputs the measurement parameter to the trend monitoring unit 35. The trend monitoring unit 35 performs trend monitoring and the like on the measurement parameter output from the selection unit 332. Note that, in the processing described above, when the classification results of the behavior of the measurement parameter of the N cycle and the N+1 cycle are the same in step S3, the selection unit 332 may select this measurement parameter and output the selected measurement parameter to the abnormality sign monitoring device 4. In addition, in the processing described above, when the mean values of the measurement parameter of the N cycle and the N+1 cycle are the same in step S4, the selection unit 332 may select this measurement parameter and output the selected measurement parameter to the abnormality sign monitoring device 4. Moreover, in the processing described above, when the classification results of a measurement parameter of the N cycle and the N+1 cycle are the same in step S3 or when the inclinations of a measurement parameter of the N cycle and the N+1 cycle are the same in step S5, the selection unit 332 may select this measurement parameter and output the selected measurement parameter to the abnormality sign monitoring device 4. Here, a reason for performing the processing of step S9 will be further described. Even if the behavior of the N cycle that is a normal behavior and the behavior of the N+1 cycle are different from each other, as the behavior of the measured value, the difference may be normal in some cases. For example, when the measurement device 21 is replaced after an operation of the N cycle, the behavior of the N+1 cycle may be different from the behavior of the N cycle. In addition, for example, when work is performed in a period of the N+1 cycle, the behavior of the N+1 cycle may be different from the behavior of the N cycle. The purpose of processing of step S9 is to select a normal measurement parameter to be output to the abnormality sign monitoring device 4 as described above. Note that, an example in which the classification unit 331 performs linear approximation on the measurement parameter to classify the behavior into three types has been described in the example described above, but the present invention is not limited thereto. The classification unit 331 may perform second-order approximation, third-order approximation on the measured value over a time to classify the behavior. Alternatively, the behavior having a change during a predetermined time is stored in the storage unit 32, and the classification unit 331 may classify the behavior with reference to the information stored in the storage unit 32. The behavior having a change during a predetermined time is, for example, a trapezoidal behavior, a behavior that repeats a plurality of times of rising and falling, a behavior that repeats a high measured value and a low measured value like a rectangular wave, or the like. In addition, the classification unit 331 may classify the behavior into two types or more, or may classify the behavior into 4 types or more. Next, a configuration example and an operation example of the abnormality sign monitoring device 4 will be described. FIG. 7 is a diagram which shows a configuration example of the abnormality sign monitoring device 4 according to the present embodiment. As shown in FIG. 7, the abnormality sign monitoring device 4 includes an acquisition unit 41, a storage unit 42, and an abnormality sign determination unit 43. The acquisition unit 41 acquires a measurement parameter output and selected by the monitoring target selecting device 3. The storage unit 42 stores the measurement parameters acquired by the acquisition unit 41 according to time series for each measurement parameter. In addition, the storage unit 42 stores a threshold value for determining whether there is an abnormality sign. The storage unit 42 stores a predicted value of a change in measured value according to time and an actual value when normal for each measurement parameter. Furthermore, the storage unit 42 stores a deviation threshold value between the actual value and the predicted value for each measurement parameter. The abnormality sign determination unit 43 derives each correlation value representing correlation strength of two measurement parameters among the plurality of measurement parameters. The abnormality sign determination unit 43 uses a correlation value obtained by adding all the derived correlation values of the measurement parameters as an abnormality indication value. The abnormality sign determination unit 43 determines (detects) that there is an abnormality sign when the monitoring indication value has exceeded the alarm transmission threshold value. The abnormality sign determination unit 43 determines (does not detect) that there is no abnormality sign when the monitoring indication value is equal to or less than the alarm transmission threshold value. The abnormality sign determination unit 43 outputs a result of detecting an abnormality sign to the abnormality diagnosis device 5 when it is detected that there is an abnormality sign. Note that an abnormality index value is an index of a degree of abnormality. Moreover, the alarm transmission threshold value is a threshold value for determining whether there is an abnormality sign. Next, an outline of the processing procedure of the sign monitoring performed by the abnormality sign monitoring device 4 will be described. The abnormality sign determination unit 43 selects two measurement parameters among the plurality of measurement parameters. FIG. 8 is a diagram which shows an example of two measurement parameters selected by the abnormality sign determination unit 43. In the example shown in FIG. 8, the abnormality sign determination unit 43 selects a “AA line outlet flow rate” as a first measurement parameter. Then, the abnormality sign determination unit 43 first selects a “BB pump bearing temperature” as a second measurement parameter. Next, the abnormality sign determination unit 43 selects an “AA line inlet temperature” as the second measurement parameter. Then, the abnormality sign determination unit 43 selects a “CC line outlet temperature” as the second measurement parameter. In this manner, the abnormality sign determination unit 43 selects all combinations of two measurement parameters among the plurality of measurement parameters. Then, the abnormality sign determination unit 43 obtains a correlation value indicating correlation strength between the selected first measurement parameter and second measurement parameter for all the combinations. The abnormality sign determination unit 43 obtains a sum of correlation values of all the combinations of the first measurement parameter and the second measurement parameter as an abnormality indication value. FIG. 9 is a diagram which shows an example of an abnormality indication value and an alarm transmission threshold value. In FIG. 9, the horizontal axis represents time, and the vertical axis represents an abnormality indication value. A broken line g11 represents an alarm transmission threshold value, and a waveform g12 represents a change in abnormality indication value over a time. In the example shown in FIG. 9, the abnormality indication value exceeds the alarm transmission threshold value at a time t1. The abnormality sign determination unit 43 obtains a contribution degree for the abnormality index value of each measurement parameter at the time t1 when the monitoring indication value has exceeded the alarm transmission threshold value. FIG. 10 is a diagram which shows an example of the contribution degree of a measurement parameter for a generated event. In FIG. 10, the horizontal axis represents a type of a measurement parameter and the vertical axis represents a contribution degree. In the example shown in FIG. 10, examples of a measurement parameter with a high contribution degree to the abnormality index value include a parameter A, a parameter B a parameter C, a parameter D, a parameter E, and a parameter F in order. As shown in FIG. 10, a measurement parameter with the highest contribution degree at a time t is the parameter A in FIG. 9. In this manner, the abnormality sign determination unit 43 extracts a measurement parameter with a high contribution degree at a time at which the abnormality indication value has exceeded the alarm transmission threshold value. Note that the abnormality sign determination unit 43 may extract at least one measurement parameter with a high contribution degree and extract two or more. The abnormality sign determination unit 43 individually monitors a measurement parameter by comparing an actual value and a predicted value for the extracted measurement parameter with a high contribution degree. FIG. 11 is a diagram which shows an example of measurement parameter monitoring. In FIG. 11, the horizontal axis represents time and the vertical axis represents a measured value. In addition, a waveform g21 represents an actual value and a waveform g22 represents a predicted value. In the examples shown in FIGS. 9 and 10, the abnormality sign determination unit 43 selects the parameter A as a measurement parameter with a high contribution degree. Then, the abnormality sign determination unit 43 reads a predicted value (or an actual value during normal) of a change in the measured value over a time of the parameter A stored in the storage unit 42. The abnormality sign determination unit 43 compares the predicted value and the actual value and monitors a deviation state between the predicted value and the actual value. The abnormality sign determination unit 43 monitors the deviation state between the actual value and the predicted value of the parameter A after the time t1, and detects that an abnormality sign occurs in the parameter A at a time t2 at which a difference between the actual value and the predicted value is equal to or greater than a deviation threshold value. The abnormality sign determination unit 43 outputs a result of detecting an abnormality sign to the abnormality diagnosis device 5 when it is detected that there is an abnormality sign. Note that the abnormality sign monitoring device 4 may perform notification from a notifier (not shown) when an abnormality indication value has exceeded an alarm transmission threshold value. In addition, when the abnormality indication value has exceeded the alarm transmission threshold value, the abnormality sign monitoring device 4 may cause a displayer (not shown) to display a measurement parameter with a high contribution degree and the contribution degree. Note that the abnormality sign method described above is an example, and the present invention is not limited thereto. For example, the method may be performed as described in Japanese Unexamined Patent application. First Publication No. 2015-62730 of Patent Document 1 in the prior art document. As described above, the abnormality sign monitoring device 4 is intended to detect an abnormality sign in an operation cycle. For this reason, the abnormality sign monitoring device 4 may have a difficulty to detect an abnormality sign due to a broken correlation when respective behaviors of measurement parameters between the N cycle and the N+1 cycle are different. For this reason, in the present embodiment, when the behavior of the N+1 cycle is abnormal, this measurement parameter is selected not to be output to the abnormality sign determination unit 43. Alternatively, in the present embodiment, this measurement parameter is excluded from the measurement parameter to be output to the abnormality sign monitoring device 4. In other words, in the present embodiment, an abnormality sign caused by a discontinuous change of a measurement parameter is screened in advance, and is excluded from an input of the abnormality sign monitoring device 4 that monitors an abnormality in an operation cycle. As a result, according to the present embodiment, erroneous detection of the abnormality sign monitoring device 4 can be reduced by selecting a measurement parameter to be output to the abnormality sign monitoring device 4. As a result, according to the present embodiment, it is possible to improve accuracy of the abnormality diagnosis system 1 at the time of identifying an abnormal event. In addition, in the present embodiment, instead of inputting a measurement parameter itself to the abnormality sign monitoring device 4, a monitoring target is selected by the monitoring target selecting device 3 described above and is input to the abnormality sign monitoring device 4. Moreover, in the present embodiment, a behavior of a measurement parameter before abnormality sign monitoring is analyzed (patterned) and classified. In addition, comparison is performed between operation cycles (for example, a previous operation cycle and a current operation cycle) in the present embodiment. As described above, in the present embodiment, a measurement parameter having a normal behavior is selected from a plurality of measurement parameters acquired by the monitoring target selecting device 3 and the selected measurement parameter is output to the abnormality sign monitoring device 4 when an abnormality sign of the nuclear power generation plant is detected on the basis of a correlation value indicating a mutual relationship between the measurement parameters. As described above, in the present embodiment, even if the correlation is broken when the measurement parameter changes discontinuously, such as an operation cycle is different, a measurement parameter having a normal behavior is selected and the selected measurement parameter is output to the abnormality sign monitoring device 4. As a result, according to the present embodiment, it is possible to solve a problem that makes it difficult to detect an abnormality sign in an operation cycle and to identify an abnormal event. In addition, in the example described above, an example in which the abnormality diagnosis system 1 is applied to the nuclear power generation plant has been described, but the present invention is not limited thereto. The abnormality diagnosis system 1 can be applied to a thermal power generation plant, a hydraulic power generation plant, a wind power generation plant, a solar power generation plant, and the like. In this case, the first period and the second period may be periods corresponding to the respective power generation plants. Note that a program for realizing all or a part of functions of the monitoring target selecting device 3 in the present invention is recorded in a computer-readable recording medium and a computer system is caused to read and execute this program recorded in the recording medium, and thereby all or a part of the processing performed by the monitoring target selecting device 3 may be performed. Note that the “computer-readable recording medium” herein includes hardware such as peripheral devices and an OS. In addition, the “computer system” includes a WWW system having a homepage providing environment (or a display environment). In addition, the “computer-readable recording medium” refers to a flexible disk, a magneto-optical disc, a portable medium such as a ROM and a CD-ROM, and a storage device such as a hard disk embedded in the computer system. Furthermore, the “computer readable recording medium” includes those that hold a program for a certain period of time like a volatile memory (RAM) in the computer system that is a server or a client when the program is transmitted via a network such as the Internet or a communication line such as a telephone line. In addition, the program described above may be transmitted from a computer system in which this program is stored in a storage device and the like to another computer system via a transmission medium or by a transmission wave in a transmission medium. Here, the “transmission medium” for transmitting the program refers to a medium having a function of transmitting information, like a network such as the Internet or a communication line (communication line) such as a telephone line. Moreover, the program described above may be for realizing a part of the functions described above. Furthermore, the program may also be a so-called difference file (difference program) that can realize the functions described above in combination with a program already recorded in the computer system. As described above, although modes for implementing the present invention have been described using the embodiments, the present invention is not limited to these embodiments, and various modifications and substitutions can be made within a range not departing from the gist of the present invention. According to at least one aspect among the aspects described above, the monitoring target selecting device can reduce erroneous detection and identify an abnormal event and the like effectively in the abnormality diagnosis system. 1 Abnormality diagnosis system 3 Monitoring target selecting device 4 Abnormality sign monitoring device 5 Abnormality diagnosis device 31 Acquisition unit 32 Storage unit 33 Target selection unit 34 Output unit 35 Trend monitoring unit 331 Classification unit 332 Selection unit 41 Acquisition unit 42 Storage unit 43 Abnormality sign determination unit 51 Acquisition unit 52 Storage unit 53 Abnormality diagnosis control unit |
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054065992 | summary | BACKGROUND OF THE INVENTION 1. Field of the Invention The invention is generally related to nuclear fuel assemblies and particularly to the fabrication of spacer grids used in nuclear fuel assemblies. 2. General Background Fuel assemblies for nuclear reactors are formed from a number of fuel rods held in position radially by spacer grids located along the length of the fuel rods. The spacer grids are formed from slotted metal strips containing stamped features, which are crisscrossed on edge to define individual cells for each fuel rod. The crossing metal strips are aligned at approximate right angles to each other and then welded together. The alignment method currently used is generally known as the pinning and framing process. After the metal strips are placed in the crisscross pattern, the next step is the insertion of pins into the individual cells of the egg crate. The "egg crate" is used to refer to the unwelded spacer grid inner structure comprised of intersecting, slotted strips that form a checkerboard-like array of cell locations. A tooling pin is then inserted into each cell. The pin interfaces with the base strip areas, as opposed to the contact points in each cell, to position and hold the strips in the proper orientation. The pinned egg crate is then placed into a welding frame that contains pressure pads which secure the egg crate to the frame. Pressure applied directly to the outer tooling pins is transferred to the inner locations through each strip and the adjacent pins. The load is repeatedly applied and released to act as a settling process to remove any tiny gaps or misalignments that may be present. The grid is then ready to be welded. This leaves a need for an improved fixturing system that can be used to align the egg crate with a minimum of steps and equipment, and provide greater dimensional control. SUMMARY OF THE INVENTION The present invention addresses the above need for a more straightforward manner in which to prepare the spacer grid egg crate for intersection welding. What is provided is a weld fixture for receiving and aligning the fuel rod contact points of a nuclear fuel bundle spacer grid, which eliminates the individual pin placement and settlement process in the previous equipment. Two plates each have a series of intermittent intersecting slots on one side used to align and position the fuel rod contact points referred to as outboard hardstops. The intermittent slots are not equally spaced, and the slot pitch is different in each intersecting direction to account for shrinkage during the welding process. One plate has a bore therethrough at each corner and a counterbore in a selected number of the polygons defined by the intermittent slots. A second plate has a bore therethrough at each corner and a counterbore in each of the polygons formed by the intermittent slots. Both plates have bores therethrough substantially at the intersection of the slots. A guide rod is received through each corner bore such that the intersecting slots on each plate are facing and aligned with each other. The guide rod in the orientation indicator corner is larger than the others to assure that the two fixture plates are properly oriented during assembly. Guide cell pins are positioned on the first plate in the recessed milled areas referred to as guide tube cell locations. Short alignment pins, located in counterbores in the polygons formed by the slots, are used to align and position fuel rod contact points referred to as inboard hardstops in the required locations on the first plate. The second plate contains tapered alignment pins, located in counterbores in the polygons formed by the slots, which are used to align and position fuel rod contact points referred to as softstops, with the exception of locations that also require positioning of an inboard hardstop, in which a combination pin is used to align and position both contact points simultaneously. The slots and assorted pins properly align the contact points of the spacer grid and the two plates are bolted together to secure the spacer grid in place. Laser welding of the interior intersections of the strips is accomplished by directing a laser beam through the bores at the slot intersections. |
claims | 1. A process for the treatment of a solution used for the decontamination of a radioactively contaminated surface, the solution containing radioactive metal ions and an organic complexing agent, the process comprising treating the solution with a reagent suitable for the destruction of the complexing agent and containing a metal capable of existing in more than one oxidation state, and raising the pH of the resultant solution to a level at which the metal of the reagent precipitates or flocs out of the solution. 2. A process according to claim 1 wherein the reagent is capable of providing active oxygen. claim 1 3. A process according to claim 2 wherein the active oxygen is in the form of a hydroxyl radical. claim 2 4. A process according to claim 1 wherein the reagent is Fenton""s reagent. claim 1 5. A process according to claim 1 wherein the reagent includes one or more catalysts. claim 1 6. A process according to claim 5 wherein the reagent includes one or more transition metal catalysts. claim 5 7. A process according to claim 6 wherein the reagent includes iron, chromium and/or copper. claim 6 8. A process according to claim 1 wherein the solution is additionally treated with electromagnetic radiation. claim 1 9. A process according to claim 8 wherein the electromagnetic radiation is UV or visible radiation. claim 8 10. A process according to claim 1 wherein the solution is treated at an ambient temperature. claim 1 |
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047724455 | description | DESCRIPTION OF THE PREFERRED EMBODIMENT Conventionally, control systems utilize sensors 30 (FIG. 2) to measure critical variables in a system controlled by a control system 35. The system being controlled may be in an aircraft or spacecraft, a chemical manufacturing plant, etc. The present invention can be applied to any such system, but will be described below for sensors 30 in a pressurized light water nuclear reactor. The sensors 30 provide sensor signals via lines 40 to an analog/digital converter 45, such as the A/D converter on an Intel 88/40 single board computer which can provide some preprocessing. Even using the best shielding available, the sensor signals received by the analog/digital converter 45 will contain some amount of noise since the sensors typically see process noise in the parameter they are monitoring. An example is steam generation level noise caused by the boiling in the steam generator. In addition, the sensor signals may contain direct current (DC) drift due to: poor initial calibration of the sensors 30, drifting of the sensors 30 during operation or failure of the sensors 30. The present invention identifies DC drift and noise signals present in the sensor signals by analyzing digitized sensor signals output by the analog/digital converter 45 in a processor 50, such as an Intel 86/05. The results of the analysis can be displayed on an auxiliary display 55 or used to control the nuclear reactor control system 35 after conversion to analog signals in a digital/analog converter 60. The first step in measuring direct current drift and noise according to the present invention is to generate component parity vector signals as described above with reference to equations (1) through (17). As illustrated in FIG. 1, the parity vectors define coordinates 100 that may be limited by an envelope 110 which defines the limit of the noise seen by a long sequence of measurements. Averaging the coordinates 100 over time produces a centroid 120 which corresponds to composite direct current drift, since the effects of noise will be averaged out. The centroid 120 can be calculated using standard matrix arithmetic. Given the coordinates of the centroid 120, it is possible to generate sensor direct current drift signals in the direction of the q or (l-1) measurement axes 130, 140 and 150 corresponding to sensors #1, #2 and #3, respectively. Since, as described above, the parity vector has l-1 dimensions when there are l sensors, the maximum number of constituent sensors corresponding to the sensor direct current drift signals is l-1. An example will be given below for l equals three, since the resulting parity vector will have two dimensions and can be easily illustrated. However, the number of sensors which can be handled by the present invention is not limited to three. If only two sensors are available, it may be possible to calculate one or more estimated sensor signals by assuming certain relationships between other measured parameters and the parameter being measured by the two remaining sensors (see the EPRI report on Research Project 1541 referenced above). One method for identifying the constituent sensors is to consider equations (16) and (17). Since the direct current drift signal is simply an average parity vector signal, if an average residual .eta..sub.j is kept for each of the sensors, the sensor producing the smallest average residual .eta..sub.1 will produce the smallest projection p.sub.1 according to equation (17). In other words, the sensor producing the smallest average residual .eta..sub.1 contributes the least to the DC drift signal and is, therfore, not a constituent sensor. The first method of identifiying constituent sensors may be better understood in view of a second method for identifying the contituent sensors of the DC drift signal. The second method is most easily explained by considering the graphical representation of two-dimensional parity. In the case of a two-dimensional composite DC drift signal or vector 155, defined by the centroid 120 and the origin 156, there are at most two constituent sensors which may have corresponding component DC drift signals. The constituent sensors can be identified by the angle between the vector 155 and a given measurement axis. Given the definition of matrix V, in particular equation (9) which defines an upper triangular matrix, the first column has only a single non-zero element V.sub.11 and thus can be easily assigned to the X-axis. Thus, in FIG. 3, the sensor signal provided by sensor number 1 is assigned to the X-axis 130. The matrix elements in the sensor signals from sensors 2 and 3 are thus displayed along axes 140 and 150 forming +120.degree. and -120.degree. angles with the positive X-axis, respectively. The angle .phi. formed between vector 155 and the X-axis 130 identifies the constituent sensors. If the angle .phi. is greater than zero and less than 60.degree., then the parity vector signal of DC drift signal will include a sensor signal from sensor number 1 with a positive deviation from the average and a sensor signal from sensor number 3 with a negative deviation, as illustrated in FIG. 3. Similarly, positive values of angle .phi. between 60.degree. and 120.degree., and 120.degree. and 180.degree. result in constituent sensors 2 and 3, and 1 and 2, respectively, while negative values of .phi. between 0.degree. and -60.degree., -60.degree. and -120.degree., and -120.degree. and -180.degree. have corresponding constituent sensors 1 and 2, 2 and 3, and 1 and 3, respectively, as illustrated in FIG. 3. After the constituent sensors have been identified, sensor DC drift signals can be found for each of the constituent sensors using geometric and trigonometric relationships. The case of three measurements of a scalar parameter will be used. Two examples of the computational methods will be given, but it is should be understood that others are possible. In the first method, it is noted that the measurement axis 150 of sensor #3 is defined by y=.sqroot.3 x (see the third column of matrix in equation (14)), while y=-.sqroot.3 x defines the measurement axis 140 of sensor #2. In four out of the six regions defined in FIG. 3, sensor #1 is a constituent sensor and the sensor #1 DC drift can easily be found by finding the X-intercept of a line 160 (FIG. 1) parallel to the measurement axis of the other constituent sensor (#2 or #3). The other sensor (#2 or #3) DC drift has a Y-coordinate equal to the Y-coordinate of the composite DC drift and its X-coordinate can be found by solving the appropriate equation above defining the measurement axis for that sensor. Thus, the component DC drift signals 210 and 220 in FIG. 1 can be easily found. In the case of a composite DC drift signal in a region in which the constituent sensors are 2 and 3, calculation of the component DC drift signals is slightly more difficult, but the same principles are involved. See for example, the composite DC drift vector 155' in FIG. 3 in a region having constituent sensors #2 and #3. A second method of calculating sensor DC drift signals utilizes the residual signals corresponding to each of the sensors. As noted above, the projection along each of the measurement axes is defined by equation (17). However, .sqroot.l/(l-1) is simply a scale factor which puts calculations using the residuals into the same scale as those using matrix V, defined in equation (14). It is possible to perform the calculations directly on residuals .eta..sub.j and multiply by the scale factor .sqroot.l/(l-1) at a later time or perform scaling using some other factor. With reference to FIG. 3, the coordinates of the vector 155' representing composite DC drift and the sensor #2 and sensor #3 DC drift signals represented by vectors 220' and 230' can be calculated from the projections p.sub.j along the measurement axes, as defined by equation (17). By definition, a line 170 between the tip of the DC drift vector 10' and the tip p.sub.3 of the projection along the sensor #3 measurement axis 150 forms an angle of 30.degree. with the measurement axis 130, since the smallest angle between all of the measurement axes if 60.degree.. Therefore, X=-p.sub.3 /sin 30.degree.. The X-coordinate X.sub.DC equals the projection p.sub.1 along the sensor #1 measurement axis 130 and the Y-coordinate Y.sub.DC of the composite DC vector 10' is defined by equation (24). EQU Y.sub.DC =(p.sub.1 -X) tan 30.degree. (24) The sensor #2 and #3 DC drift signals represented by vectors 220' and 230' can be calculated from the composite DC drift vector 155' utilizing the parallelogram having sides formed by vectors 220' and 230' and having a major diagonal formed by vector 155', together with the triangle defined by the tips of vectors 155', 220' and p.sub.3. The distance D between p.sub.3 and the tip of vector 10' can be found by vector subtraction or as D=p.sub.3 cos .theta., where .theta.=120.degree.-arctan Y.sub.DC /X.sub.DC. Thus, the sensor #2 DC drift signal DC.sub.2 =-D/sin 60.degree. and the sensor #3 DC drift signal DC.sub.3 =p.sub.3 -DC.sub.2 cos 60.degree.. When one of the consituent sensors of the composite DC drift signal is sensor #1, the sensor DC drift signals are easily calculated. For example, the sensor #1 and #3 DC drift signals DC.sub.3 and DC.sub.1 represented by vectors 210 and 220 (FIG. 1), respectively, are found as DC.sub.3 =-Y.sub.DC /sin 60.degree. and DC.sub.1 =X.sub.DC +DC.sub.3 cos 60.degree.. Increased sensitivity to sudden changes in DC drift can be provided according to the present invention by weighting the average which produces the centroid 120. Such weighting is provided by equation 25, where X.sub.i,j is the weighted running average for sensor direction i of the parity vector in the jth sample of the sensor signals, (1-W) is a weighting coefficient in which W is a weighting function and X.sub.i,j is the sensor #i parity vector signal for in the jth sample. The weighting function W may, for example, be 0.01 or 0.02, representing an average over the last 100 to 50 samples, respectively. EQU X.sub.i,j =(1-W)X.sub.i,j-1 +W.multidot.X.sub.i,j (25) Equation (25) represents first order lag, but other equations may be used to provide a desired response, representing, e.g., second order lag, for recent samples of the sensor signals, as is known the art. Other variations are also possible. For example, the parity vector can be averaged in two dimensions and periodically (e.g., every 50 or 100 samples) the sensor components can be calculated, as described above. As noted above, noise on lines 40 (FIG. 2) cause fluctuations in the parity vector corresponding to the sample of the sensor signals. Thus, in a recent sample of the sensor signals, a parity vector 310 (FIG. 5) may be produced having coordinates which are different from the coordinates of the composite DC drift vector 155. Using standard vector arithmetic, a composite instaneous noise signal can be calculated and represented by vector 320. The composite instaneous noise signal has sensor instaneous noise signals corresponding to constituent sensors. A magnified drawing of the region on display screen 20 bounded by positive values of sensor #1 and negative values of sensor #3 is illustrated in FIG. 6. Component instaneous noise signals corresponding to constituent sensors and represented by vectors 340 and 350 can be found by transposing the axes 130, 140 and 150 to the centroid 120 to produce a coordinate system for noise signals defined by axes 130', 140' and 150', corresponding to sensors #1, #2 and #3, respectively. After the constituent sensors of the noise signal 230 are identified, the sensor instantaneous noise signals represented by vector 340 and 350 can be found using the same methods used to find the sensor DC drift signals. The average amount of noise in each sensor signal can be found using a weighted RMS average of the RMS instantaneous noise signals. Thus, a sensor noise signal N.sup.2.sub.k,j can be calculated using standard matrix arithmetic and equation (26), where N.sup.2.sub.k,j is the sensor noise signal for sensor number k in the jth sample of the sensor signals, W.sub.n is a weighting function for noise signals which may or may not be the same as the weighting function W for DC drift signals, X.sub.j is the parity vector signal for the jth sample of the sensor signals, X.sub.j is the direct current drift vector, and k is a unit vector along the axis for sensor number k. EQU N.sup.2.sub.k,j =(1-W.sub.n)N.sup.2.sub.k,j-1 +W.sub.n [(X.sub.j -X.sub.j).multidot.k].sup.2 (26) The lines parallelling each of the axes 130, 140 and 150 in FIGS. 1 and 2 indicate the error bounds of each of the sensors. In the above example, it has been assumed that the error bound for each of these sensor signals is the same and has a value of b, as used in equations (4), (5a), (5b) and (9). However, it is possible for the error bounds of the sensors to differ and this can be taken into account in the parity-space algorithm. The many features and advantages of the present invention are apparent from the detailed specification, and thus it is intended by the appended claims to cover all such features and advantages of the system which fall within the true spirit and scope of the invention. Further, since numerous modifications and changes will readily occur to those skilled in the art, it is not desired to limit the invention to the exact construction and operation illustrated and described, accordingly, all suitable modifications and equivalents may be resorted to falling within the scope and spirit of the invention. |
044977670 | description | DESCRIPTION OF A PREFERRED EMBODIMENT Referring now to the drawing, and more particularly to FIG. 1 thereof, there is illustrated therein a portion of a fusion reactor system of the Tokamak-type. More specifically, there is depicted in FIG. 1 of the drawing, a compression hub, generally designated by reference numeral 10, constructed in accordance with the present invention. Cooperatively associated with the compression hub, as shown in FIG. 1, are a plurality of superconducting magnets 12. Inasmuch as all of the superconducting magnets 12 are of identical construction, it has been deemed appropriate to designate each of the magnets by means of the same reference numeral, i.e., 12. In accordance with the illustrated embodiment of the invention, the superconducting magnets 12 comprise six in number. Although the compression hub 10 is intended to be employed in a Tokamak-type fusion reactor system, it is not deemed necessary for purposes of obtaining an understanding of the present invention that a complete description of the nature of the construction and the mode of operation of such a fusion reactor system be set forth herein and/or illustrated in the drawing. Rather, it is deemed sufficient to merely note herein that in accordance with the mode of operation of such a fusion reactor system, thermal power is generated as a consequence of the ignition of plasma. Moreover, there exists a need to effect confinement of the plasma. The latter function, in turn, is accomplished magnetically through the use of a plurality of superconducting magnets, such as the superconducting magnet 12 shown in FIG. 1 of the drawing. Namely, the magnets 12 are capable of generating intense magnetic fields of sufficient strength to achieve the desired confinement of the plasma. In order to be operative for its intended purpose, there are basically two major functional requirements that the compression hub 10 must be capable of fulfilling. First, the compression hub must be capable of successfully resisting the intense forces produced by the magnets 12 tending to draw the latter together towards a common point. Secondly, the compression hub 10 must be susceptible to being cooled to the same relative temperature as the superconducting magnets 12; namely, to a cryogenic temperature of approximately 4.2.degree. kelvin. Moreover, the latter cooling of the compression hub 10 must be achieved without adversely affecting the structural strength thereof. Proceeding now with a description of the nature of the construction of the compression hub shown in FIG. 1, reference will be had for this purpose particularly to FIGS. 2, 3 and 4 of the drawing. Thus, with reference first to FIG. 3 of the drawing, there is depicted therein a prior art form of compression plate, the latter being designated therein generally through the use of the reference numeral 14. In accord with the teachings of the prior art, and as exemplified by the description thereof contained in previously referenced U.S. Pat. No. 4,174,254, the compression plate 14 is designed to be employed as one of a multiplicity of such plates which when suitably assembled one with another collectively function to produce a prior art form of compression hub. Inasmuch as the details of construction and the mode of operation thereof can be adequately found set forth in the aforementioned patent, it is not deemed necessary that they be reiterated herein. Rather, it is deemed sufficient to merely take note of the fact that the compression hub 14 is suitably constructed so as to embody the requisite means whereby the strength and cooling characteristics that a compression hub must possess in order to be operative for its intended purpose when employed in a Tokamak-type fusion reactor system are achievable therewith. To this end, the compression plate 14 is formed of aluminum, stainless steel, or some other metal suitable for employment for purposes of resisting substantial forces at cryogenic temperatures. Further, the compression plate 14 embodies a polygonal shape, i.e., that of a hexagon. In addition, the upper and lower planar surfaces of the compression plate 14 both have formed therein a suitable number of recesses (not shown). The latter recesses (not shown) are suitably located in spaced relation one to another and are suitably dimensioned so as to each be capable of receiving therewithin a shear member 16. The function of the shear members 16 is both that of a force resisting member and that of a spacer. Namely each shear member 16 functions as a torque and shear force resisting member to resist the forces imparted to the compression plate 14 as a consequence of the action of the superconducting magnets cooperatively associated therewith. With regard to the spacer function, each shear member 16 operates to effect a spacing between adjoining, superimposed compression plates 14. The shear members 16 are maintained in the aforementioned recesses (not shown) by virtue of the fact that adjoining, superimposed compression plates 14 are interconnected along their perimetric surfaces such as by means of welding. The inherent strength of the individual compression plates 14 as well as the manner of effecting the interconnection therebetween is operative to provide a prior art form of compression hub that embodies the requisite degree of strength so as to be usable in Tokamak-type fusion reactor system. In conclusion, provision is also made for coolant flow between adjoining, superimposed compression plates 14 by providing each of the latter with a suitable number of openings. More specifically, as shown in FIG. 3 the compression plate 14 is provided with an opening 18 formed therein so as to be located in proximity to one of the hexagonal corners thereof. Moreover, the compression plate 14 has a larger opening 20 formed at the center thereof so as to extend completely therethrough the same as the previously described opening 18. For purposes of clarity of description, another small opening, identified by reference numeral 22, is depicted in FIG. 3 by means of phantom lines. The latter is shown located adjacent one of the other hexagonal corners of the compression plate 14. The opening 22 is intended to depict an opening similar to the opening 18, but one which is suitably provided in the compression plate (not shown) that is located below the compression plate 14 illustrated in FIG. 3. To summarize, the openings 18, 20 and 22 function to establish, along with similar openings provided in others of the multiplicity of compression plates 14, a fluid flow passage for coolant through the compression plates 14 in a manner which can be found more fully described in U.S. Pat. No. 4,174,254. In accord therewith, a prior art form of compression hub is provided that is capable of being cooled to the desired cryogenic temperature. The compression hubs embodying prior art forms of construction that have been known heretofore are disadvantageously characterized in the fact that the constructions which they embody permit eddy currents, which may be induced therein as a consequence of the occurrence of changes in magnetic field flux, to circulate therethrough. The existence in such compression hubs of freely circulating eddy currents in turn provide various losses in the form, for example, of the creation of heat losses, that can have an adverse effect on the performance characteristics of the compression hub. Thus, a need has been evidenced for a new and improved form of compression hub, which is advantageously characterized in the fact that eddy currents, which may be induced in the compression hub from changes in magnetic field flux, are prevented from circulating therethrough. Namely, a need has been evidenced for a compression hub with eddy current prevent means. Moreover, the compression hub 10 shown in FIG. 1 and which is now to be described hereinafter comprises such a compression hub with eddy current prevent means. As best understood with reference to FIGS. 4 and 2 of the drawing, the compression hub 10 embodies, yet to be described, eddy current prevent means, the latter being operative to impede the circulation through the compression hub 10 of eddy currents that may be induced therein due to changes in magnetic field flux. In accord with the preferred form of the invention, the aforesaid eddy current prevent means comprises means with which each of the compression plates that collectively comprise the compression hub 10 is provided. More specifically, the compression hub 10, as best understood with reference to FIG. 2, consists of a multiplicity of compression plates 24 that are superimposed one upon the other so as to form a layered assembly thereof. Each of the compression plates 24 embodies a polygonal shape, i.e., that of a hexagon. However, some other form of a multi-sided figure could be utilized without departing from the essence of the invention. Moreover, each of the plates 24 is formed from a suitable material that is capable of providing the strength characteristics desired at cryogenic temperatures. Preferably, the compression plates 24 are suitably interconnected along their perimeters as a result of being welded one to another. As will be described more fully hereinafter, the compression plates 24 are each provided with suitable flow passages whereby coolant can be made to flow through the compression hub 10. To this end, the compression hub 10 is provided with suitable inlet means and outlet means whereby coolant can be fed to the compression hub 10 and after the passage therethrough can be removed therefrom. Inasmuch as such inlet means and outlet means are of conventional construction which is well-known to those skilled in the art, it is not deemed necessary to further describe them herein. In accord with the illustration of the compression hub 10 in the drawing of the instant application, the aforementioned outlet means has been depicted in FIG. 2 wherein it is identified by the reference numeral 26, whereas a showing of the aforementioned inlet means is omitted from the drawing. Proceeding now with a more detailed description of the nature of the construction of the compression plates 24, reference will be had for this purpose to FIGS. 4 and 5 of the drawing. In accord with the present invention, alternative forms of construction for the compression plates 24 are disclosed. More specifically, alternative constructional forms for the eddy current prevent means with which the compression plates 24 are provided are disclosed. Thus, referring first to FIG. 4 of the drawing, there is illustrated therein a compression plate 24 which in structure bears a resemblance to the prior art form of compression plate 14 described previously above, and illustrated in FIG. 3. That is, the compression plate 24 as noted above embodies a hexagonal configuration and is made of a material that possesses the strength desired at cryogenic temperatures. In addition, the compression plate 24 shown in FIG. 4 is provided with a plurality of suitable recesses (not shown) in each of which a shear member 28 is designed to be suitably received such that the latter is captured therein when another compression plate 24 is superimposed thereover. Like the shear member 16 with which the prior art compression plate 14 is provided, the shear member 28 is intended to not only offer resistance to the torque and shear forces which are imparted to the compression hub 10 from the superconducting magnets 12, but also to establish the desired spacing between adjacent compression plates 24. Continuing with the description of the compression plate 24 of FIG. 4, the latter further includes a plurality of openings, i.e., the small opening 30 provided therein adjacent one of the hexagonal corners of the plate 24 and the larger opening 32 formed substantially at the center thereof. The openings 30 and 32 are operative in the manner of the openings 18 and 20 of the compression plate 14 to establish a flow passage through the compression plate 24 as well as between adjacent compression plates 24. To this end, there is depicted in phantom lines in FIG. 4 another small opening, identified therein by the numeral 34, which is intended to denote the small opening similar to the opening 30, which is formed in the plate 24 that is located directly below the plate 24 that is shown in FIG. 4. Note is taken here of the fact that preferably the compression plates 24 are suitably positioned relative to each other so as to form a layered assembly wherein the openings 30 are offset as between adjacent plates 24. This is best understood with reference to FIG. 2 of the drawing. The compression plate 24, unlike the prior art compression plate 14, embodies an eddy current prevent means that is operative to impede the circulation of induced eddy currents therethrough. In this end, the compression plate has a radial cut provided therein. The latter radial cut extends from the outer perimeter of the plate 24 to the inner perimeter thereof, i.e., the surface that defines the circumference of the large opening 32. Furthermore, the radial cut extends completely through the plate 24, i.e., through the entire thickness of the latter. Finally, in accord with the best mode embodiment of the invention, the afore-described radial cut which effectively forms an interruption in the circumference of the plate 24 has inserted therein a suitably dimensioned and configured piece 36 of insulative material. Any suitable known type of material possessing the desired insulative characteristics and capable of being employed at the cryogenic temperatures to which the compression plates 24 are designed to be cooled may be employed. Moreover, any conventional form of securing means may be employed for purposes of retaining the piece 36 of insulative material in place within the afore-described radial cut. It is to be understood that in accord with the best mode embodiment of the invention each of the compression plates 24 which the compression hub 10 embodies would be provided with such a radial cut in which a piece 36 of insulative material is emplaced. Moreover, as best understood with reference to FIG. 2 of the drawing, the compression plates 24 are preferably suitably arranged when in the assembled condition such that the insulative pieces 36 are offset relative to each other at least as between adjacent plates 24. For purposes of completing the description of the compression plates 24 of FIG. 4, note is taken of the slots 38 that are provided in the outer surfaces of each of the six sides thereof. The slots 38 extend the full thickness of the plate 24, and are intended to be operative for purposes of effecting the interconnection of the superconducting magnets 12 to the compression hub 10. To this end, each of the superconducting magnets 12 embodies a protrusion 40, as shown in FIG. 1, which is designed to be received within a corresponding one of the slots 38. Inasmuch as the function of the slots 38 and protrusions 40 is not related directly to the subject matter of the present invention, any further discussion of the nature of the construction and/or mode of operation thereof has been omitted from herein. Should a further description thereof be desired, reference may be had for this purpose to the teachings thereof that are to be found contained in the prior art. Turning next to a consideration of the structure that is to be found illustrated in FIG. 5 of the drawing, the latter Figure depicts a compression plate constructed in accord with the teachings of the present invention that embodies an alternative form of eddy current prevent means. More specifically, there is shown in FIG. 5, a compression plate, generally designated by the reference numeral 24', which is basically similar in construction and mode of operation to the compression plate 24 illustrated in FIG. 4 that has been described hereinabove. In view of the similarity therebetween, any elements of the compression plate 24' of FIG. 5 that find correspondence with an element of the compression plate 24 of FIG. 4 is identified in FIG. 5 through the use of the same reference numeral that has been employed in FIG. 4 for designating the similar element therein, but with the addition in FIG. 5 of a prime to the numeral. Thus, by way of a brief description of the nature of the construction of the compression plate 24', the latter embodies the shape of a hexagon, and is formed from a suitable material possessing adequate strength at cryogenic temperatures. A plurality of shear members 28' are received in recesses (not shown) provided for this purpose in both the upper and lower planar surfaces of the plate 24'. Openings 30' and 32' of differing dimensions are suitably formed in the plate 24' so as to be operative as flow passages for coolant. There is also shown in phantom lines in FIG. 5 an opening 34' that is formed in the plate 24' which lies immediately below the plate 24' that appears in FIG. 5. Slots 38' are provided in the outer surface of each of the six sides of the plates 24', and are designed to cooperatively receive therewithin the protrusions 40 with which the superconducting magnets 12 are provided. As stated above, the compression plate 24' of FIG. 5 differs structurally from the compression plate 24 of FIG. 4 insofar as concerns the eddy current prevent means which the former embodies. By way of reiteration, in both cases, the function of the eddy current prevent means is to impede the circulation of induced eddy currents through the compression plate, be it the plate 24 of FIG. 4 or the plate 24' of FIG. 5. To this end, both the plate 24 and the plate 24' are provided with at least one interruption that is formed in the circumference thereof. In the case of the plate 24 of FIG. 4, as has been described above, the interruption takes the form of a radial cut that is suitably located therein, and which is designed to receive therewithin a piece 36 of insulative material. However, in the case of the plate 24' of FIG. 5, the latter in essence embodies two interruptions that are suitably located therein so as to be aligned along a common axis. More specifically, the aforesaid two interruptions are preferably created by making the plate 24' in two segments that are designed to mate with each other such that a space exists therebetween, i.e., so that the aforedescribed two interruptions exist therebetween. Moreover, each of the two interruptions is designed to extend from the outer surface of the plate 24' to the inner surface thereof, i.e., the inner surface thereof that defines the circumference of the opening 32'. Thus, in essence, the aforesaid two interruptions coact to provide the plate 24' with a through cut that extends across the entire width and thickness of the plate 24'. Finally, in accord with the teachings of the present invention, a piece of insulative material, 42 and 44, respectively, is inserted in each of the two interruptions that exist by virtue of forming the plate 24' as two mating segments. Briefly then, the two pieces 42 and 44 function in a manner similar to the piece 36 with which the plate 24 is provided. In conclusion, it is to be understood that the plates 24' could be substituted for the plate 24 to form the compression hub that is illustrated in FIGS. 2 and 1 of the drawing. Thus, in accordance with the present invention, there has been provided a novel and improved compression hub that is designed to be cooperatively associated with a plurality of superconducting magnets in a Tokamak-type fusion reactor system. Moreover, the subject compression hub of the present invention embodies a sufficient structural strength as to be capable of resisting the intense forces produced by the superconducting magnets that tend to draw the latter together towards a common point whereat the compression hub is located. In addition, in accord with the present invention, the compression hub embodies a construction that permits the latter to be cooled to a temperature that is commensurate with the operating temperature of the superconducting magnets, while yet enabling the compression hub to retain the structural strength required thereof. Further, the compression hub of the present invention embodies means operative to impede the circulation of eddy currents therethrough, while yet possessing the strength and cooling characteristics desired of a compression hub. Additionally, in accordance with the present invention, a compression hub is provided embodying such eddy current prevent means wherein the latter consists of an interruption provided in the surface of the elements that collectively comprise the compression hub. Also, the compression hub of the present invention embodies eddy current prevent means wherein the interruption provided in the surface of the elements takes the form of a radial cut in which an insulative material is inserted. Finally, in accord with the present invention, a compression hub is provided wherein the interruption provided in the surface of the elements is effected by fabricating the elements from multiple segments that in the assembled state are separated one from another by means of insulative material. While only one embodiment of my invention has been shown, it will be appreciated that modifications thereof, some of which have been alluded to hereinabove may readily be made thereto by those skilled in the art without departing from the essence of the invention. I, therefore, intend by the appended claims to cover the modifications alluded to herein as well as all other modifications which fall within the true spirit and scope of my invention. |
description | This application claims the benefit of U.S. Provisional Application No. 61/878,584, filed on Sep. 16, 2013 and U.S. Provisional Application No. 61/925,417, filed on Jan. 9, 2014. The entire teachings of the above applications are incorporated herein by reference. There are 15 actinide elements, each with several important isotopes. All actinide isotopes are unstable to radioactive decay involving emission of alpha or beta particles along with gamma rays, as also are all isotopes of the next five atomic numbers below the actinides. Instability and ease of fission of at least some actinide isotopes generally increases with ascending atomic number and spontaneous fission also becomes common in the higher actinides. All actinide atoms are fissionable, meaning each can be fissioned if its atomic nucleus is struck by a sufficiently energetic neutron. Actinide isotopes can be classified according to whether they are fissile, meaning that they can be fissioned by slow neutrons having room-temperature thermal motion energies of about 0.025 eV. Examples of fissile actinide isotopes include uranium-233, uranium-235, plutonium-239 and plutonium-241, but of these only uranium-235 is found in nature. Only the fissile actinide isotopes can support fission chain reactions, since emitted fission daughter neutrons having enough energy to fission other non-fissile actinides are rare. Uranium-235 with a 704 million year half-life is the only naturally occurring fissile isotope. Uranium-238, which has a 4.47 billion year half-life is 138 times more abundant and thorium-232, with a 14.05 billion year half-life, is about 500 times more abundant. Both can be fissioned, releasing about 200 MeV of energy per atom. However, they are only fissionable, not fissile. A sustained fission chain reaction is impossible with either of these more plentiful isotopes. Isotopic enrichment is a difficult industrial process in which a mixture of two or more isotopes of an element is divided into two different mixtures, an “enriched” mixture with an increased concentration of one isotope and a “depleted” mixture with a depressed concentration of the same isotope. Light water reactors (LWRs), the reactor design currently responsible for producing the majority of the world's nuclear power, rely on the rare uranium-235 isotope as fuel, leaving most of the uranium-238 isotope unused along with the thorium-232 isotope which LWRs entirely ignore. Indeed, the total utilization of mined uranium is only about 1%, with 99% discarded as depleted uranium or as the main component of spent nuclear fuel (SNF). Two alternative physics pathways exist to make use of the two naturally abundant actinide isotopes, uranium-238 and thorium-232, as follows: Pathway one: Provide a source of sufficiently energetic neutrons to induce fissions without a chain reaction. Pathway two: Transmute the fissionable isotopes into fissile isotopes, then fission them in a chain reaction. Initially, there was no known source of fast neutrons with enough generation efficiency to cause net energy release from the first pathway. That changed when the first H-bomb was tested, but for non-explosive applications it remained true that no energy-efficient source of fast neutrons was available. Isotopes able to be transmuted into fissile isotopes by absorbing a neutron, followed in some cases by beta decay processes, are known as fertile isotopes. All non-fissile actinides are fertile in this sense. Thus, the second pathway for the two fertile and fissionable, but not fissile, natural actinide isotopes is based on the following chained nuclear reaction sequences: Plutonium-239 and uranium-233 support fission chain reactions as well as natural uranium-235 does. Every critical nuclear fission reactor incorporating either some uranium-238 or some thorium-232 causes these fissile fuel production reactions to occur. The ratio of the rate of production of new fissile atoms divided by the rate of fissioning fissile atoms is an important reactor parameter termed the Conversion Ratio (CR) if less than unity or the Breeding Ratio (BR) if greater than unity. Typical CR values are 0.6 for LWRs and can exceed 0.9 for a molten salt reactor (MSR) with a graphite moderator. Reaching or exceeding unity would imply converting all fertile atoms into fissile atoms then fissioning them. To exceed unity using uranium-238, it is necessary to use fissile fuel with high plutonium-239 content, minimize neutron captures in structural material, surround the core with an optimized uranium-238 blanket, and frequently recycle the fuel and blanket through a reprocessing center in order to chemically extract bred plutonium from the blanket and insert it into the core while also removing neutron-absorbing fission products. Experimental Breeder Reactor 1 (EBR-1), the world's first liquid metal cooled fast breeder reactor (LMFBR) began operation in December 1951, producing 200 kW of electricity from its 1.4 MW thermal power. By 1953 it had demonstrated a net breeding gain, thus confirming the conceptual design of a fuel breeder using plutonium fuel with a non-moderating coolant. Much larger LMFBR designs for electricity production are highly constrained but have been built and operated in several countries, all exhibiting BR values slightly exceeding one. In principle such fission breeders could consume most of their actinide feedstock input streams. However, they have not been widely deployed, partly because breeders have higher costs than LWRs both for initial capital outlays and ongoing plutonium fuel recycling, but also due to fears about breeder reactor safety and special breeder concerns about terrorism and weapons proliferation. Ever since the aforementioned fission breeder design difficulties, costs, and constraints were recognized, there have been efforts to find alternative approaches to harvesting fission energy from the more abundant non-fissile but fissionable actinides. Other than the breeder reactor, the only non-fusion approach ever suggested was Carlo Rubbia's 1995 “Energy Amplifier” which relied on spallation. In nuclear spallation, a beam of very high energy ions emerging from a particle accelerator, typically hydrogen ions with energies between 800 MeV and 7,000 MeV per proton, is focused on a heavy metal target, typically of mercury, lead, or tantalum. Each spallation impact of a very high energy proton on a heavy metal nucleus then sprays out typically 20 to 30 high energy neutrons. However, some protons may fail to cause spallation so the efficiency may not be high. In the Energy Amplifier scheme, high energy spallation neutrons would then cause fissions in thorium-232 or uranium-238 via pathway one, thus releasing even more neutrons which in turn would be absorbed causing pathway two transmutation chains ending in uranium-233 or plutonium-239. A concern about this method is whether the very high energy investment needed per spallation neutron could be offset by the energy content of the fissile fuel produced. Another concern is the high cost and large size of present particle accelerators. Unlike accelerator driven systems, use of a fusion neutron source may be less of a concern since fusion releases its own nuclear energy. Fusion schemes can be classified according to whether their fusion fuel feeds are deuterium only (DD) or deuterium tritium (DT). If the neutron source is a fusion system using a feedstock of deuterium only, then half of its resulting DD fusion reactions would produce 2.45 MeV neutrons. These do not carry enough energy for pathway one but are adequate for pathway two. If instead a fusion neutron source uses a 50/50 DT feedstock of deuterium and tritium then almost all neutrons produced would be 14.1 MeV neutrons adequate for pathway one. Furthermore, for identical fusion plasma temperature and pressure conditions, the neutron flux will be two orders of magnitude more intense than in the DD fuel case. While it was recognized in the 1950s that neutron bombardment of fissionable actinide isotopes could greatly expand fissile fuel supplies, there were no controlled fusion neutron sources with adequate energy efficiency. Particle accelerators can easily produce fusion reactions but coulomb scattering is so strong that their typical efficiencies are only about 0.001%. After some initial analyses the subject of hybrid systems was not pursued further. This changed in 1969 when Soviet researchers announced their tokamak device had confined a plasma with temperatures approaching the thermonuclear fusion range. After an international team confirmed the temperatures, other researchers around the world built their own tokamaks and began experiments with fusion-relevant plasmas. A fusion concept developed in the mid-1970s envisioned a non-Maxwellian ion velocity distribution plasma known as the Two Component Tokamak (TCT). This TCT scheme using neutral beams was the basis for the Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory (PPPL). TFTR performance culminated in a plasma fusion energy gain (Q), i.e., fusion power divided by plasma heating power, of about Q=0.28. Neutral beams were subsequently used at the Joint European Torus (JET) near Oxford in the U.K. to achieve a higher fusion energy gain factor of about Q=0.65. Although these energy gain results are still far too low for pure fusion energy systems, the TCT approach is attractive for Fission Fusion Hybrid (FFH) schemes where the large energy multiplication ratio of the resulting fissions can underwrite the continuous investment of power fed back to a neutral beam plasma heating system. Today it remains the most plausible scheme for producing via fusion the 14.1 MeV neutrons needed for a hybrid nuclear reactor. A 1974 Lawrence Livermore Laboratory paper reported the results of neutronics simulations of various FFH blanket options using solid materials, including the graded use of various moderators. It predicted an optimized fission blanket power about ten times the DT fusion power with tritium breeding self-sufficiency in the blanket and also a net blanket production of plutonium-239 ranging, for different blanket options, from 2.24 to 4.51 atoms per DT fusion neutron. The report concluded that FFH technology could eliminate the need for isotopic enrichment and could use the then-existing national stockpile of depleted uranium for fuel, thus producing from this source alone a thousand years of electrical power for the US. This and subsequent published FFH studies envisioned a stand-alone fissile fuel factory which would produce and export plutonium for use in solid fuel rods to be fabricated for and fissioned in other reactors. In proposed FFH systems, fissions only occur in a subcritical fission blanket. In a recent (2009) Gaithersburg, Md. workshop organized by the Department of Energy titled “Research Needs for Fusion-Fission Hybrid Systems,” that limitation was elevated to become a definition: “A fusion-fission hybrid is defined as a subcritical nuclear reactor consisting of a fusion core surrounded by a fission blanket. The fusion core provides an independent source of neutrons, which allows the fission blanket to operate subcritically.” Disclosed herein are embodiments of a hybrid molten salt reactor advantageously incorporating a critical molten salt reactor with a source of energetic neutrons surrounded by a molten salt blanket in which fissions occur. Embodiments of the hybrid molten salt reactor (HMSR) may fission any combination of supplied actinides within a self-contained reactor system. Some embodiments can efficiently produce electrical power without requiring fuel enrichment, solid fuel fabrication, or spent fuel reprocessing and recycling, thus eliminating any need to transport concentrated fissile material. In some embodiments, where fission product inventories in the molten salt are limited to suitably low levels, actinides are entirely fissioned within the HMSR, thus fully exploiting the energy potentially available from actinides and removing them from the radioactive waste stream. In addition, certain long-lived fission products can be transmuted to shorter-lived isotopes within the HMSR without isotopic separation by simply not removing those chemical elements, while the quantity of others can be diminished by removing their chemical elements at optimized rates. In an example embodiment a hybrid molten salt reactor comprises a source of energetic neutrons, the energetic neutrons having a typical energy per neutron of 14 MeV or greater, a critical molten salt reactor comprising a reactor vessel, a loop comprising a path in the reactor vessel and around the source of energetic neutrons, and a molten salt, the molten salt containing a mixture of fissile actinides and fertile actinides, the molten salt circulating in the loop. The circulating molten salt has a sustained exothermic nuclear reaction comprising: (i) the fissile and fertile actinides irradiated by the energetic neutrons when exposed to the source of energetic neutrons, the energetic neutrons inducing subcritical nuclear fission and generating daughter neutrons, (ii) the fissile actinides in the circulating molten salt undergoing critical nuclear fission when circulating through the vessel of the critical molten salt reactor and generating daughter neutrons, and (iii) a portion of the fertile actinides capturing a portion of the daughter neutrons and, the captured daughter neutrons inducing transmutation of the portion of fertile actinides into fissile actinides. In some embodiments, the hybrid molten salt reactor further includes a blanket of tanks surrounding the source of energetic neutrons, with the molten salt circulating through the blanket of tanks, and a thickness and an arrangement of the blanket of tanks enabling an adequate fraction of the energetic neutrons to be absorbed in the molten salt to maintain a sufficient inventory of fissile actinides in the molten salt to maintain criticality of the critical molten salt reactor. The thickness and arrangement of the blanket of tanks enables the molten salt to absorb a sufficient portion of the energetic neutrons and generated daughter neutrons to maintain a desired fissile inventory. The blanket of tanks can be chemically and mechanically compatible with the molten salt, with each tank having separate plumbing connections for liquid inflow and outflow. The separate plumbing connections enable draining the tank based on gravity. The blanket of tanks can further include mechanical supports and plumbing connections configured for rapid replacement using remote handling equipment. The hybrid molten salt reactor can further include a controller adjusting an average power level of the source of energetic neutrons to maintain fission criticality in the critical molten salt reactor. The fission induced by the energetic neutrons and absorption of the resulting fission daughter neutrons by fertile actinides maintains fissile actinides in the molten salt at a concentration necessary for fission criticality in the molten salt reactor. The critical molten salt reactor can further include neutron absorbing control rods adapted to be partially inserted into the molten salt reactor in order to reduce a stable operating temperature of the molten salt reactor and adapted to be fully inserted into the molten salt to completely halt and preclude fission chain reactions. The hybrid molten salt reactor may include a fission product removal system enabling removal of one or more fission products from the molten salt. In some embodiments, the hybrid molten salt reactor includes a fuel system adding actinide salts to the molten salt at a rate which compensates for the loss by fission of the actinides previously dissolved in the molten salt. The fuel system can enable continuous fueling of the HMSR without stopping the critical nuclear fission chain reactions in the critical molten salt reactor. The hybrid molten salt reactor may include a fission product removal system enabling removal of one or more fission products from the molten salt, wherein the fission product removal system removes the fission products at a rate enabling indefinitely maintained fission in the critical molten salt reactor. In some embodiments, actinide fueling and fission product removal is sufficient to enable 100% fission energy utilization of the fertile actinides. The hybrid molten salt reactor may include a pump system to pump the molten salt around the source of energetic neutrons and through the vessel of the critical molten salt reactor. The hybrid molten salt reactor may include a heat exchanger receiving heat produced by the HMSR, wherein the pump system pumps the molten salt through the heat exchanger. The critical molten salt reactor may include a moderator. The moderator can be lithium hydride using the deuterium isotope of hydrogen and using lithium enriched in the lithium-7 isotope. The moderator can be a graphite core. The molten salt can contain lithium fluoride enriched in the lithium-7 isotope and sodium fluoride. The molten salt can contain fertile thorium-232 and the daughter neutrons transmute the fertile thorium-232 into fissile uranium-233. The molten salt can contain fertile uranium-238 and the daughter neutrons transmute the fertile uranium-238 into fissile plutonium-239. For either natural isotope, neutron absorption with gamma ray emission immediately transmutes the original nuclide to a far less stable isotope of the same element. Such neutron captures are most common with low energy incident neutrons but also occur with neutrons of higher energy. Uranium-239, produced by neutron absorption in natural uranium-238, then beta-decays with a 23.5 minute half-life into non-fissile neptunium-239, which in turn beta-decays with a 2.36 day half-life into the fissile plutonium-239 isotope, whose half-life is 24,100 years. Similarly, thorium-233, produced by neutron absorption in natural thorium-232, beta-decays with a 22.3 minute half-life into the non-fissile protoactinium-233 isotope, which in turn beta-decays with a 27.0 day half-life into the fissile uranium-233 isotope, whose half-life is 159,200 years. The source of energetic neutrons can be a nuclear fusion device producing the energetic neutrons from the fusion of hydrogen isotopes. The source of energetic neutrons can be a spallation device producing the energetic neutrons by impacting energetic ions on a target material. The molten salt can contain fertile thorium-232 and neutrons either produced as fission daughters or resulting from (n,2n) and/or (n,3n) reactions transmute it into fissile uranium-233 which then fissions. The molten salt can contain a mixture of one or more actinides either from spent nuclear fuel wastes of nuclear fission reactors or from any other source, wherein absorption of neutrons either produced as fission daughters or resulting from (n,2n) and/or (n,3n) reactions converts fertile actinides into fissile actinides which then fission. The molten salt can contain natural mined uranium that has not been isotopically enriched, the natural mined uranium containing fertile uranium-238 and fissile uranium-235, the fission daughter neutrons transmuting the fertile uranium-238 into fissile plutonium-239 and the fissile plutonium-239 undergoing fission with the uranium-235. The molten salt can contain fertile uranium-238 or depleted uranium which by definition is primarily uranium-238 and neutrons either produced as fission daughters or resulting from (n,2n) and/or (n,3n) reactions transmute it into fissile plutonium-239 which then fissions. Another example embodiment of the HMSR combines a critical Molten Salt Reactor (MSR) with an external source of high energy neutrons by routing the molten salt which flows in a loop through the MSR's reactor core, through the MSR's external heat exchanger and circulating pump, to also flow through a blanket of tanks adjacent to the external neutron source, before returning in the single loop to the MSR's core. The adjacent blanket of tanks surrounds enough of the external neutron source so that most neutrons from the external neutron source enter the tanks and are absorbed by actinides carried by molten salt flowing through those tanks. The embodiment of the HMSR also includes a control scheme which adjusts the power level of the external source of high energy neutrons so as to maintain fissile actinides in the molten salt at concentrations which provide the MSR's necessary reactivity. An embodiment, in which MSR power is separately controlled to follow variations in the aggregate total external load demand for electric power, adjusts the time-averaged high energy neutron source power to be a fraction of the MSR power where that fraction is itself continuously increased or decreased based on comparison of measurements of MSR reactivity with a fixed reactivity setpoint. Past hybrid reactor design concepts have not in general included a critical reactor. In particular, they have not included critical molten salt reactors such that dissolved nuclear fuels would flow internally around a single shared loop including both the critical molten salt reactor and a separate zone for irradiation by high energy neutrons. The high energy neutron source may either be a nuclear fusion reactor device producing 14.1 MeV neutrons from the fusion of hydrogen isotopes, or it may be an accelerator-driven spallation device producing neutrons of that or even higher energies. The molten salt reactor (MSR) is a design in which the fuel is liquid and mobile. MSR designs employ liquid salt mixtures including dissolved fissile fuel nuclides and frequently also fertile nuclides such as thorium-232 or uranium-238. The fractions of different mixture components are usually chosen near eutectic points of phase diagrams so that the mixture's melting point, which typically is well above room temperature, is minimized and so that any unexpected salt freeze-up event would produce uniform solidification without separating into dissimilar mixture components. Since fluorine is the strongest possible oxidizer for chemically binding metals, most MSR designs have adopted fluorine-based molten salts such as UF4, PuF3 and ThF4 dissolved in a carrier liquid mixture of NaF, ZrF4, LiF and/or BeF2. Other proposed MSR designs used chloride-based carrier salts (e.g., NaCl) mixed with uranium chloride and plutonium chloride in order to achieve a harder fission neutron energy spectrum and somewhat lower melting temperatures. Historically, the first MSR built and operated was the Aircraft Reactor Experiment (ARE) which operated at Oak Ridge National Laboratory (ORNL) in 1954 (without ever flying). Later, the 8 MW Molten Salt Reactor Experiment (MSRE) at ORNL operated more than 17,000 hours from 1965 through 1969, testing fission power operations with molten salt mixtures employing successively uranium-235, uranium-233, and then trace plutonium-239 as fissile fuel components. A detailed design of a thorium to uranium-233 breeder reactor employing the MSR concept, the Molten Salt Breeder Reactor (MSBR), was completed at ORNL in the early 1970s but never built. However, since the 1970s, engineering interest in the MSR has continued. The MSR class of designs has been adopted internationally as one of the Generation IV reactor design families chosen in the past decade to be developed for possible future use, and there is substantial technical research interest in MSR's within the international nuclear engineering community. The initial attraction of using molten fuel was its very strong negative temperature coefficient that results when thermal expansion causes liquid fuel to leave a moderated reactor core region. This can result in reactor power adjusting itself naturally to follow load demand variations without active feedback control, and it also is a safety feature. The most striking characteristic of MSRs is that their nuclear fuels are liquid and thus can be made to flow. This confers several advantages as follows: (i) A strongly negative temperature coefficient enhances stability and safety. (ii) Fuel can be moved by gravity from the reactor core to passively cooled dump tanks in an emergency. One simple design passively initiates a dump if a solidified salt plug melts. (iii) No solid fuel needs to be fabricated. (iv) Fission products can be continuously removed and make-up fuel added while operating. (v) A low radioactivity source term for accidental releases can be achieved by the continuous removal of fission products from the molten salt, thus maintaining a low fission product inventory in the MSR. (vi) Low reactivity margins for criticality optimized designs are feasible, thus reducing the extent of worst-case possible criticality excursions. (vii) External cooling becomes possible because of fuel flow. Heat can be removed in external heat exchangers located outside the critical core region, and thus away from where the fission chain reaction occurs, instead of using space within the reactor core for heat transfer to a coolant as is required for all solid fuel designs. This allows a more compact fission core design in a MSR. (viii) Delicate solid fission fuel and cladding structures vulnerable to meltdown damage in light water reactors (LWRs) are entirely eliminated. (ix) Life-limiting damage to solid fuel caused by fissions which increase the number of atoms and their volume in the solid is eliminated since ionic liquids have no molecular or crystalline structure to damage. These attributes are thought to simplify the safety situation for a MSR design. Indeed, MSRs are not subject to core meltdown accidents since, tautologically, their fuel is already melted during normal operation. MSR salts are chemically stable so they cannot burn or explode. Furthermore, an MSR thermal conversion cycle using helium in a closed Brayton cycle instead of water/steam would also avoid any steam explosion hazard. FIG. 1 illustrates a schematic system diagram of an embodiment of a hybrid molten salt reactor. This depicted embodiment is intended for use in a stationary location since gravity affects some of its features. FIG. 1 shows a single closed loop of pipes 101 connecting the following four components: a blanket of tanks 110, which surrounds an energetic neutron source 111 adjacent to but not part of the loop 101, a molten salt reactor 120, a molten salt heat exchanger 130, and a molten salt circulating pump 140. The loop of interconnecting pipes 101 and associated components 110, 120, 130, 140 is oriented with respect to the vertical direction so that the molten salt heat exchanger 130 is located at the highest elevation in the loop 101 and, in particular, is located above the molten salt reactor 120 to which it is directly connected. An additional pipe not part of the loop 101 extends upward from the highest elevation in the loop 101 to a closed pressurizer volume 105 located above the loop. Another additional pipe not part of the loop extends downward from the lowest elevation in the loop 101, through a salt freeze plug 109 to a set of dump tanks 199 located below the loop 101. The total aggregate volume of the dump tanks 199 equals or exceeds the closed volume for molten salt 10 above the dump tanks 199. A molten salt liquid mixture 10 of different ionic salt components including actinides fills the entire loop 101 and extends upwards above the loop into the lower part of a closed pressurizer volume 105 which serves to accommodate expansion and contraction of the molten salt 10, e.g., as a result of molten salt 10 temperature changes over time. Gas in contact with the molten salt fills the upper part of the pressurizer volume 105. The molten salt mixture 10 also extends downwards to the salt freeze plug 109 location where deliberate external heat leakage mechanisms, not shown in FIG. 1, cause the salt temperature to stay below the salt's freezing/melting temperature, thus plugging the downward path within the piping 101 by solid salt 109 so that during normal operation molten salt 10 stays in the loop 101 and does not drain into the dump tanks 199. Also not shown is the system to restore the salt freeze plug 109 and molten salt 10 from the dump tanks 199 back to the main loop 101 in order to resume normal operation. A particular embodiment of the HMSR based on fluorides which is compatible with FIG. 1 is the eutectic mixture of about 44.5 mole % LiF, 24.1 mole % NaF, and 31.4 mole % of the fluorides of actinides such as ThF4 or UF4. Such mixtures are liquid between approximately 490° C. to well above 1000° C. and so could be used with loop temperatures in the 600° C. to 700° C. range. Each actinide can assume multiple chemical combinations with fluorine, e.g., uranium can form UF3, UF4, and UF6. To control their relative abundances in the molten salt a redox chemistry gas control system 173 of FIG. 1 can be configured injecting gases at a low point into the molten salt, letting their bubbles ascend due to their buoyancy, and withdrawing them from the pressurizer volume. For instance, bubbling fluorine gas through the molten salt based on fluorides decreases the concentration of UF3, while increasing concentrations of UF4 and UF6, while bubbling hydrogen gas through the molten salt thus forming HF gas in the pressurizer decreases the UF6 concentration while increasing concentrations of UF3 and UF4. Continuing to refer to FIG. 1, with a molten salt 10 mixture of fluorides, fresh actinide fuels 11 can be most conveniently added to the loop 101 as fluorides with an actinide addition system 172. Much of the existing stockpile of depleted uranium is at present stored as UF6 and natural uranium is typically converted to UF6 gas before being enriched. However, since fluorine is the strongest oxidizer it is not difficult to prepare actinide fluorides from any other chemical form. If spent nuclear fuel from LWRs, which is mostly UO2, is pulverized then heated in a fluorine atmosphere, it burns to form uranium fluoride compounds while releasing oxygen. Gaseous fission products such as xenon tend to eventually rise to and collect within the pressurizer's gas volume 105 to be processed by a gas removal system 171, but this process can be greatly accelerated by sparging. In the sparging process, a gas, preferably helium, is injected at a low point into the molten salt and its many bubbles rise up to the pressurizer gas volume from which they are continuously removed. The large surface area of gas bubbles rising through the molten salt 101 increases the rate at which fission product gas atoms are combined into bubbles which are rapidly removed from the molten salt due to their buoyancy. Some fission products 12 such as the noble metals do not form soluble fluorides and may be processed by a fission product removal system 174, which may mechanically remove from a secondary loop by filtration or centrifugal separation. Others may be removed from a secondary loop by distillation. Electrochemical means provide ways to remove still other fission products from the molten salt. An alternative embodiment of the HMSR system using chloride salts is possible and can provide a salt mixture with chemical bonds almost as strong as with fluorine but having a lower melting/freezing temperature than fluorides. However, it would not then be possible for the HMSR to avoid transmuting both natural chlorine-35 and natural chlorine-37 into radioactive chlorine-36, which has a half-life of 301,000 years. Alternative embodiments based on bromide or iodide salts are also possible and would also result in lower melting/freezing temperatures without producing any long half-life radioactive product but would have somewhat weaker chemical bonds than fluorine or chlorine and would need to contend with the complicating fact that both bromine and iodine are also fission products themselves. Neither fluorine nor chlorine is a fission product. During normal operation of the HMSR system the loop's 101 molten salt 10 is in steady rapid motion, circulating around the loop 101 because of the pumping action of the molten salt circulating pump 140 which itself is driven by an electric motor 141. Molten salt pumps 140 successfully used in the past have been of the vertical shaft type. Molten salt 10 pumped to elevated pressure exits the molten salt pump 140 in the downward direction, flowing through piping 101 until reaching the blanket tanks 110 which surround the energetic neutron source 111, flowing through the blanket tanks 110 and then on through piping 101 to enter the lower plenum of the molten salt reactor 120, then flowing upwards through the reactor core's 121 multiple parallel spaced channels in a moderator, which in the FIG. 1 depicted embodiment is a graphite moderator, then through an upper molten salt reactor 120 plenum and then loop piping 101, to and through the molten salt heat exchanger 130, then at lower pressure back through piping 101 to the molten salt circulating pump's 140 intake, thus completing the loop. During normal operation, a steady neutron flux maintains itself by critical fission chain reaction mechanisms in the molten salt reactor's 120 core region 121, causing fissions of fissile atoms to occur while the molten salt 10 containing those fissioning atoms within the liquid is flowing through the reactor core's 121 multiple spaced channels through the moderator. Most fissions occur in this spatial core region 121 because the moderator material filling the spaces between flow channels scatters and thereby slows fission daughter neutrons passing through it from their initial fission spectrum energies typically near 1 MeV per neutron to lower energies approaching the thermal range, e.g., 0.025 eV to 1 eV. The physics of fission includes the fact that for fissile atoms the fission cross sections which together with neutron flux determine fission rates are greatly increased if the initiating incident neutrons have low energy. These fissions occurring in the molten salt 10 as it flows through the molten salt reactor 120 directly heat the molten salt 101 liquid volumetrically without requiring any heat transfer through surfaces. Thus, molten salt reactor power density is not restricted by fuel-to-coolant heat transfer limits of the reactor core's 121 design, as is the case for all solid fueled fission reactors. Molten salt 10 exits the top of the molten salt reactor 120 at a higher temperature than its temperature when it entered at the reactor's bottom, because of the heat immediately released by fissions in the core 121. Most types of fission daughter atoms have a ratio of neutrons to protons in their nuclei too large for stability, so they are temporarily radioactive, typically emitting beta electrons and gamma rays with half-lives ranging from fractions of a second to much longer. During normal steady fission operations these decay processes contribute about 7% of total power, and they also cause direct volumetric heating of the molten salt distributed around the entire molten salt loop. The molten salt heat exchanger 130 contains two flowing fluids, the molten salt 10 and an intermediate heat transfer fluid 20, not containing any actinides, in a secondary loop 1602. There are multiple possible intermediate heat transfer fluids 20 possible, each compatible with the embodiment depiction of FIG. 1. The heat exchanger 130 keeps the two fluids 10, 20 separated from each other by a compatible solid membrane through which heat flows via thermal conduction from the molten salt 10 on the higher temperature side of the membrane to the intermediate fluid 20 on the lower temperature side. This heat exchanger 130 thus serves to export the high temperature heat produced in the reactor 121 without exporting radioactivity. Heat exchanger technology is a well-known art, and standard heat exchanger parameters determine relationships between flow rates of the two fluids, their incoming and outgoing temperatures, their heat transfer rates, and their flow pressure drops. By adopting a counterflow scheme and increasing the membrane's total surface area, completely efficient heat transfer can be approached arbitrarily closely, albeit with increasing heat exchanger cost. Alternatives to the FIG. 1 embodiment may optionally include more than one intermediate heat transfer 20 fluid with high temperature heat transferred from the molten salt 10 to the first fluid, then through another heat exchanger (not shown) to the second fluid, etc., before exporting the high temperature heat for external uses such as electric power production. Although such features would increase HMSR cost they would decrease the chance of releasing radioactivity because of a physical failure of a single heat exchanger membrane. Continuing to refer to FIG. 1, the high temperature heat transferred through the molten salt heat exchanger 130 to the intermediate heat transfer 20 fluid is further transferred to a thermal conversion system 150 which converts its energy content into a combination of mechanical work in the form of rotating shaft torque and low temperature heat. The mechanical work in turn operates a generator 170 producing electrical power while the low temperature heat is exhausted into an adjacent external heat sink, either the atmosphere or a very large body of water, from which in either case the low temperature heat ultimately radiates to space. Different conventional thermal conversion schemes are possible, each having its own advantages. The particular scheme depicted in the FIG. 1 embodiment is an open Brayton cycle in which air is first compressed by a compressor 151, then heated by a second heat exchanger 152 transferring to the air high temperature heat from the intermediate heat transfer fluid 20, then is expanded through a turbine 153 mounted on the same rotating shaft 154 as the compressor 151, and finally is exhausted back to the atmosphere at a low temperature which is warmer than the air intake temperature, thus carrying away the low temperature exhaust heat into the atmosphere without requiring any low temperature heat exchanger equipment to transfer heat into the air. Turbine torque exceeds compressor torque, providing the net mechanical power used to operate the electrical generator 160 which is also mounted on the same shaft 154. Although the depicted open Brayton cycle has the advantage of low capital cost, other conversion schemes such as closed Brayton cycles and Rankine cycles can obtain higher thermal conversion efficiencies. Thus, as with the design selection of heat exchanger parameters, cost-benefit trade-off studies should guide the selection of the thermal conversion system from the range of conventional options. Normal operation of any molten salt reactor 120 includes an inherently stable load-following characteristic. If thermal conversion power increases to satisfy a suddenly increased electrical demand, suddenly exceeding the rate at which heat is being transferred into the intermediate heat transfer fluid 20 from the molten salt 10, then the average temperature of the entire intermediate loop 160 will lower as more heat is withdrawn from it than is being added. The reduction in temperature of the intermediate loop 160 will in turn cause an increased cooling of the molten salt loop 101, causing its temperature to also decrease as long as fission power remains constant. However, the reduced molten salt 10 temperature causes reactivity to increase via two mechanisms. Reduction of effective resonance cross sections because of the Doppler coefficient increases the fraction of neutrons, which avoid capture and reach low energy where they are effective in causing fissions. Thermal contraction of the molten salt 10 as its temperature is reduced removes some molten salt from the pressurizer's expansion volume 105 and increases the density of molten salt 10 in the molten salt reactor's 120 core region 121 and thereby the amount of fissile material there. The reactivity increase from both effects causes reactor power to increase at a slow exponential rate set by the net positive reactivity and the delayed neutron fraction of the fuel mixture. Reactor power then increases until the net reactivity returns to zero, which does not occur until the average molten salt 10 temperature in the reactor 120 returns to the same temperature that prevailed before the electrical load demand increase. At that time, the reactor power will be at a higher value than before, balancing the increased load demand. A similar sequence for decreases in load demand illustrates this load following behavior, with increases in the temperature of the intermediate fluid 20 and the molten salt 10, increases in resonant absorption of neutrons and decreases in the fissile material within the core region 121, followed by stabilization at a reduced fission power level but the same molten salt 10 temperature. Thus, the inherent self-regulating characteristic of a molten salt reactor 120 tends to keep its temperature constant regardless of power demand. Not shown in FIG. 1, some electric power produced by the generator 160 must be fed back with a controller 180 to operate: (a) the energetic neutron source 111, (b) the molten salt circulating pump 140, (c) the intermediate fluid pump 161, and (d) other conventional auxiliary molten salt reactor support systems (not shown), including electrical heaters and dump tank pumps. The temperature at which a molten salt reactor stabilizes is not constant but instead varies with its excess reactivity, which in turn depends on its concentration of fissile atoms and its concentration of neutron-absorbing species, in particular of certain fission products such as xenon-135. It also depends on control rod position (not depicted in FIG. 1). In a conventional MSR, neutron-absorbing control rods can be initially inserted into the molten salt, then gradually withdrawn as the fissile inventory and thus the excess reactivity are reduced by cumulative fission. Here, the control rods' motion has the effect of maintaining the natural stabilization temperature of the molten salt in its inherent load-following behavior at a constant value, but that effect only lasts until the control rods reach their fully withdrawn positions as the initial excess reactivity disappears. Although fertile material is converted to fissile material within a MSR by neutron absorptions followed in some cases by beta decays, for Conversion Rate (CR) values less than one these only slow but do not prevent the net consumption of fissile isotopes. The constant-temperature behavior can be extended, in principle indefinitely, by continuously removing neutron-absorbing fission product “ash” from the molten salt while also adding fresh fissile material to the molten salt at rates compensating for the fission rate. However, most available actinides are not fissile so this scheme would require isotopic enrichment and would not fully utilize the energy potentially available in natural actinides that are not fissile. In embodiments of the HMSR, additional net conversion of fertile actinide isotopes to fissile isotopes is provided by an energetic neutron source (111 in FIG. 1), which may be either an accelerator-driven spallation neutron source or a deuterium-tritium (DT) fusion neutron source. For example, using the previously described eutectic mixture of fluorides consistent with the FIG. 1 embodiment, neutronics simulations show that an 80 cm thick molten salt blanket surrounding the neutron source is almost as effective as an infinitely thick blanket. This is because more than 99% of total neutrons are absorbed within a 80 cm thick blanket, where the total includes the 14 MeV source neutrons, fission daughter neutrons produced in the blanket, and (n,2n) neutrons produced in the blanket. For such a 80 cm thick blanket holding this mixture of fluoride salts with the actinide component predominately uranium-238, simulations show that each incoming 14 MeV neutron results in 0.413 fissions in the blanket, mostly of uranium-238, releasing 1.483 fission daughter neutrons thus averaging about 3.6 daughter neutrons per fission. Additionally about 0.243 more neutrons per 14 MeV neutron are released by (n,2n) reactions. About 2.28 neutrons per incoming 14 MeV neutron are absorbed in uranium-238, producing 2.28 atoms of uranium-239 which after two beta decay steps becomes 2.29 fissile plutonium-239 atoms. If the molten salt reactor 120 exhibits a Conversion Ratio of between CR=0.90 and CR=0.95 then the quantity of fissile material exported from the blanket per DT neutron is effectively multiplied by a conversion factor ranging between 10 and 20, while being fissioned. Thus, the number of fertile atoms converted to fissile atoms and fissioned as a result of a single 14 MeV neutron and the action of the HMSR system is in the range from 22.8 to 45.6. In terms of energy released, each single 14 MeV energetic neutron irradiating the blanket of tanks surrounding the Energetic neutron source leads to about 83 MeV of additional fission energy being released immediately in that blanket plus eventually another 4560 MeV to 9120 MeV in the molten salt reactor from fissions of initially non-fissile actinides. Thus in a HMSR the power carried by neutrons in the energetic neutron source can be a small fraction of the total plant power released by fissioning initially non-fissile actinides, i.e., less than one percent. In embodiments with neither fission product removal nor actinide addition implemented, the HMSR's energetic neutron irradiation of the molten salt still provides the benefit of extending operating period durations between refueling outages by increasing the system's conversion from fertile to fissile actinides. With embodiments in which fission product removal and fertile actinide addition are both effective enough to continually maintain MSR criticality, it becomes possible to never recycle the molten salt. The strategy then becomes to never remove actinides from the molten salt and to never remove the molten salt containing its inventory of actinides from the HMSR. Simulations show this strategy is effective in that the actinide inventory eventually stabilizes even though actinides continue to be added. An embodiment of the HMSR would initiate operation with the particular mixture of different actinides that does not change over time, while a related feedstock of mostly or completely fertile actinides is continuously added and while fission power is produced and converted into electricity. An attractive aspect of this embodiment is that simulations show that such stationary mixtures have all fissile components of the molten salt denatured by larger concentrations of non-fissile isotopes of the same elements. For instance, such a stationary mixture of plutonium isotopes produced from uranium is 60% non-fissile plutonium-242, 21% non-fissile plutonium-240, and 1% non-fissile plutonium-238, with only 12% as fissile plutonium-239 and 6% as fissile plutonium 241. This denatured situation reduces weapons proliferation risks since a difficult isotopic separation step would be required to create purified fissile material from this mixture. The strategy of never removing actinides from the molten salt can be extended to also not remove certain long-lived fission products from the molten salt. However, a difficulty arises because the choice to not remove a fission product applies to all isotopes of an element. This is because it is not expected to be economic to treat different isotopes of a fission product element differently from each other. Simulations show that this non-removal strategy is quite successful for long-lived fission product isotopes of cesium, iodine and technetium. For some other long-lived fission products this strategy fails due to the unending build-up of stable fission product isotopes of the same elements. For these elements a different strategy of removal at an optimized rate could still provide the benefit of somewhat reducing the quantity of long-lived isotopes that would either need disposal in a geological repository or isotopic separation followed by transmutation. The maintenance of MSR criticality requires that the conversion of fertile to fissile isotopes effected by the energetic neutron source must occur at a rate which, on average, counterbalances the consumption by fission of fissile isotopes. This balance is accomplished by a control system whose commands modulate the rate of production of the energetic neutrons. Such modulation does not need to be continuous, and may even be accomplished by repetitively switching the neutron source off thus producing zero neutrons and then on at a maximum neutron production rate, while adjusting the timing of those switching transitions. Feedback measurements needed to implement this control system include molten salt temperatures and flow rates, neutron flux in the MSR, and MSR control rod positions. Other measurements may also be included if found useful. If an embodiment's energetic neutron source is based on DT fusion rather than spallation, then it is also necessary for the system to breed tritium which is accomplished by using a molten salt recipe that includes lithium having an isotopic composition ratio, lithium-6 to lithium-7, chosen to cause the needed amount of tritium production. Although in a pure DT fusion system the requirement of breeding one replacement tritium atom for each DT fusion neutron released is anticipated to be a difficult constraint to be met through neutron-consuming reactions with lithium, the situation is different in a hybrid where fission further multiplies the neutrons. In the HMSR configuration, most of the needed tritium breeding would occur in the MSR where low energy neutrons are plentiful. The minimum amount of tritium production occurs if the lithium used is pure lithium-7, in which case only high energy neutrons participate. As the lithium-6 concentration is increased, tritium production monotonically increases. Tritium is recovered from the molten salt along with other hydrogen isotopes such as protium which is sometimes injected to reduce the content of a halogen species such as fluorine for redox control, then is separated by conventional means such as cryogenic distillation. Since hydrogen isotopes diffuse through hot metal walls, it is important to forecast tritium leakage as part of the design. If leakage is excessive, it may be necessary to use double walled containment with helium circulating between the walls, undergoing continuous chemical scrubbing to keep its tritium partial pressure low. The physics advantages of using a high energy neutron source include the ability to fission any actinide, fissile or not, and its larger yield of daughter neutrons which can cause transmutation of fertile actinides to fissile actinides. The energetic neutrons from the external source carrying 14.1 MeV or more energy per neutron can cause fission of any actinide nucleus that they strike, and when such fissions caused by such high energy neutrons occur, the number of fission daughter neutrons produced per fission is significantly greater than the number of daughter neutrons produced by fissions caused by lower energy neutrons. If each such daughter neutron is absorbed by a fertile actinide atomic nucleus in the molten salt, that nucleus is either converted immediately into a fissile actinide or it starts a decay process which after some days have elapsed converts it into a fissile nuclide. In either case, the nuclide remains in solution within the HMSR's molten salt, cycling back and forth between the critical MSR and the external neutron source. When the MSR eventually fissions the new fissile actinide atoms created by the nuclear processes initiated by neutron irradiation of a blanket tank, some of the daughter neutrons released by those fissions are absorbed within the MSR region, also by fertile actinide isotopes dissolved in the molten salt, producing an additional fraction, (CR), of the fissile atoms from the blanket tank. Those are also converted to fissile species and fissioned, releasing yet more daughter neutrons, producing an additional fraction, (CR)2 of those fissile atoms. Those in turn produce an additional fraction, (CR)3 which in turn produce yet another fraction (CR)4. This thus forms an infinite series of recursive logical steps represented by a geometric series. This geometric series is equivalent to multiplying the externally caused production of fissile species in the blanket of tanks by the factor, 1/(1−CR), where CR is the conversion ratio of the critical reactor. For instance, a graphite-moderated MSR with a conversion ratio of CR=0.95 provides a 20-fold increase in the effectiveness of the external neutron source in exploiting the more abundant actinide isotopes. Fission cross sections of fissile atoms are very much larger for slow neutrons than for neutrons at the higher energies at which daughter neutrons resulting from fission are born. Neutrons slow down, i.e., are moderated, through scattering collisions with atomic nuclei. The use of a special moderator material in a critical reactor configuration can therefore allow the reactor to be smaller or to incorporate less fissile material than is possible without a moderator. Thus, most critical reactor designs include moderators, but some do not. On the other hand, materials also absorb neutrons, a behavior that is undesired in a moderator. For example, because of its hydrogen content water is an excellent moderator which facilitates physically small reactor designs. However, it absorbs so many neutrons that reactor designs using water as a moderator cannot sustain a critical chain reaction using natural uranium, instead requiring uranium enriched in its fissile isotope. Heavy water incorporating deuterium as a moderator is somewhat less effective so requires a larger moderator volume but its lower neutron absorption allows critical reactor designs using natural uranium. Other compounds incorporating deuterium may function similarly, e.g., lithium hydride using the deuterium isotope of hydrogen and lithium enriched in the lithium-7 isotope. Pure graphite, which has been used as a moderator in MSRs, requires even more volume to slow neutrons but similarly absorbs few neutrons so has been used to moderate reactors fueled with natural uranium. Without the past six decades of research, there would be no fusion neutron sources, the spallation neutron sources would be embryonic, and there would be no MSR operating experience nor even a data base of molten salt properties. Most important for modern nuclear engineering is the vast information content of Evaluated Nuclear Data Files (ENDF), which can be used with suitable neutronics computer codes to design and evaluate nuclear energy systems without actually constructing and experimentally operating them. The use of such codes and ENDF or similar nuclear data files is essential to predicting the interplay of neutron elastic and inelastic scattering, absorption, fissions, fission daughter neutrons, (n,2n) and higher neutron releasing reactions. Although multiple molten salt formulations are possible and may have significant advantages, a particular example starting recipe for an embodiment of the HMSR based on fluorides is the eutectic mixture of about 44.5 mole % LiF, 24.1 mole % NaF, and 31.4 mole % of fluorides of actinides such as ThF4 or UF4. Stated in atom % units these are equivalent to about 66 atom % fluorine, 15 atom % lithium, 11 atom % actinides and 8 atom % sodium. Stated in mass % (or weight %) these are equivalent to about 62 mass % actinides, 31 mass % fluorine, 5 mass % sodium and 2 mass % lithium. This particular mixture recipe melts uniformly at a temperature below 500° C. and remains a liquid at temperatures well above 1000° C. Importantly, the special metal alloy, Hastelloy-N, developed for its long-term high-temperature chemical compatibility with fluoride-based molten salts in the MSRE radiation environment, is usable for steady molten salt temperatures up to 706° C. with brief excursions to higher temperatures. Thus, with this particular embodiment's molten salt recipe and with existing metallic alloy materials, it is feasible to operate a thermal conversion cycle for electricity production within a temperature range between the mixture's freezing/melting point temperature and the temperature rating of existing compatible materials, e.g., 600° C. to 700° C. Alternatively, if and when graphite container materials become available, the operating molten salt temperature of a different embodiment can be elevated above 1000° C. Such high temperatures would allow electricity generation at higher efficiency and lower cost and may also make new applications practical such as the production of hydrogen fuel from water. Neutronics simulations of neutron irradiation of this particular molten salt recipe using uranium-238 and lithium-7 show that blanket tanks with a molten salt thickness of 80 cm are able to absorb over 99% of neutrons including together the 14 MeV source neutrons, the fission daughter neutrons, and neutrons produced by other reactions such as (n,2n), (n,3n), etc. Thus, in a commercial HMSR the neutron source blanket tanks would not need to be thicker than 80 cm and may have somewhat less thickness if design optimization finds that advantageous. Simulated fission power in these subcritical blanket tanks was found to be between five and six times the DT fusion power. This is less than the factor of ten predicted in historical FFH studies, but those studies did not include fluorine, sodium and lithium, which provide as their main benefit fuel mobility. Neutronics simulations of an embodiment of the entire HMSR system predict that the ratio of MSR power to energetic neutron source power exceeds one hundred. This result was robust; the neutron source power was less than 1% of overall HMSR plant power regardless of whether the actinide fuel feedstock was uranium-238, thorium-232, depleted uranium, or the mix of actinides found in LWR spent fuel rods. This result implies that a certain level of energy inefficiency in the neutron source can be tolerated in an economically successful HMSR system. For instance, if 30% of HMSR plant power were thermally converted at 33% efficiency to electricity and fed back to operate a neutron source which itself is only 10% efficient, the resulting energetic neutrons comprising 1% of output power would suffice to maintain fissile fuel for 100% of output power of which 70% could be converted to yield 23% of output power as electricity for sale. An embodiment of the HMSR confers additional benefits if fission products are continuously removed from the molten salt fast enough that criticality is maintained. Chemical engineering technology is able to separate any chemicals. At issue is whether separation processes which are both economical and compatible with continuing MSR operations can be developed for all fission products. If so then it will never be necessary to shut down for “ash removal” and the molten salt can be used indefinitely without being removed. The resulting new embodiment leaves actinides dissolved in the molten salt until they fission and provides no systems for actinide removal. This HMSR embodiment was simulated in neutronics and transmutation codes; they predict that actinide inventories eventually stabilize while fresh actinides continue to be added and fission products without actinides are actively removed. Thus, this embodiment makes it possible to entirely remove actinides with their long radioactive half-lives from the waste stream while harvesting 100% of their fission energy. Embodiments of the HMSR also mitigate proliferation concerns associated with the LWRs fuel cycle. HMSR fuel needs no enrichment, no fabrication and no reprocessing. In its best-practice implementation where the HMSR operates without interruption, interest by proliferators in its waste stream would be greatly reduced by the complete absence of any actinides there. An HMSR neither imports nor exports enriched fissile fuel, and it consumes its fissile fuel internally as fast as the fissile fuel is created, thus maintaining its low inventory. There would be no reason for an HMSR to have equipment on hand to remove any dissolved actinides, fissile or not. Additionally, simulations show that HMSR inventories evolve towards states where each fissile isotope is denatured by non-fissile isotopes of the same element, thus making futile their diversion for explosive weapons uses without additional isotopic enrichment. HMSR simulations also showed that some long-lived fission products can be internally transmuted so that, like actinides, they also are absent from the HMSR waste stream. Thus, embodiments of the HMSR can greatly reduce the need for long term radioactive waste storage in a geological repository. HMSR Nuclear Process Analyses It is conventional practice in modern nuclear engineering to use computer-based modeling to the extent feasible. Similar to trends in other engineering disciplines, this alternative to the construction, operation, data measurement, acquisition and interpretation of physical experiments minimizes potential costs and risks, especially during technical development. Most significant nuclear reactions have by now been extensively studied. Committees reviewing published reaction rate measurements have periodically chosen consensus models for Evaluated Nuclear Data File (ENDF) cross section libraries. These ENDF database libraries, available from the National Nuclear Data Center at Brookhaven National Laboratory, can accurately predict most nuclear reaction rates in a real physical system if used with an appropriate nuclear computer code and system model. Many such nuclear codes exist for different applications, most developed at national laboratories of the US Government's Department of Energy. A computer study was done to establish how well a particular embodiment of the HMSR would perform in consuming all supplied actinides if neutron losses were realistically minimized. It is summarized here in order to assist persons skilled in the art to use this invention without undue experimentation. The system of nuclear codes known as SCALE developed at Oak Ridge National Laboratory (ORNL) was used. ORNL's reference for the SCALE code package used is as follows: Scale: A Comprehensive Modeling and Simulation Suite for Nuclear Safety Analysis and Design. Version 6.1, June 2011. ORNL/TM-2005/39. Available from Radiation Safety Information Computational Center at Oak Ridge National Laboratory as CCC-785. Modules from the SCALE 6.1 system of codes were used along with the 238-group ENDF/B-VII Release 0 cross section library. It was necessary to develop additional interfacing code modules since the HMSR system with its two distinct nuclear reaction zones does not conform to any conventional configurations for which SCALE analyses have been automated. The Energetic Neutron Source was modeled as an isotropic uniform density volumetric spherical source of 3.5 m radius emitting neutrons in group 4 which runs from 13.84 MeV to 14.55 MeV. To guarantee ample moderation the MSR's graphite was modeled as a matrix of vertically oriented 15 cm/side prismatic hexagonal blocks with 3.5 cm diameter central molten salt channels. Overall cylindrical MSR size was set to 8.8 m for both diameter and height. Although larger than typical modern power reactors this is similar to the size of the graphite moderated reactors which operated at Hanford during the 1940s. The molten salt mixtures modeled were 44.5 mole % lithium fluoride (LiF), 24.1 mole % sodium fluoride (NaF) and 31.4 mole % total (HM)Fx where HM (Heavy Metal) represents actinide species and x ranges from 4 for thorium through uranium to 3 for plutonium and higher. Initial computer runs were made using SCALE's XSDRNPM code module in order to choose the thickness of the molten salt blanket surrounding the source. It was decided this should be functionally equivalent in terms of leakage to an infinite thickness. The goal was set that neutron leakage be between 0.5% and 1.0% of the total of (1) the energetic source neutrons plus (2) the net additional neutrons produced within the blanket by (n,2n) or (n,3n) reactions or by fission. With the molten salt actinide content set to be entirely uranium-238, an 80 cm thickness yielded the required leakage. This blanket thickness was adopted for all models. Predicted neutron reactions in this molten salt blanket carrying uranium-238 are summarized in Table 1. For each energetic source neutron, 0.21872 fissions of uranum-238 occur releasing about 43 MeV of fission energy plus (0.82170−0.21872=0.60298) additional daughter neutrons beyond those consumed to initiate the fissions. There are also additional neutrons released by (n,2n) and (n,3n) reactions, totaling 0.00557+0.02632+0.00197+0.12613+2*(0.03793)=0.23585. Of the net total 1.83883 neutrons, 1.66253 are captured by other uranium-238 nuclides converting them into uranium-239 which beta decays into fissile plutonium-239. TABLE 1Calculated Neutron Reactions in 80 cm Thick Blanket Containing U238 asSole ActinideReactions per 14.1 MeVNuclide insourceMolten SaltReactionneutronlithium-7(n, 2n)0.00557(n, 3n)0(n, γ)0.00067fluorine-19(n, 2n)0.02632(n, 3n)0(n, γ)0.00918sodium-23(n, 2n)0.00197(n, 3n)0(n, γ)0.01053uranium-(n, 2n)0.12613238(n, 3n)0.03793fissions0.21872fissiondaughters0.82170(n, γ)1.66253 The ORIGEN module of SCALE6.1 was then used in subsequent runs to simulate evolution of the molten salt's isotope inventory caused by reactions both in the MSR and in the blanket. ORIGEN's inputs include single-group collapsed cross sections, neutron flux, exposure duration, steady continuous removal rates (sec-1) for each element, and continuous addition rates for a set of fueling isotopes. Simulated powers of the MSR vs. the neutron source were adjusted to maintain keff=1 criticality using the facts that non-breeding MSR operation depletes fissile inventories, while operation of the energetic neutron source increases fissile inventories. For this adjustment, a software feedback loop functioning as a switching controller first evaluated keff criticality for the MSR based on the molten salt's most recent evolved isotope inventory, then either ran ORIGEN for the blanket if keff<1 or for the MSR if keff≧1. Over successive loop iterations the average power ratio adjusted itself to maintain keff near unity. The energetic neutron source power was separately adjusted to maintain average wall loading at 0.5 MW/m2 DT fusion power equivalent. Since all actinides have energy to be harvested through fission and if not fissioned have long half-lives requiring long term isolation, the overall objective was to find steady-state operating conditions with no actinides ever removed. These consist of steady power levels, continuous removal rates (sec−1) for fission/transmutation products, steady continuous addition rates for actinide fueling isotopes, and an associated steady inventory of isotopes in the molten salt consistent with keff=1 criticality of the MSR. Simulations were run until averaged changes in the molten salt's isotope inventory became negligible, signaling steady-state conditions. Actinide removal rates were held at zero, thus requiring that actinides go in but never come out. Eight different cases were run resulting in different final simulated values for total fission product inventories and for fission to energetic neutron power ratios. In cases 1 through 6, the initial actinide inventories and the continuously added actinides were entirely uranium-238. In Case 7 they were a typical spent nuclear fuel mixture from light water reactors and in Case 8 they were thorium-232. Case 1 suppressed fission product (FP) generation, equivalent to infinite FP removal rates. Case 2 assigned arbitrary removal rates to each FP element. Case 3 reduced removal rates by a factor of ten while Case 4 increased them by a factor of 10. It is significant that although Cases 2-4 varied FP removal rates by a factor of 100, they converged to steady solutions in which all actinides were fully consumed. Having shown the HMSR can consume all actinides, attention in Case 5 turned to other components of the radioactive waste stream. Table 2 lists the longest half-life radioactive fission product isotopes sorted by their half-lives. Its point is that there are only a few radioactive FP isotopes with half-lives so long that they present a challenging long term waste disposal problem similar to unfissioned actinides. TABLE 2Radioactive Fission Products Sorted by Half-LifeFissionsteady inventoryYieldper annual fissionseqZASymbolNameHalf-Life%*rate153129129Iiodine-12915.7 million yrs0.81.81E+05246107107Pdpalladium-1076.5million yrs1.21.13E+05355135135Cscesium-1352.3million yrs6.92.29E+054409393Zrzirconium-931.53million yrs5.51.21E+055347979Seselenium-79327thousand yrs0.041.89E+02650126126Sntin-126230thousand yrs0.13.32E+017439999Tctechnetium-99211thousand yrs6.11.86E+04862151151Smsamarium-15190yrs0.56.49E−01950121121mSntin-121m43.9yrs0.000053.17E−051055137137Cscesium-13730.2yrs6.32.79E+0011389090Srstrontium-9028.9yrs4.51.88E+001248113113mCdcadmium-113m14.1yrs0.00081.63E−0413133Htritium12.32yrs0.023.55E−0314368585Krkrypton-8510.76yrs0.23.10E−021563155155Eueuropium-1554.76yrs0.085.49E−031661147147Pmpromethium-1472.62yrs2.258.50E−021755134134Cscesium-1342.07yrs0.00082.39E−051844106106Ruruthenium-1061.02yrs0.405.89E−03*Total fission yield is 200% Case 5 attempted to apply the successful actinide non-removal strategy to onerous long-lived FPs by zeroing removal rates for ten elements, i.e., iodine, palladium, cesium, zirconium, selenium, tin, technetium, samarium, strontium, and cadmium, with the objective of transmuting their long-lived radioactive isotopes into shorter lived or stable isotopes without requiring expensive isotopic separation. The resulting inventories failed to stabilize due to rapid build-up of stable non-radioactive isotopes of four of them, i.e., palladium, samarium, selenium, and especially zirconium. Case 6 returned those four elements to their Case 2 removal rates. The non-removal strategy with stable actinide inventories then obtained small Case 6 inventories for the other six FP elements not removed, i.e., iodine, cesium, tin, technetium, strontium, and cadmium, but some of their inventories, were still growing albeit very slowly due to their stable non-radioactive isotopes. A small positive removal rate may be appropriate for these. Steady fission product inventories in Cases 1-4 and 6-7 were all less than 0.8 atom % while their ratios of fission power to energetic neutron power varied from 244 to 1053. Steady fission product inventories showed an inverse correlation with steady power ratios, but sensitivities to individual fission product isotopes would have required many more runs to determine. Case 7 simulated SNF in the initial load and in continuous fueling. Although it had a high FP inventory its high power ratio results from SNF's inclusion of some fissile material with its uranium-238. Case 8 simulated thorium-232 fueling, showing the HMSR can consume all actinides from a thorium cycle, albeit with a power ratio of only 184. Cases 1 through 4 and 6 through 8 all demonstrated stable actinide inventories and stable criticality with fresh actinides being added and no actinides ever removed. Case 5 failed to converge and was halted after its FP fraction rose above 8.62 atom %. Table 3 lists steady isotope inventories for fueling cases with uranium-238, with SNF from light water reactors, and with thorium. It is noteworthy that these cases have low fissile concentrations and fissile isotopes are mixed with non-fissile isotopes of the same elements. TABLE 3Steady Isotope Inventories (atom %),Case 2Case 7Case 8Isotope(U238)(SNF)(Th232)li715.12414.95814.769f1965.89366.43466.268na238.2048.0848.010FPs0.1010.7390.527th2300.0000.0000.005th2320.0000.00010.027pa2310.0000.0000.002pa2330.0000.0000.002u232*0.0000.0000.006u233*0.0000.0000.208u2340.0000.0000.101u235*0.0000.0000.009u2360.0050.0290.053u2389.7898.8430.002np2370.0020.0020.003np2390.0000.0030.000pu2380.0040.0030.002pu239*0.0470.0420.000pu2400.0820.0650.001pu241*0.0250.0220.000pu2420.2220.1980.001pu2440.0010.0010.000am2410.0010.0000.000am2430.0900.0840.001cm2420.0010.0010.000cm2440.1780.2180.001cm245*0.0040.0060.000cm2460.1750.2090.001cm247*0.0050.0060.000cm2480.0470.0540.000*fissile isotopes Results of this study confirm the HMSR can completely consume all supplied actinides using uranium, SNF or thorium fuels, and that the steady fission to energetic neutron power ratio is sufficiently large for the neutron source to be less than 1% of total plant power. While this invention has been particularly shown and described with references to example embodiments thereof, it will be understood by those skilled in the art that various changes in form and details may be made therein without departing from the scope of the invention encompassed by the appended claims. |
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claims | 1. A magnetic field coil arrangement for a magneto-optical trap, comprising:a first transparent substrate having a first surface;a second transparent substrate having a second surface opposite from the first surface;one or more side walls coupled between the first and second transparent substrates;a first set of magnetic field coils on the first surface of the first transparent substrate; anda second set of magnetic field coils on the second surface of the second transparent substrate, the second set of magnetic field coils in an offset alignment with the first set of magnetic field coils;wherein the first and second sets of magnetic field coils are configured to produce a magnetic field distribution that mimics a quadrupole magnetic field distribution in a central location between the first and second transparent substrates. 2. The magnetic field coil arrangement of claim 1, wherein the first and second transparent substrates each comprise a glass panel. 3. The magnetic field coil arrangement of claim 1, wherein the first set of magnetic field coils are electrically connected to one or more power sources, and the second set of magnetic field coils are electrically connected to one or more power sources. 4. The magnetic field coil arrangement of claim 1, wherein the first set of magnetic field coils includes a first coil, a second coil, and a third coil, in a substantially planar configuration and spaced apart from each other around a central location on the first surface of the first transparent substrate. 5. The magnetic field coil arrangement of claim 4, wherein the second set of magnetic field coils includes a fourth coil, a fifth coil, and a sixth coil, in a substantially planar configuration and spaced apart from each other around a central location on the second surface of the second transparent substrate. 6. The magnetic field coil arrangement of claim 5, wherein:the first coil is connected to a first current source such that a current flows in a counter clockwise direction around the first coil;the second coil is connected to a second current source such that a current flows in a clockwise direction around the second coil; andthe third coil is connected to a third current source such that a current flows in a clockwise direction around the third coil. 7. The magnetic field coil arrangement of claim 6, wherein:the fourth coil is connected to a fourth current source such that a current flows in a clockwise direction around the fourth coil;the fifth coil is connected to a fifth current source such that a current flows in a counter clockwise direction around the fifth coil; andthe sixth coil is connected to a sixth current source such that a current flows in a counter clockwise direction around the sixth coil. 8. The magnetic field coil arrangement of claim 5, wherein:the first coil is connected to a first current source such that a current flows in a clockwise direction around the first coil;the second coil is connected to a second current source such that a current flows in a clockwise direction around the second coil; andthe third coil is connected to a third current source such that a current flows in a clockwise direction around the third coil. 9. The magnetic field coil arrangement of claim 8, wherein:the fourth coil is connected to a fourth current source such that a current flows in a counter clockwise direction around the fourth coil;the fifth coil is connected to a fifth current source such that a current flows in a counter clockwise direction around the fifth coil; andthe sixth coil is connected to a sixth current source such that a current flows in a counter clockwise direction around the sixth coil. 10. A magneto-optical trap device, comprising:a vacuum cell comprising:a first transparent panel having a first surface;a first set of magnetic field coils on the first surface of the first transparent panel;a second transparent panel having a second surface opposite from the first surface;a second set of magnetic field coils on the second surface of the second transparent panel, the second set of magnetic field coils in an offset alignment with the first set of magnetic field coils;one or more side walls coupled between the first and second transparent panels; anda vacuum chamber enclosed by the first and second transparent panels, and the one or more sidewalls;a plurality of power sources electrically connected to the first and second sets of magnetic field coils; anda plurality of laser devices each configured to direct a laser beam through a respective magnetic field coil in the first and second sets of magnetic field coils such that the laser beams intersect along orthogonal axes in a central location of the vacuum chamber;wherein the first and second sets of magnetic field coils produce a magnetic field distribution that mimics a quadrupole magnetic field distribution in the central location of the vacuum chamber. 11. The magneto-optical trap device of claim 10, wherein the first and second transparent panels each comprise a glass panel. 12. The magneto-optical trap device of claim 10, wherein the first set of magnetic field coils includes a first coil, a second coil, and a third coil, in a substantially planar configuration and spaced apart from each other around a central location on the first surface of the first transparent panel. 13. The magneto-optical trap device of claim 12, wherein the second set of magnetic field coils includes a fourth coil, a fifth coil, and a sixth coil, in a substantially planar configuration and spaced apart from each other around a central location on the second surface of the second transparent panel. 14. The magneto-optical trap device of claim 13, wherein:the first coil is connected to a first current source such that a current flows in a counter clockwise direction around the first coil;the second coil is connected to a second current source such that a current flows in a clockwise direction around the second coil; andthe third coil is connected to a third current source such that a current flows in a clockwise direction around the third coil. 15. The magneto-optical trap device of claim 14, wherein:the fourth coil is connected to a fourth current source such that a current flows in a clockwise direction around the fourth coil;the fifth coil is connected to a fifth current source such that a current flows in a counter clockwise direction around the fifth coil; andthe sixth coil is connected to a sixth current source such that a current flows in a counter clockwise direction around the sixth coil. 16. The magneto-optical trap device of claim 10, wherein the vacuum cell further comprises an additional magnetic field coil on the first surface that substantially surrounds the first set of magnetic field coils. 17. The magneto-optical trap device of claim 16, wherein the vacuum cell further comprises an additional magnetic field coil on the second surface that substantially surrounds the second set of magnetic field coils. 18. A method of fabricating a vacuum cell for a magneto-optical trap, the method comprising:forming a first set of magnetic field coils on a first surface of a first transparent substrate;forming a second set of magnetic field coils on a second surface of a second transparent substrate;attaching the first and second substrates to one or more side walls such that the first surface is opposite from the second surface, and the second set of magnetic field coils is in an offset alignment with the first set of magnetic field coils; andforming a vacuum chamber enclosed by the first and second transparent substrates, and the one or more sidewalls, wherein the first and second sets of magnetic field coils produce a magnetic field distribution that mimics a quadrupole magnetic field distribution in a central location of the vacuum chamber. 19. The method of claim 18, wherein the first set of magnetic field coils includes a first coil, a second coil, and a third coil, which are formed in a substantially planar configuration and spaced apart from each other around a central location on the first surface of the first transparent substrate. 20. The method of claim 19, wherein the second set of magnetic field coils includes a fourth coil, a fifth coil, and a sixth coil, which are formed in a substantially planar configuration and spaced apart from each other around a central location on the second surface of the second transparent substrate. |
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050193236 | claims | 1. A method of producing Iodine-124, said method comprising: placing a target means comprising copper in a nickel plating solution and electroplating said target means with nickel; placing the resulting target means in an isotopically enriched Tellurium-124 dioxide plating solution and electroplating said target means with Tellurium-124; placing the resulting target means in line with a deuteron beam of a cyclotron, thereby irradiating the Tellurium-124 and creating Iodine-124 by the.sup.124 Te(d,2n).sup.124 I reaction; and separating the Iodine-124 from the target means. the target means is a copper metal plate which is first milled and uniformly lapped, said copper metal plate being sanded, washed with distilled water, and dried prior to electroplating. Tellurium-124 electroplating is accomplished by means of a solution of isotopically enriched Tellurium-124 dioxide dissolved in a solution of potassium hydroxide and by means of a platinum electrode. the target thickness is between 10 and 14 milligrams per square centimeter. the Iodine-124 is used as a radioactive standard for nuclear detection calibration. placing a target means comprising copper in a nickel plating solution and electroplating said target means with nickel; placing the resulting target means in a Tellurium-124 plating solution and electroplating said target means with Tellurium-124; placing the resulting target means in line with the particle beams of a cyclotron, thereby irradiating the Tellurium-124 and creating Iodine-124; separating the Iodine-124 from the target means; and combining the Iodine-124 with meta-iodobenzylguanidine. the Iodine-124 is combined with the meta-iodobenzylguanidine in a method comprising the mixing of meta-iodobenzylguanidine sulphate with copper nitrate in a borosilicate serum vial, adjusting the pH to about 5, heating the solution to 150 degrees centigrade, cooling, adding a sodium biphosphate buffer solution, and passing the filtrate through an anion-exchange resin. creating a target matrix by electroplating Tellurium-124 onto a nickel surface of a water cooled copper plate, whereby the resulting Tellurium-124 concentration is at least 0.1 milligrams per square centimeter, bombarding the electroplated tellurium for about four hours with a 50 microampere beam current comprising deuteron particles having a particle energy of at least 6.5 MeV, thereby producing an Iodine-124 product, allowing said iodine product to decay for about 40 hours, and separating Iodine-124 from the target matrix. the Tellurium-124 concentration of the target matrix is about 10 to 14 milligrams per square centimeter. the resulting Iodine-124 is incorporated into a substance selected from the group including the following: a steroidal group, an aryl group, a substituted aryl group, a vinyl group, an aryl group capable of coupling with antibodies, an aromatic amine, an aromatic isocyanate, benzoic acid, a substituted benzoic group, a vinylestradial group, monoclonal antibodies, polyclonal antibodies, steroids, cholesterol derivatives, estrogen derivatives, hormones, and proteins. irradiation is conducted in the range of 25 to 80 microamperes deuteron beam current with irradiation doses ranging from 100 to 500 microampere hours. disposing on a target means a substantial amount of Tellurium-124, irradiating said Tellurium-124 and transforming a substantial amount of said Tellurium-124 into Iodine-124 by the .sup.124 Te(d,2n).sup.124 I reaction, chemically removing said Iodine-124 from said target means to produce a solution having radioisotopes of iodine which are primarily Iodine-124, removing from said solution a substantial portion of deleterious salts, whereby autoradiolytic decomposition of said Iodine-124 is substantially reduced. preparing an ion exchange column and passing said solution through said column, and heating said solution to reduce the volume so that its final concentration is about 15 to 120 mCi per 1.0 ml or greater specific activity. said Iodine-124 is separated from trace Tellurium, and said solution is subsequently purified by removing salts from said solution by preparing an ion exchange column and passing said solution through said column, and heating said solution to reduce its volume so that its concentration is about 15 to 120 mCi per 1.0 ml. or greater specific activity. 2. The method of producing Iodine-124 of claim 1 wherein: 3. The method of producing Iodine-124 of claim 2 wherein: 4. The method of producing Iodine-124 of claim 3 wherein: 5. The method of producing Iodine-124 of claim 1 wherein the irradiated target is placed in a solution of sodium hydroxide solution containing hydrogen peroxide and water, subsequently, the solution is transferred to a vessel containing aluminum powder, thereafter the solution so purged with air, then carbon dioxide gas, particles in the solution are then filtered out and passed through a cation-exchange column. 6. The method of making Iodine-124 in claim 1 whereby: 7. A method of synthesizing Iodine-124 labeled meta-iodobenzylguanidine, said method comprising: 8. The method of claim 7 wherein 9. A method of synthesizing Iodine-124 to a purity of about 99.5%, said method comprising: 10. The method of synthesizing Iodine-124 of claim 9 whereby: 11. The method of synthesizing Iodine-124 of claim 10 whereby: 12. The method of making Iodine-124 of claim 9 wherein: 13. A method of making and purifying an Iodine-124 solution comprising the steps of: 14. A method of making and purifying an Iodine-124 solution in accordance with claim 13 wherein said step of removing salts from said solution includes: 15. A method of making and purifying an Iodine-124 solution in accordance with claim 13 wherein: |
summary | ||
claims | 1. A scanning electron microscope system comprising:input means for inputting pattern design data of a pattern to be formed on a photomask designed by taking an optical proximity effect into account;SEM image acquiring means for acquiring a SEM image of a resist pattern formed by exposing a resist film coated on a surface of a substrate with light through the photomask;simulating means for simulating a circuit pattern to be formed on a substrate by using the pattern design data inputted by the input means;superposing means for superposing the SEM image of the resist pattern formed by the SEM image acquiring means and the simulated circuit pattern formed by the simulating means;geometrical feature extracting means for extracting one- or two-dimensional geometrical features indicating differences between the superposed SEM image and the simulated circuit pattern; andevaluating index calculating means for calculating evaluation indices for evaluating a correction of the optical proximity effect of the mask pattern formed on the photomask by using information about the one- or two-dimensional geometrical features provided by the feature extracting means. 2. The dimension measuring scanning electron microscope system according to claim 1, wherein the geometrical features to be extracted by the feature extracting means includes all or some of widths of component lines of a circuit pattern, distance between a design position of an end part of the circuit pattern and an actual position of the same end part of the circuit pattern receded from the design position, interpattern distance between adjacent circuit patterns, rounding degree of a corner of the circuit pattern, radius of a circular pattern in the circuit pattern, ratio between major and minor axes of an elliptic pattern in the circuit pattern, and area of the circuit pattern. 3. The dimension measuring scanning electron microscope system according to claim 1 further comprising an exposure simulator including an input means for entering design data on the exposure circuit pattern and exposure conditions and capable of calculating data on a transfer circuit pattern to be formed when the substrate is exposed to design data on the exposure circuit pattern on the basis of the design data on the exposure circuit pattern and the exposure conditions entered by the input means, wherein the evaluation index calculating means calculates data on the positional relation between the transfer circuit pattern and the SEM image of the exposure circuit pattern by subjecting the data on a transfer circuit pattern obtained by the exposure simulator and the SEM image of the exposure circuit pattern to an aligning process, and the design data on the exposure circuit pattern and the data on the SEM image of the exposure circuit pattern are superposed by replacing the calculated data on the transfer circuit pattern with the design data on the exposure circuit pattern. 4. The dimension measuring scanning electron microscope system according to claim 3, wherein the calculated data on the simulated circuit pattern calculated by the exposure simulator is data on a shape of a section formed by cutting a light intensity distribution on a plane corresponding to a proper light intensity or data on a shape of a section in close resemblance with the SEM image of the mask pattern among shapes of sections formed by cutting the light intensity distribution on planes corresponding to different light intensities respectively. 5. The dimension measuring scanning electron microscope system according to claim 1, wherein the evaluation index calculating means includes a display means for displaying an image formed by superposing part of or all the images defined by the data on the SEM image of the mask pattern and the design data on the mask pattern. 6. The dimension measuring scanning electron microscope system according to claim 1, wherein the evaluation index calculating means includes a display means for displaying quantities of some of or all the geometrical features. 7. A circuit pattern evaluating system comprising:input means for inputting design data on a mask pattern formed on a photomask;SEM image acquiring means for acquiring a SEM image of a resist pattern formed by exposing a resist film coated on a substrate with light through the photomask on which the mask pattern is formed;simulating means for simulating a pattern to be formed on the substrate by using the inputted design data on the mask pattern;matching means for superposing the SEM image of the resist pattern acquired by the SEM image acquiring means and the pattern simulated by the simulating means;resist pattern evaluating means for evaluating a quality of the resist pattern by determining differences between the superposed SEM image of the resist pattern and the simulated pattern; anddisplay means for displaying information obtained by the resist pattern evaluating means on a screen. 8. The circuit pattern evaluating system according to claim 7 the resist pattern evaluating means evaluates the resist pattern in terms of all or some of widths of component lines of a circuit pattern, distance between a design position of an end part of the circuit pattern and an actual position of the same end part of the circuit pattern receded from the design position, interpattern distance between adjacent circuit patterns, rounding degree of a corner of the circuit pattern, radius of a circular pattern in the circuit pattern, ratio between major and minor axes of an elliptic pattern in the circuit pattern, and area of the circuit pattern. 9. The circuit pattern evaluating system according to claim 7, wherein the simulating means calculates a light intensity distribution when the mask pattern is used for exposure, and uses a shape of a section formed by cutting the light intensity distribution on a plane corresponding to a proper light intensity or a shape of a section in close resemblance with a measured image of the mask pattern among shapes of sections formed by cutting the light intensity distribution on planes corresponding to different light intensities respectively, as a simulated circuit pattern. 10. The circuit pattern evaluating system according to claim 7, wherein the display means displays an image formed by partly or totally superposing the SEM image of the resist pattern and the simulated circuit pattern formed by the simulating means by the matching means. 11. The circuit pattern evaluating system according to claim 7 further comprising a warning means for giving a warning when deviations of parts of the SEM image of the resist pattern from corresponding parts of the simulated circuit pattern representing the quality of the resist pattern are outside corresponding tolerances. 12. A circuit pattern evaluating method comprising the steps of:inputting design data of a pattern to be formed on a photomask by taking an optical proximity effect into account;acquiring a SEM image of a resist pattern formed by exposing a resist film coated on a substrate with light through a photomask on which a mask pattern is formed by using the design data of the pattern;simulating a circuit pattern to be formed on the substrate from the inputted design data of the photomask;superposing the SEM image of the resist pattern and the simulated circuit pattern;extracting one- or two-dimensional geometrical features indicating differences between the SEM image and the simulated circuit pattern from the superposed SEM image and the simulated circuit pattern; andcalculating evaluation indices for evaluating a corrected optical proximity effect of the mask pattern formed on the photomask by using information on the one- or two-dimensional geometrical features extracted by the step of extracting one- or two-dimensional geometrical features. 13. The circuit pattern evaluating method according to claim 12, wherein the information about the geometrical features used by the step of calculating the evaluation indices includes all or some of widths of component lines of a circuit pattern, distance between a design position of an end part of the circuit pattern and an actual position of the same end part of the circuit pattern receded from the design position, interpattern distance between adjacent circuit patterns, rounding degree of a corner of the circuit pattern, radius of a circle in the circuit pattern, ratio between major and minor axes of an elliptic pattern in the circuit pattern, and area of the circuit pattern. 14. The circuit pattern evaluating method according to claim 12, the step of entering the design data enters also exposure conditions for exposing the substrate to exposure radiation, and the step of forming a simulated circuit pattern forms a simulated pattern that is expected to be formed on the substrate when the mask pattern defined by the entered design data on the design mask pattern and the exposure conditions are used for exposure. 15. The circuit pattern evaluating method according to claim 14, wherein the exposure conditions entered in the step of entering the design data includes wavelength of exposure radiation, numerical aperture of an objective lens, apparent size of a light source and at least one of pieces of information about defocus. 16. A circuit pattern evaluating method comprising the steps of:inputting design data on a mask pattern to be formed on a photomask;acquiring a SEM image of a resist pattern formed by exposing a resist film coated on a substrate with light through a photomask on which the mask pattern is formed;simulating a circuit pattern to be formed on the substrate by using the inputted design data on the mask pattern;superposing the SEM image of the resist pattern and the simulated circuit pattern;evaluating a quality of the resist pattern on the basis of differences between the SEM image and the simulated circuit pattern derived from the superposed SEM image and the simulated circuit pattern; anddisplaying results of evaluation of the quality of the resist pattern. 17. The circuit pattern evaluating method according to claim 16, wherein the step of evaluating the quality of the resist pattern evaluates the resist pattern in terms of all or some of widths of component lines of a circuit pattern, distance between a design position of an end part of the circuit pattern and an actual position of the same end part of the circuit pattern receded from the design position, interpattern distance between adjacent circuit patterns, rounding degree of a corner of the circuit pattern, radius of a circular pattern in the circuit pattern, ratio between major and minor axes of an elliptic pattern in the circuit pattern, and area of the circuit pattern. 18. The circuit pattern evaluating method according to claim 16, wherein the step of forming a simulated circuit pattern calculates a light intensity distribution when the mask pattern is used for exposure, and uses a shape of a section formed by cutting the light intensity distribution on a plane corresponding to a proper light intensity or a shape of a section in close resemblance with a measured image of the mask pattern among shapes of sections formed by cutting the light intensity distribution on planes corresponding to different light intensities respectively, as a simulated circuit pattern. 19. The circuit pattern evaluating method according to claim 16, wherein the step of displaying results of evaluation of the quality of the resist pattern displays an image formed by partly or totally superposing the SEM image of the resist pattern and the simulated circuit pattern formed by the simulating means by the matching means. 20. The circuit pattern evaluating method according to claim 16, wherein the step of evaluating the quality of the resist pattern gives a warning when deviations of parts of the SEM image of the resist pattern from corresponding parts of the simulated circuit pattern representing the quality of the resist pattern are outside corresponding tolerances. |
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abstract | Open pit mine (OPM) structures are modified or built new for use in disposing of low-level radioactive/nuclear waste (LLW). A drainage system is added to the OPM to drain water, such as, but not limited to, rain water, out of a volume of the OPM and to a particular geologic zone located far below the OPM that is isolated away from the local water table. Cells are formed within the volume of the OPM that are configured to receive the LLW. Cells are added to the OPM from a bottom towards a top of the OPM. Void spaces around the LLW materials within the cells are filled in with a protective-medium to mitigate against radionuclide migration away from the LLW materials within the cells. The protective-medium may be a blend of carbon nanotubes and a foam cement slurry. The carbon nanotubes may be made from reacting ethylene with vermiculite. |
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062366998 | abstract | Compliance with administrative limits on cumulative exposure of control rod groups in the reactor core, is monitored by computing the incremental effective exposure for each group commensurate with core power, for each time increment at which each group is within the position range where an administrative limit is imposed. The increments of effective exposures for each group are accumulated, and the accumulated effective exposure for each group is compared with the administrative limit to each group. This comparison is then displayed to the reactor operator, preferably using either a "rolling wheel" or "sliding bar" format. |
055919839 | claims | 1. A radiation emitting apparatus, comprising: a source of radiation for providing a substantially unshaped radiation beam in a given beam direction, and a collimator for shaping said radiation beam, said collimator comprising, first and second layers of a plurality of elongated radiation blocking leaves, frame means for supporting said leaves, and moving means for moving said leaves, said leaves of each layer being arranged adjacent one another so as to form two opposed rows of adjacently positioned leaves and being movable by said moving means in a longitudinal direction (Y) which is generally traverse to the beam direction so as to define a radiation beam shaping field between the opposed ends of the leaves, and said layers being arranged one above another in the beam direction and offset in a lateral direction (X) generally transverse to the beam direction and orthogonal to the longitudinal direction (Y) so that spaces between adjacent leaves of said first and second layers are positioned over and under, respectively, leaves of said second and said first layers, respectively; wherein each row of said collimator leaves comprises a plurality of adjacently positioned relatively narrow width leaves bounded on at least one end with a relatively wider width end leaf which is also movable in the longitudinal direction. a source of radiation for providing a substantially unshaped radiation beam in a given beam direction, and a collimator for shaping said radiation beam, said collimator comprising, first and second layers of a plurality of elongated radiation blocking leaves, frame means for supporting said leaves, and moving means for moving said leaves, said leaves of each layer being arranged adjacent one another so as to form two opposed rows of adjacently positioned leaves and being movable by said moving means in a longitudinal direction (Y) which is generally traverse to the beam direction so as to define a radiation beam shaping field between the opposed ends of the leaves, and said layers being arranged one above another in the beam direction and offset in a lateral direction (X) generally transverse to the beam direction and orthogonal to the longitudinal direction (Y) so that spaces between adjacent leaves of said first and second layers are positioned over and under, respectively, leaves of said second and said first layers, respectively; wherein each of said collimator leaves comprises an identical arrangement of collimator leaves; and wherein each layer of collimator leaves comprises a central plurality of a set of opposed and adjacently positioned relatively narrow width leaves bounded at one end with a set of a relatively wider width leaf. a source of radiation for providing a substantially unshaped radiation beam in a given beam direction, and a collimator for shaping said radiation beam, said collimator comprising, first and second layers of a plurality of elongated radiation blocking leaves, frame means for supporting said leaves, and moving means for moving said leaves, said leaves of each layer being arranged adjacent one another so as to form two opposed rows of adjacently positioned leaves and being movable by said moving means in a longitudinal direction (Y) which is generally traverse to the beam direction so as to define a radiation beam shaping field between the opposed ends of the leaves, and said layers being arranged one above another in the beam direction and offset in a lateral direction (X) generally transverse to the beam direction and orthogonal to the longitudinal direction (Y) so that spaces between adjacent leaves of said first and second layers are positioned over and under, respectively, leaves of said second and said first layers, respectively; wherein each of said collimator leaves comprises an identical arrangement of collimator leaves; and wherein said first and second layers of said collimator are positioned so that the order of said leaf arrangement in said first layer is a mirror image of the order of the leaf arrangement in said second layer. a first layer of a plurality of elongated radiation blocking leaves, said leaves being arranged adjacent one another so as to form two opposed rows of adjacently positioned leaves and being movable in a longitudinal direction (Y) which is generally traverse to the beam direction so as to define a radiation beam shaping field between the opposed ends of the leave; a second layer of a plurality of elongated radiation blocking leaves, said leaves of said second layer also being arranged adjacent one another so as to form two opposed rows of adjacently positioned leaves and being movable in a longitudinal direction (Y) which is generally traverse to the beam direction so as to define a radiation beam shaping field between the opposed ends of the leaves; frame means for supporting said leaves of said first layer and said leaves of said second layer; and moving means for moving said leaves of said first layer and said leaves of said second layer in the longitudinal direction; wherein said first and second layers are arranged one above another in the beam direction and are offset in a lateral direction (X) generally transverse to the beam direction, so that spaces between adjacent leaves in said first and second layers are positioned over and under, leaves of said second and said first layers; and wherein each row of said collimator leaves comprises a plurality of adjacently positioned relatively narrow width leaves bounded on at least one end with a relatively wider width end leaf which is also movable in the longitudinal direction. a first layer of a plurality of elongated radiation blocking leaves, said leaves being arranged adjacent one another so as to form two opposed rows of adjacently positioned leaves and being movable in a longitudinal direction (Y) which is generally traverse to the beam direction so as to define a radiation beam shaping field between the opposed ends of the leaves; a second layer of a plurality of elongated radiation blocking leaves, said leaves of said second layer also being arranged adjacent one another so as to form two opposed rows of adjacently positioned leaves and being movable in a longitudinal direction (Y) which is generally traverse to the beam direction so as to define a radiation beam shaping field between the opposed ends of the leaves; frame means for supporting said leaves of said first layer and said leaves of said second layer; and moving means for moving said leaves of said first layer and said leaves of said second layer in the longitudinal direction; wherein said first and second layers are arranged one above another in the beam direction and are offset in a lateral direction (X) generally transverse to the beam direction, so that spaces between adjacent leaves in said first and second layers are positioned over and under, respectively, leaves of said second and said first layers; wherein each of said collimator layers comprises an identical arrangement of collimator leaves; and wherein each layer of collimator leaves comprises a central plurality of a set of opposed and adjacently positioned relatively narrow width leaves bounded at one end with a set of a relatively wider width leaf. a first layer of a plurality of elongated radiation blocking leaves, said leaves being arranged adjacent one another so as to form two opposed rows of adjacently positioned leaves and being movable in a longitudinal direction (Y) which is generally traverse to the beam direction so as to define a radiation beam shaping field between the opposed ends of the leaves; a second layer of a plurality of elongated radiation blocking leaves, said leaves of said second layer also being arranged adjacent one another so as to form two opposed rows of adjacently positioned leaves and being movable in a longitudinal direction (Y) which is generally traverse to the beam direction so as to define a radiation beam shaping field between the opposed ends of the leaves; frame means for supporting said leaves of said first layer and said leaves of said second layer; and moving means for moving said leaves of said first layer and said leaves of said second layer in the longitudinal direction; wherein said first and second layers are arranged one above another in the beam direction and are offset in a lateral direction (X) generally transverse to the beam direction, so that spaces between adjacent leaves in said first and second layers are positioned over and under, respectively leaves of said second and said first layers; wherein each of said collimator layers comprises an identical arrangement of collimator leaves; and wherein said first and second layers of said collimator are positioned so that the order of said leaf arrangement in said first layer is a mirror image of the order of the leaf arrangement in said second layer. 2. The radiation emitting apparatus of claim 1, wherein each row of said collimator leaves is bounded at its other end by a relatively wider width trimmer leaf movable in the lateral direction. 3. The radiation emitting apparatus of claim 2, wherein each of said leaves that are movable in said longitudinal direction are able to be retracted by an amount sufficient to permit said trimmer leafs to be individually extendable into said radiation beam so as to define a width for said radiation field. 4. A radiation emitting apparatus, comprising: 5. The radiation emitting apparatus of claims 4, wherein an other end of said narrow width leaves is bounded with a relatively wider width leaf movable in the lateral direction. 6. The radiation emitting apparatus of claims 5, wherein said first and second layers of said collimator are positioned so that the order of said leaf arrangement in said first layer is a mirror image of the order of the leaf arrangement in said second layer. 7. A radiation emitting apparatus, comprising: 8. A multileaf collimator for use in a radiation system providing a radiation beam in a given beam direction, comprising: 9. The radiation emitting apparatus of claim 8, wherein each row of said collimator leaves is bounded at its other end by a relatively wider width trimmer leaf movable in the lateral direction. 10. The radiation emitting apparatus of claim 9, wherein each of said leaves that are movable in said longitudinal direction are able to be retracted by an amount sufficient to permit said trimmer leafs to be individually extendable into said radiation beam so as to define a width for said radiation field. 11. A multileaf collimator for use in a radiation system providing a radiation beam in a given beam direction, comprising: 12. The radiation emitting apparatus of claim 11, wherein an other end of said narrow width leaves is bounded with a relatively wider width leaf movable in the lateral direction. 13. The radiation emitting apparatus of claims 12, wherein said first and second layers of said collimator are positioned so that the order of said leaf arrangement in said first layer is a mirror image of the order of the leaf arrangement in said second layer. 14. A multileaf collimator for use in a radiation system providing a radiation beam in a given beam direction, comprising: |
description | The present invention relates at least generally to the field of detecting fissile, high energy neutrons and, more particularly, to an advanced fissile neutron detection system and associated method. Governments mobilize radiation detectors to attempt to stop the illicit movement of nuclear material such as plutonium and uranium. Previous approaches to neutron detection have relied upon an isotope of helium gas, helium-3 or 3He, a limited resource generated during the construction and/or decommissioning of nuclear weapons which is already showing signs of a global short supply. Due to increasing 3He shortages and the resulting increase in associated costs, neutron detectors utilizing 3He cannot be economically deployed at scales. Efforts to develop replacement technologies have been initiated, however, none of these efforts have produced a cost effective, scalable solution. The lack of scalable technology has limited the evolution of existing systems to meet evolving threats. Specifically, current modeling efforts show that the deployment of a large, networked array of detection technologies where the detectors are placed at potential points of attack, material source locations, and discreetly at randomized points of transportation pathways will lead to the greatest increase of overall security against nuclear threats. Plutonium and highly enriched uranium (HEU) materials that can be used in a nuclear weapon emit both gamma rays and neutrons. After the attacks on Sep. 11, 2001, the U.S. government sought to strengthen border defenses against smuggled Special Nuclear Materials (SNM). To detect SNM, federal, state, and local governments initially deployed detection units using 3He gas in proportional counters wrapped in high-density polyethylene (HDPE) a technology pulled from physics laboratories and the nuclear power industry. Polyvinyltoluene (PVT) plastics coupled to photomultiplier tubes (PMT), pulled from the scrap-steel industry, were used to detect gamma rays emitted by HEU, as well as other dangerous radioactive sources that could be used to create a radiological dispersive device. Handheld devices, which have better gamma ray energy resolution than PVT, supported the main scanning capabilities of these larger 3He and PVT detectors. This initial detection capability had challenges. The initial deployment of neutron detectors severely depleted the limited stockpile of 3He, driving costs sky-high and limiting scalability of deployment. Equally problematic were the number of false positive alarms that were due to the poor energy resolution of PVT, increasing overall scanning times and limiting the usability of the systems. Multiple government R&D programs over the past ten years have invested in 3He alternatives for neutron detection, as well as improved energy resolution gamma ray detection units. However, while some alternative materials have emerged, Applicants believe that none of the R&D programs succeeded in reducing the cost of these systems. Given that 1.2 million kilograms of Pu has been produced since World War II, and its key signature is neutron emission, neutron detection is now considered a non-negotiable component of threat detection capability. In view of the foregoing, Applicants recognize that new neutron detection solutions are needed. Applicants further recognize that the solution should: Be low cost and independent of 3He. This will enable scalable, affordable solutions; Have low probability for gamma-ray induced false positives by having high gamma ray rejection; Be rugged and long lived for compatibility with military CONOPS; and Hit metrics of capture area and efficiency to detect the desired threats as a major advance in the overall reduction of nuclear threats. The foregoing examples of the related art and limitations related therewith are intended to be illustrative and not exclusive. Other limitations of the related art will become apparent to those of skill in the art upon a reading of the specification and a study of the drawings. The following embodiments and aspects thereof are described and illustrated in conjunction with systems, tools and methods which are meant to be exemplary and illustrative, not limiting in scope. In various embodiments, one or more of the above-described problems have been reduced or eliminated, while other embodiments are directed to other improvements. In general, a fissile neutron detection system is described for detecting incident fissile neutrons. In one aspect of the disclosure, the fissile neutron detection system includes an ionizing thermal neutron detector arrangement including an inner peripheral shape that at least substantially surrounds a moderator region for detecting thermal neutrons that exit the moderator region but is at least generally transparent to the incident fissile neutrons. A moderator arrangement is disposed within the moderator region for converting the incident fissile neutrons in the moderator region to thermal neutrons which exit the moderator region to then enter the thermal neutron detector arrangement for detection of at least some of the thermal neutrons to produce an electrical current as a detector output with the moderator arrangement having an outer peripheral shape that is at least generally complementary to the inner peripheral shape and the moderator arrangement and includes lateral extents such that any given dimension that bisects the lateral extents includes a length that is greater than any thickness of the moderator arrangement transverse to the lateral extents. In another aspect of the disclosure, the fissile neutron detection system includes an ionizing thermal neutron detector arrangement including an inner peripheral shape that at least substantially surrounds a moderator region for detecting thermal neutrons that exit the moderator region but is at least generally transparent to the incident fissile neutrons. A moderator arrangement is disposed within the moderator region for converting the incident fissile neutrons in the moderator region to thermal neutrons which exit the moderator region to then enter the thermal neutron detector arrangement for detection to produce an electrical current as a detector output with the moderator arrangement having an outer peripheral shape that is at least generally complementary to the inner peripheral shape and the moderator arrangement includes major widthwise and major lengthwise lateral extents such that any given dimension across the lengthwise and widthwise lateral extents includes a length that is greater than any thickness of the moderator arrangement transverse to the lateral extents. Although the following Detailed Description will proceed with reference being made to illustrative embodiments, many alternatives, modifications and variations thereof will be apparent to those skilled in the art. The following description is presented to enable one of ordinary skill in the art to make and use the invention and is provided in the context of a patent application and its requirements. Various modifications to the described embodiments will be readily apparent to those skilled in the art and the generic principles taught herein can be applied to other embodiments. Thus, the present invention is not intended to be limited to the embodiments shown, but is to be accorded the widest scope consistent with the principles and features described herein including modifications and equivalents, as defined within the scope of the appended claims. It is noted that the drawings are not to scale and are diagrammatic in nature in a way that is thought to best illustrate features of interest. Descriptive terminology can be used with respect to these descriptions, however, this terminology has been adopted with the intent of facilitating the reader's understanding and is not intended as being limiting. The present disclosure brings to light a fissile neutron detection system. The term fissile neutron is intended to refer to a high energy neutron that is typically emitted by Pu or HEU. By way of example, the energy of a fissile neutron can be in the range from 100 keV to 10 MeV. The fissile neutron detection system can include a first thermal neutron detector and a second thermal neutron detector. The first thermal neutron detector and the second thermal neutron detector can each include: a chamber containing at least one active material that emits at least one ionizing particle upon exposure to thermal neutrons and at least one electrode. The system can further include a neutron moderator disposed proximate the first thermal neutron detector and the second thermal neutron detector in a space between the first thermal neutron detector and the second thermal neutron detector. The neutron moderator can include a hydrogen-containing material that transitions at least a portion of high-energy incident fissile neutrons to low-energy thermal neutrons. In some implementations, the fissile neutron capture efficiency of such a system can exceed 50%. In some implementations, the active material can include one or more sheets of a solid material such as lithium-6 (6Li) or boron-10 (10B) that can emit a number of charged particle(s) upon capture of a thermal neutron. In some implementations, the neutron moderator can include one or more solid thermoplastic materials, such as high-density polyethylene (HDPE), which includes a high percentage of atoms per cm3 of hydrogen. A fissile neutron detection method is provided. In an embodiment, the method can include transitioning at least some incident fissile neutrons to thermal neutrons by passing the incident fissile neutrons through at least one neutron moderator proximate a first thermal neutron detector and a second thermal neutron detector. The neutron moderator is located in a space between the first thermal neutron detector and the second thermal neutron detector. The neutron moderator includes a hydrogen containing material that causes the transition of the at least some incident fissile neutrons to thermal neutrons. The method further includes impinging at least a portion of the thermal neutrons exiting the neutron moderator on either of: at least one active material disposed in a chamber of the first thermal neutron detector or, at least one active material disposed in a chamber of the second thermal neutron detector. The method also includes generating, by the first thermal neutron detector and the second thermal neutron detector, a current proportional to the number of thermal neutrons impinging on the active material in the first thermal neutron detector and on the active material in the second thermal neutron detector. The relatively high-energy fissile neutron (energy level >100 keV) enters the neutron moderator and collides with hydrogen nuclei within the moderator. The collisions reduce the energy level of the fissile neutron to a relatively low-energy thermal neutron (energy level <0.1 eV). The collisions may also cause a portion of the incident fissile neutrons to exit the moderator along a vector that does not intersect the first thermal neutron detector or the second thermal neutron detector. Thus only a portion of the incident fissile neutrons may be captured by the thermal neutron detectors. Using the detector/moderator/detector arrangement described herein offers a significant cross-sectional area for fissile neutron capture, providing capture efficiencies of greater than 60%. Such capture efficiency exceeds the capture efficiency of other neutron detection systems such as systems using 10B straw detectors placed in a moderator block which offer significantly less cross-sectional area for neutron capture. At least some of the thermal neutrons exiting the neutron moderator enter the first thermal neutron detector or the second thermal neutron detector. Within the thermal neutron detector, the thermal neutron impinges on an active material. In some instances, the active material can capture the thermal neutron and generate a number of daughter particles such as an alpha particle (two protons/two neutrons) and triton (one proton/two neutrons). At least some of the daughter particles can ionize a readout gas within the neutron detector. The drift electrons and ionized readout gas cause a charge flow within the neutron detector. The charge flow can be captured by an electrode as an electrical current. One or more properties of the electrical current can be indicative of the fissile neutron detected by the thermal neutron detector. Another embodiment of a fissile neutron detection system is provided which includes at least one thermal neutron detector. Each thermal neutron detector can include a body having a length, a width, and a thickness defining a closed chamber; the length and the width of the chamber greater than the thickness of the chamber, at least one active material that emits at least one ionizing particle upon exposure to thermal neutrons, the active material disposed within the chamber and at least one electrode disposed in the chamber. The fissile neutron detection system can further include at least one neutron moderator disposed proximate the at least one neutron detector. Each neutron moderator can include a material that transitions at least a portion of high-energy incident fissile neutrons to low-energy thermal neutrons. An embodiment of a fissile neutron detection method is also provided. The embodiment can include transitioning at least some incident fissile neutrons to thermal neutrons by passing the incident fissile neutrons through at least one neutron moderator disposed proximate at least one thermal neutron detector. Each thermal neutron detector can include a body having a length, a width, and a thickness defining a closed chamber; the length and the width of the chamber greater than the thickness of the chamber. Each neutron detector can further include at least one active material that emits at least one ionizing particle upon exposure to thermal neutrons, the active material disposed within the chamber and at least one electrode disposed in the chamber. The method can additionally include impinging at least 60% of the thermal neutrons exiting the neutron moderator on the at least one active material disposed in the chamber of the at least one neutron detector. The method can further include generating, by the at least one thermal neutron detector, a current at the at least one electrode, the current proportional to the number of thermal neutrons impinging on the at least one active material in the at least one thermal neutron detector. Still another embodiment of a fissile neutron detection system is provided. The fissile neutron detection system can include a first thermal neutron detector, a second thermal neutron detector, and a neutron moderator. Each of the neutron detectors can include a body having a length, a width, and a thickness that define a hermetically sealed, continuous chamber, wherein the length and the width of the body exceed a thickness of the body. Each thermal neutron detector can also include at least one active material disposed within the hermetically sealed chamber, the at least one active material to emit at least one charged particle upon exposure to a thermal neutron. Each thermal neutron detector can additionally include at least one electrode disposed within the chamber and electrically isolated from the body. The neutron moderator can be disposed proximate the first thermal neutron detector and the second thermal neutron detector. The neutron moderator can be disposed in a space between the first thermal neutron detector and the second thermal neutron detector. The neutron moderator can include one or more hydrogen-containing materials that transition at least a portion of high-energy incident fissile neutrons to low-energy thermal neutrons that are detectable by the first thermal neutron detector or the second thermal neutron detector. As used herein, the terms “top” and “bottom”, as well as other descriptive terminology, are intended to provide a relative rather than absolute reference to a location and is not intended to limit a positional relationship merely on the basis of descriptive nomenclature. Thus, inverting an object having a “top cover” and a “bottom cover” can place the “bottom cover” on the top of the object and the “top cover” on the bottom of the object. Such configurations should be considered as included within the scope of this disclosure. As used herein, the terms “first,” “second,” and other similar ordinals are intended to distinguish a number of similar or identical objects and not to denote a particular or absolute order of the objects. Thus, a “first object” and a “second object” can appear in any order including an order in which the second object appears before or prior in space or time to the first object. Such configurations should be considered as included within the scope of this disclosure. FIG. 1A is a front elevational view of an illustrative embodiment of a fissile neutron detection system 100 that includes a first thermal neutron detector 102A, a second thermal neutron detector 102B (collectively, “thermal neutron detectors 102”) and a neutron moderator 150 that is at least partially disposed proximate first thermal neutron detector 102A and second thermal neutron detector 102B in a space, cavity or volume formed between first thermal neutron detector 102A and second thermal neutron detector 102B, in accordance with at least one embodiment of the present disclosure. FIG. 1B is a partial sectional view of a portion of fissile neutron detection system 100 shown circled in FIG. 1A, in accordance with at least one embodiment of the present disclosure. FIG. 1B diagrammatically depicts operational level details of fissile neutron detection system 100. FIG. 1C is a side elevation of illustrative fissile neutron detection system 100 depicted in FIG. 1A, in accordance with at least one embodiment of the present disclosure. FIG. 1D is a partial sectional view of fissile neutron detection system 100 shown circled in FIG. 1C, in accordance with at least one embodiment of the present disclosure. FIG. 1D diagrammatically depicts operational level details of fissile neutron detection system 100. FIG. 1E is a partial sectional view of an embodiment of an isolator 128 through which an electrode 116 is disposed within thermal neutron detector(s) 102. Although first thermal neutron detector 102A and second thermal neutron detector 102B are depicted and described in association with FIGS. 1A-1E, other arrangements that include only a single neutron detector 102 are considered as embodiments of the concepts described herein. Neutron moderator 150 can be a single or multi-piece member having any number or combination of shapes or configurations. Regardless of shape and/or configuration, neutron moderator 150 has a surface area that can be defined by the external sides of neutron moderator 150. Such surface area can include exposed (e.g., outwardly facing) exterior surfaces, hidden (e.g., inwardly facing) exterior surfaces, or any combination thereof. For example, the surface area of a planar neutron moderator 150 having an annulus therethrough would include the surface area of the plane, the “edges” of the plane, and the surface area about the perimeter of the annulus. A majority of the surface area of neutron moderator 150 can be disposed proximate one or more thermal neutron detectors 102. In various embodiments, neutron moderator 150 can include a proximate surface area (i.e., a surface area that is in contact and/or nearest to the thermal neutron detectors, as opposed to extending beyond the periphery or moderator region defined by the thermal neutron detector(s)) that is substantially greater than the total moderator surface area in one embodiment and greater than approximately 60% of the total moderator surface area in another embodiment. The moderator or moderator arrangement serves to define what may be referred to as the lateral extents of the detection system. For example, the lateral extents of an orthorectangular moderator likewise define a rectangle. The lateral extents can define any suitable shape without limitation. A flux of fissile neutrons can be can be incident on the fissile neutron detection system generally originating along either one or both of two major receiving directions with a first major receiving direction 152a diagrammatically indicated by arrows and a second major receiving direction 152b diagrammatically indicated by arrows. These example receiving directions are at least generally opposite of one another. While the operation of fissile neutron detection system 100 can conveniently be described and understood based on the first and second opposing major receiving directions, it should be appreciated that the incident fissile neutrons or a flux(es) thereof can be received from a wide range of receiving directions, most of which are not normal to the opposing major outwardly facing surfaces of the overall assembly. Thus, the detection of fissile neutrons is not limited to the first and second major receiving directions. However, while Applicants recognize that the fissile neutron detection systems described herein are certainly capable of efficiently detecting fissile neutrons over a wide range of incident angles, these designs and methods are nevertheless explicitly intended for optimization with respect to the major receiving directions rather than omnidirectional receiving capability. Indeed, many distinguishing aspects of these teachings are considered by Applicants to be related to performance optimization that could be rendered somewhat less clear if any specification of receiving directionality was simply removed from consideration, and at least some aspects of these teachings could be rendered moot in applications wherein directional requirements stray sufficiently far from those that are discussed herein. Applicants recognize that thermal neutron detectors 102A and 102B are at least substantially transparent to the incident fissile neutrons. That is, the incident fissile neutrons generally do not interact with the thermal neutron detectors, but instead travel through the thermal neutron detectors to then reach moderator 150. The neutrons are most often detected by the thermal neutron detector(s) subsequent to moderation/slowing. Turning to FIG. 1B, at least a portion of the incident fissile neutrons, one of which is indicated by the reference number 160, impinging upon neutron moderator 150 can enter, strike, or otherwise impinge upon neutron moderator 150. Since the present descriptions relating to fissile neutron 160 apply essentially to all incident fissile neutrons that are subsequently detected, the reference number 160 can interchangeably reference a flux or plurality of incident fissile neutrons. Based on the descriptions above, it should be appreciated that fissile neutron 160 may have already transited through thermal neutron detector 102A. A path 161 for fissile neutron 160 through moderator 160 is shown as a straight line, however, it is noted that this path can be complex, and considered as a chain of causality between its origination and end points rather than an actual trajectory. Details with respect to this chain of causality will be described at various points hereinafter. Within neutron moderator 150, energy of the fissile neutron can be reduced in elastic scattering events in which the fissile neutrons collide with hydrogen nuclei within the moderator. As a result of these collisions, the energy level of fissile neutron 160 is reduced to that of a thermal neutron which is indicated by the reference number 162. Also, as a result of the random nature of these collisions, low-energy thermal neutron 162 can exit the neutron moderator along the same or a different vector than incident fissile neutron 160. By placing a majority of the surface area of neutron moderator 150 proximate to each neutron detector 102, the probability of detecting thermal neutron 162 exiting the neutron moderator is increased. In various embodiments, the probability that incident fissile neutron(s) 160 will pass through detector 102 after exiting neutron moderator 150 as thermal neutron(s) 162 can be approximately 60%. As will be described hereinafter, the percent of thermal neutrons detected can depend on a number of factors, and can be determined by a designer, at least in part, based on tradeoffs that can influence the thickness, weight and cost of the overall detector system. Each of neutron detectors 102 includes a top cover 104, a bottom cover 108, and sidewalls 120 that form a chamber 105. In some implementations, one or more gas tight or gas impervious seals 124 can be disposed in the joints formed by the sidewall 120 and the top cover 104 and the sidewall 120 and the bottom cover 108. In at least some implementations, the seals 124 can isolate or hermetically seal chamber 105 to minimize or even prevent exchange of gases or fluids between chamber 105 and the exterior environment. In some implementations, chamber 105 can include a single, continuous (i.e., uninterrupted) chamber 105. In some implementations, chamber 105 can contain one or more gases or gas combinations. For example, chamber 105 can contain a noble gas such as argon (Ar). In some implementations, at least one layer or sheet of active material 112 is arranged with a support matrix 106 (FIG. 1C) to support each layer or sheet of active material 112. In this embodiment, a number of electrodes 116 can extend partially or completely through all or a portion of the chamber 105. Isolators 128 can be disposed at locations where the electrodes 116 extend through a wall or cover of the chamber 105 to electrically isolate electrode 116 from the sidewall or cover of chamber 105. Although not depicted in FIGS. 1A-1E, in some implementations, all or a portion of the number of electrodes 116 can penetrate the top cover 104 and/or bottom cover 108 to enter the chamber 105 of the neutron detector 102 rather than penetrating sidewall 120. In such instances, one or more isolators 128 can be disposed about electrodes 116 at the point the electrodes penetrate into chamber 105. As depicted in FIGS. 1A thru 1D, during operation, neutrons impinge upon fissile neutron detection system 100. Fissile neutrons 160 can be produced, for example, by plutonium (Pu) or other highly enriched uranium (HEU) products such as may be found in nuclear or radiological explosive device. Within chamber 105 of each of thermal neutron detectors 102, a sheet of active material 112, such as lithium 6 (6Li) can be disposed on a support structure 106 such as, for example, an aluminum honeycomb matrix. Fissile neutrons impinging upon fissile neutron detection system 100 can pass through neutron moderator 150 where the energy level of at least a portion of incident fissile neutrons 160 (between 100 keV and 10 MeV) can be reduced to the energy level of a thermal neutron 162 (e.g., less than 0.1 eV). Thermal neutron(s) 162 can be captured by one of the 6Li atoms contained in active material 112. The capture of thermal neutron 162 by the 6Li atom forms a lithium 7 (7Li) atom that can decay into two daughter particles, an alpha particle 166 and a triton 168. Triton 168 and alpha particle 166 travel in opposite directions, and dissipate energy as they travel through active material 112. Upon exiting active material 112, at least some of tritons 168 or alpha particles 166, having sufficient kinetic energy, ionize atoms in readout gas 170 disposed within chamber 105. Electrons 172 produced by the ionization of readout gas 170 can drift towards electrodes 116 in the chamber and the ionized gas generated by the ionization of readout gas 170 can drift towards active layer 112. Electrons 172 that drift within amplification region 176 (i.e., the Townsend avalanche region approximately 5 times the radius of electrode 116) encounter an electric field that accelerates drifting electrons 172 to a sufficient velocity that additional readout gas 170 can be ionized. The additional ionized readout gas can create additional electrons 172 that also tend to drift toward electrodes 116 and cause additional ionization of readout gas 170. This process that occurs within the Townsend avalanche can be referred to as “gas multiplication.” Ionized atoms of readout gas 170 within the Townsend avalanche region that move towards the active layer 112 induce a current flow along electrode 116. In some implementations, the current along electrode 116 can be collected and amplified using a pulse-mode, charge-sensitive preamplifier to produce a voltage output signal 188 on an output 192. Pulse height discrimination circuitry can be used to compare the voltage output signal to a first defined threshold to determine whether fissile neutron(s) 160 has been detected (e.g., for a gas multiplication of roughly 100, and an amplification circuitry gain of approximately 1 fC/mV, pulse heights greater than approximately 250 keV can indicate the presence and/or detection of fissile neutron(s) 160). In some embodiments, the false positive detection rate of fissile neutron(s) 160 based on the first predetermined threshold can be less than 1×10−5 for a gamma ray exposure rate of 100 mR/hr. A second predetermined threshold can be selected and can be set at a value that is less than the first predetermined threshold. Voltage output signals 192 below the second predetermined threshold can be deemed as very low ionizing gamma ray events or movements of charge in fissile neutron detection system 100 that were induced by another source (e.g., thermal heat, radio frequency electromagnetic radiation, and changes in the relative position of electrodes 116 and active layer 112—known as microphonics). Voltage output signal 188 below the first predetermined threshold and above the second predetermined threshold can be indicative of gamma ray events. The detected rate of neutrons and gamma rays impinging upon fissile neutron detection system 100 can be used in radiation detection methodologies (e.g., to detect the presence of a nuclear weapon or unauthorized nuclear device). In embodiments, the composition of readout gas 170 can be maintained relatively constant over time to avoid deterioration of the gamma ray and neutron detection process. Change in readout gas 170 composition greater than 1% in the composition can affect the Townsend avalanche process. For example, nitrogen, oxygen, or water molecules that leak into the chamber 105 may not ionize as well as a readout gas 170, such as argon, in the amplification region 176 near the electrodes 116, and therefore may reduce the Townsend avalanche process near the electrodes 116 when introduced into the readout gas 170. This can reduce the ability of the readout electronics to distinguish between noise, gamma ray, and fissile neutron events, thereby decreasing the efficiency and/or accuracy of fissile neutron detection system 100. A 1% change in the composition of readout gas 170 can cause up to an 8% change in voltage output signal 188. To maintain accuracy and responsiveness of the fissile neutron detection system, it can be advantageous to limit the change in composition of readout gas 170 by minimizing the following: (1) the egress of readout gas 170 from chamber 105; and (2) the ingress of contaminants, including air constituents (nitrogen, oxygen, carbon dioxide), water, and other airborne molecules, into chamber 105. Top cover 104 and bottom cover 108 can be fabricated from one or more materials that permit the passage of thermal neutrons 162 from neutron moderator 150 to chamber 105. In at least some implementations, top cover 104 and bottom cover 108 can be fabricated from one or more suitable stainless steels, such as, an 18/8 stainless steel, a 304 stainless steel, a 304L stainless steel, a 316 stainless steel, or a 316L stainless steel. Other grades and materials can be substituted with equal efficiency. Top cover 104 and bottom cover 108 define the overall configuration of neutron detector 102. In one example, top cover 104 and bottom cover 108 can include generally planar members in such embodiments, the neutron detector 102 can have a generally planar configuration, for example a square periphery of major opposing surfaces having a side length of approximately 20 cm to 100 cm. Top cover 104 and bottom cover 108 can have other shapes, such as triangular, octagonal, hexagonal, circular, elliptical, rectangular, or even irregular shapes to fit within designated areas. Similarly, chamber 105 at least partially formed and/or bounded by top cover 104 and/or bottom cover 108 can have any shape, configuration, or regular/irregular perimeter. For example, the shapes defined by the lateral periphery chamber 105 can be generally square, generally rectangular, generally oval, generally elliptical, generally circular, generally triangular, generally polygonal, generally trapezoidal, or any other regular or irregular configuration. In some implementations, all or a portion of chamber 105 can be spherical or hemispherical and all or a portion of neutron moderator 150 can be spherical and placed concentrically within the chamber 105 of thermal neutron detector 102. In some embodiments, chamber 105 can be defined by three, mutually orthogonal, measurements, such as a length, a width, and a thickness. Major lateral extents of chamber 105 can have any suitable dimensions in accordance with a given application, and various thicknesses can be selected in accordance with the teachings and descriptions herein. In various such embodiments, top cover 104 and/or bottom cover 108 can define either or both the length and the width of chamber 105. In such embodiments, sidewall 120 can define the thickness of chamber 105. Chamber 105 can have a length and width that greatly exceed the thickness of chamber 105. In some embodiments, the length of chamber 105 measured along a first axis can exceed the thickness of chamber 105 measured along a second axis orthogonal to the first axis by a factor of approximately 20 times or greater. In some embodiments, the width of chamber 105 measured along a third axis can exceed the thickness of chamber 105 measured along the second axis orthogonal to the third axis by a factor of approximately 5 times or greater. In some embodiments, chamber 105 can have a length, measured along a first axis, of approximately 100 centimeters (cm). In some embodiments, chamber 105 can have a thickness, measured along a second axis orthogonal to the first axis, of approximately 2 to 5 centimeters (cm). In some embodiments, chamber 105 can have a width, measured along a third axis orthogonal to the first axis and the second axis, of approximately 20 centimeters (cm). In other embodiments, a number of which will be described at appropriate points hereinafter, top cover 104 and bottom cover 108 can have configurations other than planar, for example top cover 104 can include a simple or compound curved surface having a first radius while bottom cover 108 can include a similar simple or compound curved surface having a second radius that is greater or less than the first radius. Such an implementation can provide a thermal neutron detector that is curved, arced, or hemispherical. In yet other embodiments, top cover 104 and bottom cover 108 can have generally similar irregular shapes that permit the construction of thermal neutron detectors 102 having virtually any size, shape, and/or physical configuration. Such irregular shapes can, for example, advantageously permit the custom fitting of thermal neutron detectors 102 within odd or irregular shaped housings. In at least some implementations, all or a portion of top cover 104 and/or bottom cover 108 can be integrally formed with all or a portion of sidewall 120 to eliminate one or more joints between respective cover 104, 108 and sidewall 120. In some implementations, all or a portion of top cover 104 and/or bottom cover 108 can be affixed to all or a portion of sidewall 120 using one or more adhesives, by welding or brazing, or similar attachment or affixment techniques capable of providing a gas tight seal between sidewall 120, top cover 104, and bottom cover 108. In some implementations, top cover 104, bottom cover 108, and at least a portion of sidewall 120 can be integrally formed, for example using one or more casting, extrusion, injection molding, or similar processes in which all or a portion of top cover 104, all or a portion of bottom cover 108, and a portion of sidewalls 120 are seamlessly formed. In some embodiments, some or all of seals 124 between sidewall 120 and top cover 104 and/or sidewall 120 and bottom cover 108 can be formed from an elastomeric compound that is compressed or otherwise formed to the mating surfaces of sidewall 120 and top cover 104 and/or sidewall 120 and bottom cover 108. In some implementations all or a portion of seals 124 can include polyisobutylene or one or more polyisobutylene containing compounds to maintain the composition of readout gas 170 over an extended timeframe (e.g., 30 years). Beneficially, the use of flexible elastomeric seal 124 provides the ability for seal 124 to conform to the mating surfaces of sidewall 120, top cover 104 and/or bottom cover 108, filling any imperfections in the mating surfaces and minimizing the likelihood of readout gas 170 leakage through gaps formed by imperfections in the mating surfaces. In some embodiments, the quality of the mating surfaces found on top cover 104, bottom cover 108, and/or sidewalls 120 can be selected to generate uniform electric fields near electrodes 116 of the fissile neutron detection system 100 (e.g., the variance in the finish on the mating surfaces of top cover 104, bottom cover 108, and/or sidewall 120 can be equal to or less than 0.020″ inches). Providing such a surface finish on the mating surfaces improves sealing of the chamber 105 and takes advantage of the sealing properties of seal 124, at least when elastomeric, which can accommodate such fluctuations in the surface finish of the mating surfaces. The use of elastomeric seal 124 can also facilitate a low temperature manufacturing process that minimizes or even eliminates high temperature processes, such as welding or brazing, on fissile neutron detection system 100 which reduces warping and bending of the components of fissile neutron detection system 100. An elastomeric seal can also accommodate thermal expansion/contraction of the chamber components, thereby allowing a greater number of material choices for top plate 104, bottom plate 108, and sidewalls 120 such as glass, aluminum, or stainless steel. Elastomeric seal 124 can have a thickness in the range of approximately 25 micrometers (μm) to approximately 1 centimeter (cm) and a width in the range of approximately 1 cm to approximately 5 cm. Such an elastomeric seal 124 can provide less than 1% leakage, as a percentage of the volume of the chamber, of an argon-methane readout gas 170 from chamber 105, and less than 1% leakage of oxygen into chamber 105 over a 30 year period for chamber 105 having a length of approximately 0.5 m, a width of approximately 1 m, and a thickness of approximately 1 cm. In one implementation, elastomeric seal 124 can include a polyisobutylene seal 124 having a width of approximately 1.5 cm, a total surface area of 30 square centimeters (cm2), can maintain an oxygen leak rate into the chamber 105 of approximately 1.3×10−10 cm3·cm/(s·cm2·cm-Hg). A leak rate of approximately 1.3×10−10 cm3·cm/(s·cm2·cm-Hg) provides an oxygen concentration of approximately 0.75% by volume for a chamber 105 having a volume of approximately 5000 cubic centimeter (cm3) after 30 years of operation. In some embodiments, all or a portion of sidewalls 120 can be fabricated using one or more metallic materials, such as stainless steel. In some embodiments, all or a portion of the sidewall 120 can be fabricated using aluminum or an aluminum containing alloy. In some embodiments, sidewall 120 can have a mating surface or lip that, upon assembly, is disposed proximate top cover 104, bottom cover 108, or both top cover 104 and bottom cover 108. In some implementations, the mating surface can be machined or similarly finished to remove irregularities in the surface and provide a relatively smooth sealing surface. In some embodiments, readout gas 170 can include one or more pure or nearly pure noble gases, such as argon (Ar). In some embodiments, readout gas 170 can include a gas mixture, for example a gas containing 90 percent by volume (vol %) argon and 10 vol % quenching gas such as carbon dioxide or methane. In some implementations, the voltage bias applied to electrode 116 can be adjusted, controlled, or otherwise altered based at least in part on the composition of readout gas 170. In such instances, a small (e.g., 1%) change in the bias voltage applied to electrode(s) 116 can cause a larger change (e.g., up to approximately 15%) in the voltage output signal provided by or generated by electrode(s) 116. Active material 112 disposed in chamber 105 can include one or more sheets of active material disposed within the chamber, one or more layers of active material disposed within the chamber, or in some implementations, an active gas disposed within chamber 105. In some implementations, the active material can include lithium 6 (6Li), boron 10 (10B), and helium 3 (3He) as well as some combination thereof. In some embodiments, active material 112 can include a sheet of active material such as a sheet of 6Li foil that, in some embodiments, can be supported within chamber 105 by support matrix 106. The latter can be, by way of non-limiting example, a honeycomb formed, for instance from aluminum or a stainless steel mesh that can be generally planar, rather than a honeycomb. In such implementations, the length and width of sheet 112 of active material can greatly exceed the thickness of the layer of active material. In such implementations, the length and width of sheet 112 of active material can greatly exceed the thickness of neutron detector 102. In such implementations, the length and width of sheet 112 of active material can greatly exceed the thickness of chamber 105. In some embodiments using one or more sheets of 6Li foil as active material 112, each sheet of 6Li foil can have a length and a width each of which greatly exceeds the thickness of the foil. In some embodiments, the sheet of 6Li foil can have a thickness of approximately 100 micrometers (μm). In some embodiments, the sheet of 6Li foil can have a width of approximately 20 centimeters (cm). In some embodiments, the sheet of 6Li foil can have a length of approximately 100 centimeters (cm). Embodiments in which active layer 112 is disposed at an intermediate point within the chamber can advantageously detect tritons 166 emitted from both sides of active layer 112. In contrast, tritons 166 emitted from only one side of active layer 112 can be detected in embodiments in which active layer 112 is disposed proximate top cover 104 and/or bottom cover 108 rather than disposed at an intermediate point in chamber 105. In some embodiments, active material 112 can include a layer of active material such as a layer of 10B that can be disposed on substrate that is disposed within chamber 105. In some embodiments, active material 112 can include a layer of active material such as a layer of 10B that can be disposed (e.g., via chemical vapor deposition or similar processes) on all or a portion of an interior surface of top cover 104, bottom cover 108, and/or sidewalls 120 forming chamber 105. In such implementations, each of the length and width of the layer of active material 112 can greatly exceed the thickness of the layer of active material 112. In such implementations, each of the length and width of the layer of active material 112 can greatly exceed the thickness of thermal neutron detector 102. In such implementations, each of the length and width of the layer of active material 112 can greatly exceed the thickness of chamber 105. With regard to the foregoing discussions and those which follow, it should be appreciated that the terms “length” and “width” are applied in terms of describing the lateral extents of major surface areas of the various components in a pair of orthogonal directions that are transverse or orthogonal to the thicknesses of the various components as well as to the thickness of the overall fissile neutron detection system. Furthermore, insofar as a given detector is configured for operation with respect to a given receiving direction, it should be appreciated that the lateral extents of that given detector can generally be considered as being at least approximately orthogonal to that receiving direction. For the case of bidirectional detectors the two major receiving directions can be aligned along a single receiving axis with one major receiving direction antiparallel to the other major receiving direction, and the lateral extents can be oriented at least approximately orthogonal to that receiving axis. In some instances, a peripheral outline of the lateral extents of an irregularly-shaped fissile neutron detector can have a lateral extent that does not exceed the subject thickness requirements, however, the lateral extents will nevertheless be seen to circumscribe at least one major area in a plan view which major area falls within the scope of the present disclosure as well as the appended claims. In some implementations, active material 112 can include one or more active gas species, for example helium 3 (3He). In such instances, chamber 105 can be filled with one or more active gases or a mixture that includes one or more active gases. In some implementations, a combination of active sheets, active layers, and/or active gases can be disposed within chamber 105. In some implementations, all or a portion of top cover 104 and/or bottom cover 108 can be formed into a dished or tray-like form such that top cover 104 and/or bottom cover 108 form at least a portion of sidewall 120, and can, on occasion, form the entirety of sidewall 120 of chamber 105. In some implementations, neutron detector 102 can have a thickness (that includes top cover 104, sidewall 120 (if present), and bottom cover 108 of approximately 2 to 5 centimeters (cm). Top cover 104 and bottom cover 108 can have any suitable dimensions, geometry, and/or configuration to provide thermal neutron detector 102 having any suitable shape or geometry. In some implementations, thermal neutron detector 102 can be in the physical configuration of a planar structure having a length and width that greatly exceeds the thickness of thermal detector 102. In some implementations, the length of thermal neutron detector 120, measured along a first axis, can be from approximately 5 or more times the thickness of detector 102 to approximately 100 or more times the thickness of detector 102. In some implementations, the width of thermal neutron detector 102, measured along a second axis that is orthogonal to the first axis, can be from approximately 3 or more times the thickness of thermal detector 102 to approximately 50 or more times the thickness of thermal detector 102. In some implementations, thermal neutron detector 102 can have a length, measured along a first axis, of from approximately 10 centimeters (cm) or greater to approximately 1000 cm or greater; a thickness, measured along a second axis orthogonal to the first axis, of from approximately 0.5 centimeters (cm) or less to approximately 5 cm or less; and a width, measured along a third axis orthogonal to the first axis and the second axis of from approximately 30 cm to approximately 500 cm. In such implementations, top cover 104 and bottom cover 108 can have a corresponding width of from approximately 30 cm to approximately 500 cm; and a corresponding length of from approximately 10 cm or less to approximately 100 cm or less. Other thermal neutron detector 102 physical configurations are possible. For example, thermal neutron detector 102 can be curved about a single axis to provide chamber 105 that is arced or parabolic. In such an implementation, top cover 104 and bottom cover 108 can be arced or parabolic along the desired axis to provide chamber 105. In another example, thermal neutron detector 102 can be curved about two axes to provide chamber 105 that is a concave dish, a convex dish, or hemispherical. In such an implementation, top cover 104 and bottom cover 108 can be arced or dished along the respective axes to provide the arced or dished chamber 105. In some implementations, top cover 104 and/or bottom cover 108 can be fabricated using one or more stainless steels, aluminum, or one or more aluminum alloys. Top cover 104 and/or bottom cover 108 can be made of glass such as soda-lime or borosilicate glass. In some embodiments, some or all of electrodes 116 can pass through sidewall 120 of the neutron detector 102. In some embodiments, some or all of electrodes 116 can pass through top cover 104 and/or bottom cover 108 of thermal neutron detector 102. Any number of electrodes 116 can be disposed within chamber 105. Each of electrodes 116 can have any profile or shape, for example, electrodes 116 can include conductors having a round cross section with a diameter of from approximately 25 micrometers (μm) to approximately 150 μm. In embodiments, electrodes 116 can be tensioned to approximately 33% to approximately 67% of the breaking or failure limit for the electrode material. In another embodiment that is illustrated by a subsequent figure, a single feedthrough can couple to a plurality of electrodes 116. One or more isolators 128 can electrically isolate electrodes 116 from sidewall 120, top cover 104, and/or bottom cover 108 of neutron detector 102. In some implementations, one or more isolators 128 can hermetically seal about electrode 116, thereby maintaining the hermetic integrity of chamber 105. In some implementations, each of one or more isolators 128 can permit the passage of electrode 116 through an aperture extending through isolator 128. After passing electrode 116 through isolator 128, the space around isolator 128 can be filled using a material such as solder, conductive epoxy, brazing, or welding. The tube length through isolator 128 and the inner diameter of isolator 128 can be selected based on a variety of factors. For example, the shear strength of Sn-37Pb and Sn-3.5Ag solder can exceed 3000 pounds per square inch (psi). With a tension of approximately 450 grams or 1 pound on a 50 μm diameter tungsten rhenium wire, a solder length of approximately 7 millimeters (mm) would provide a safety factor of 5. Isolators 128 can include any current or future developed electrical insulator. Non-limiting examples of such electrical insulators include, but are not limited to, glass isolators, ceramic isolators, Bakelite isolators, resin isolators, epoxy isolators, and similar. In some implementations, thermal neutron detector 102 can include one or more isolator feedthrough inserts 126. Beneficially, feedthrough inserts 126 can be manufactured separate from thermal neutron detector 102 using a separate process that provides a glass-to-metal or ceramic-to-metal feedthrough assembly process. Such construction permits the formation of a hermetic seal between the isolator feedthrough inserts 126, isolator 128 and electrode 116 without requiring feedthrough inserts 126 to be incorporated during the manufacturing process of neutron detector 102. Feedthrough inserts 126 can be modularly constructed and can contain any number of electrodes 116. Feedthrough inserts 126 can be affixed to thermal neutron detector 102 via one or more processes such as welding or brazing. In some implementations, electrodes 116 can be disposed generally parallel to each other and extending from a first side (i.e., major surface) of thermal neutron detector 102 to a second side (i.e., major surface) of thermal neutron detector 102. Any suitable electrode configuration can be used, for example, implementations in which some or all of electrodes 116 are arranged in a pattern such as a star pattern in which electrodes 116 are not parallel to each other. In various embodiments, electrodes 116 can be maintained at the same potential or different potentials. For example, in thermal neutron detectors 102 using a sheet type active material 112, an electrical source 190 can maintain electrodes 116 at a positive or negative potential measured with respect to sheet-type active material 112. In some implementations, electrodes 116 can be maintained at a potential of approximately 1100 volts (V) greater than active material 112. Moderator 150 includes one or more materials capable of reducing an energy level of fissile neutron 160 to an energy level of thermal neutron 162. Such reduction in energy level of fissile neutron 160 occurs within moderator 150 as fissile neutron 160 impacts nuclei in moderator 150. Moderator 150 can include one or more materials that include a minimum of approximately 10 percent atoms per cm3. Moderator 150 can include one or more solids, one or more liquids, and/or one or more compressed gases, or combinations thereof. The use of moderators containing predominantly larger nuclei (e.g., carbon) can disadvantageously cause ricocheting (rather than the preferred slowing) of the incident fissile neutrons 160. In at least some implementations, all or a portion of moderator 150 can be disposed between first thermal neutron detector 102A and second thermal neutron detector 102B. In some implementations, no air gap or similar void is formed between moderator 150 and the exterior surface of top cover 104 and/or exterior surface of the bottom cover 108 of the neutron detector 102. In other words, the major, opposing sides (i.e., major, opposing surfaces) of the moderator can be in direct physical contact with one of the major surfaces of each thermal neutron detector 102. In some implementations, an air gap (or some other form of void or space disposed between the moderator and the detector arrangement) can exist between moderator 150 and the exterior surface of top cover 104 and/or the exterior surface of bottom cover 108 of neutron detector 102. Applicants recognize that in the context of the described embodiments the distance between the moderator or moderator arrangement and the thermal neutron detectors can generally be minimized in order to ensure that thermal neutrons do not escape the detection system after exiting the moderator. In a manner consistent with minimal gaps, embodiments that are within the scope of the present disclosure can at least substantially fill the volume (i.e., greater than 50 percent) of a moderator region that is defined between the thermal neutron detectors with moderating material. In one embodiment, at least 60 percent of the volume of the moderator region is filled by moderating material. In this regard, a moderating arrangement can include a single member or multiple members of moderating material. Interstitial gaps between multiple members do not contribute to the filled volume. With this disclosure in hand it should be clear that excessive thermal neutron detector-to-moderator spacing will generally reduce overall efficiency. Applicants recognize that for a fissile neutron detection system in which a thermal neutron detector arrangement surrounds a given moderator, as taught herein, increasing the moderator-to-detector spacing, in addition to reducing detection efficiency, will generally require the designer to increase the surface area of the detectors that make up the detector arrangement in order to insure that the thermal neutron detector arrangement continues to surround the moderator at the increased moderator-to-detector spacing. Since the active sheet layer(s) tends to be composed of relatively costly material, such as Lithium, such configurations can result in increased cost with a detrimental effect on detection efficiency. Furthermore, as described at various points previously, insofar as the thermal neutron detectors disclosed herein are designed for some degree of optimization with respect to a given receiving axis, the spacing between the moderator and each thermal neutron detector along the receiving axis can generally be minimized at least within reasonable practical limits, in order to correspondingly minimize instances whereupon thermalized neutrons can escape undetected. While not intending to be bound by theory, Applicants submit that excess space that is not filled by moderating material provides what may be referred to as sideways or grazing escape paths for thermal neutrons to exit from moderating material without thereafter entering a thermal neutron detector to thereby evade detection. For at least this reason, a person of ordinary skill in the art, having this disclosure in hand, should appreciate that for the embodiments described herein detection efficiency is typically enhanced when the volume or region defined by the detector arrangement is at least substantially filled by the moderator, as opposed to cases where moderator-detector spacing results in significant void or otherwise unfilled space within the envelope defined by the inner periphery of the thermal neutron detector arrangement. Detection efficiency is also enhanced responsive to Applicants' recognition that the detector arrangement at least substantially surrounds the moderator arrangement. In FIG. 1A, the neutron moderator is substantially surrounded by the detector arrangement since a majority of the major opposing surfaces of neutron moderator 150 are in a direct confronting relationship with thermal neutron detectors 102A and 102B and, more particularly, in such a confronting relationship with active material 112 within each thermal neutron detector. Examination of the embodiments described below will reveal that the moderator arrangement, forming the core of system 100, is at least substantially surrounded by the neutron detector arrangement. This is in sharp and opposite contrast with the accepted thinking of the prior art. Applicants submit that those of ordinary skill in the art were led to believe that the optimum architecture was to provide a moderator arrangement that surrounds a thermal neutron detector. It is further submitted that Applicants' discoveries, as brought to light herein, serve to sweep aside this misconception of the prior art based, at least in part, on the recognition that the thermal neutron detectors used herein are, at least from a practical standpoint, essentially transparent to the fissile neutrons, resulting in an entirely new and heretofore unseen detection system architecture that embodies, in its essence, a complete reversal of prior art beliefs. Without intending to be bound by theory, Applicants submit that the performance levels provided by the present disclosure are, at least in part, attributable to the capability to capture backscattered neutrons that were lost in prior art architectures. Having discussed the particular case of moderator voids resulting from moderator-to-detector spacing or any general equivalent thereof, it is noted that this consideration in no way limits customizing of the mechanical and material properties of the moderator structure, as a whole. In other words and by way of example, certain moderator materials can actually define voids as part of the characteristic structure of the material that do not contribute sideways or grazing escape paths within the intended meaning. For instance, a moderator material can include voids that make for a composite structure including solid and gaseous regions interspersed with one another. In some implementations, the thickness of moderator 150 disposed between first thermal neutron detector 102A and second thermal neutron detector 102B can have a constant thickness that is greater than the thickness of either first thermal neutron detector 102A and/or second thermal neutron detector 102B. In some implementations, moderator 150 can have a length and a width that is about the same as the length and the width of first thermal neutron detector 102A and second thermal neutron detector 102B. In embodiments, moderator 150 can have a length that is approximately 100 centimeters (cm). In embodiments, moderator 150 can have a width that is approximately 20 centimeters (cm). In embodiments, moderator 150 can have a thickness that is approximately 1 cm to 5 cm. In some implementations, the thickness of moderator 150 can be based in whole or in part on the thickness of either or both thermal neutron detectors 102 adjacent to moderator 150. In embodiments, the thickness of moderator 150 can be approximately 1 to 4 times the thickness of the adjacent neutron detector 102. In some implementations, moderator 150 can include one or more materials having a length and width that both greatly exceed the thickness of moderator 150. The reader's attention is now directed to FIG. 1F, which is a diagrammatic, partially cutaway view of fissile neutron detection system 100, in elevation, shown here for purposes of illustrating paths of incident fissile neutrons. Most of the example paths start at least initially along first major receiving direction 152a. The example paths are not intended to imply that the number of nuclei interactions in the moderator are limited, but rather are intended to illustrate major interactions in the moderator that either contribute to detection or to non-detection. On a path 194a, a fissile neutron passes through thermal neutron detector 102A and is then forward scattered and slowed by moderator 150 to a thermal neutron which is, in turn, detected by thermal neutron detector 102B at 195a. On a path 194b, the fissile neutron passes through thermal neutron detector 102A and is then backscattered and slowed by moderator 150 to a thermal neutron which is, in turn, detected by thermal neutron detector 102A at 195B. On a path 194c, the fissile neutron passes through thermal neutron detector 102A, is forward scattered and slowed through moderator 150 to a thermal neutron and then evades detection by thermal neutron detector 102B at 195c. On a path 194d, the fissile neutron passes through thermal neutron detector 102A and is slowed by moderator 150 to a thermal neutron and lost to hydrogen neutron capture at 195d. On a path 194e, the fissile neutron passes through thermal neutron detector 102A, is back scattered and slowed by moderator 150 to a thermal neutron and then passes back through thermal neutron detector 102A to evade detection at 195e. On a path 194f, the fissile neutron passes through fissile neutron detection system 100 without being slowed down in moderator 150 to evade detection at 195f. On a path 194g, the thermal neutron is incident on an end face of moderator 150 at an angle ϕ with respect to the major surfaces of the overall assembly and then slowed down and scattered by moderator 150 into thermal neutron detector 102A for detection at 195g. On a path 194h, another fissile neutron is incident on an end face of the moderator and is slowed down and scattered by the moderator into thermal neutron detector 102B, but evades detection at 195h. It should be noted that the illustrated paths are provided for purposes of enhancing the understanding of the reader. In view of these path examples, one of ordinary skill in the art will appreciate that the orientation of the incident fissile neutrons can be from any direction and still can result in conversion to thermal neutrons which are subsequently detected. At the same time, it should be appreciated, however, that Applicants have discovered that the structure of the fissile neutron detection system, as brought to light herein, provides for improved efficiency for detection of fissile neutrons that are incident upon the overall detection system at least compared with prior art systems without thermal neutron detectors at least substantially surrounding a moderator region. In this regard, most of the incident fissile neutrons will initially pass through one of the outwardly facing major surfaces of one of the thermal neutron detectors, enter the moderator and then be scattered to lose energy for detection as a thermal neutron by one of the thermal neutron detectors. It should also be noted that the number of scattering events shown for trajectories 194 a, b, c, d, e, g and h in moderator 150 in FIG. 1F has been limited in order to maintain illustrative clarity. More realistically, the number of scattering events that lead to a thermalized neutron from a fissile neutron is 10 to 40 scatters. FIG. 2A is a diagrammatic, partially cutaway view of an illustrative embodiment of a fissile neutron detection system 200 that includes fissile neutron detection system 100, described above, as a core structure having first thermal neutron detector 102A, second thermal neutron detector 102B, and neutron moderator 150 disposed proximate first thermal neutron detector 102A and second thermal neutron detector 102B at least partially within a space 201 formed between first thermal neutron detector 102A and second thermal neutron detector 102B, in accordance with at least one embodiment of the present disclosure. An external or supplemental moderator arrangement 202, thinner than moderator 150, can at least substantially surround system 100 by being positioned at least adjacent to the major outwardly facing surfaces of thermal neutron detectors 102A and 102B. In the embodiment of FIG. 2A, moderator arrangement 202 completely surrounds system 100. The supplemental moderator arrangement can be integrally formed or formed from panels of a sheet material in any suitable manner and adjoined in any suitable manner. FIG. 2B is a diagrammatic, partially cutaway view of another illustrative embodiment of a fissile neutron detection system 200′ that includes first thermal neutron detector 102A, second thermal neutron detector 102B, and a third thermal neutron detector 102C, in accordance with at least one embodiment of the present disclosure. FIG. 2B depicts a first neutron moderator 150A disposed proximate first thermal neutron detector 102A and second thermal neutron detector 102B at least partially within a space or volume formed between first thermal neutron detector 102A and second thermal neutron detector 102B and a second neutron moderator 150B disposed proximate second thermal neutron detector 102B and third thermal neutron detector 102C at least partially within a space or volume formed between second thermal neutron detector 102B and third thermal neutron detector 102C. As depicted in FIGS. 2A and 2B, neutron detectors 102 and neutron moderator(s) 150 can be at least partially enclosed by an external neutron moderator 202. In some implementations, a housing or shell 205 can be disposed about all or a portion of external neutron moderator 202. The thickness of external moderator 202 can be the same or different in various areas of the neutron detectors 102. For example, external neutron moderator 202 can have a first thickness 212 proximate at least a portion of first thermal neutron detector 102A (e.g., on the “top” or exposed portion of the fissile neutron detection system 200) and a second thickness 214 proximate at least a portion of second thermal neutron detector 102B (FIG. 2A) or third thermal neutron detector 102C (FIG. 2B)—e.g., on the “bottom” of fissile neutron detection system 200. In addition, external neutron moderator 202 can have a third thickness 216 proximate the surface of thermal neutron detectors 102. As will be discussed at one or more appropriate points hereinafter, a thickness of the supplemental moderator outward of the major surfaces of the thermal neutron detectors that is less than thickness 210 of moderator 150 can provide for enhanced detection efficiency, as will be further discussed. For the moment, it is sufficient to note that the structure of FIG. 2A can, at least in some cases, provide for a modified sensitivity at least in each of major receiving directions 152a and 152b that is greater than a fundamental receiving sensitivity of core fissile neutron detection system 100 in each of the major receiving directions. In order to maintain an equal modified receiving sensitivity in each of the major receiving directions, thickness 212 is equal to thickness 214. External moderator 202 includes one or more materials capable of reducing an energy level of an incident fissile neutron 160. Such reduction in energy level of incident fissile neutron 160 occurs within external moderator 202 as fissile neutron 160 impacts hydrogen nuclei in the material forming the external moderator 202. External moderator 202 can include one or more materials that include a minimum of approximately 10 weight percent atoms per cm3 of hydrogen. External moderator 202 can include one or more solids, one or more liquids, and/or one or more compressed gases, or combinations thereof. In at least some implementations, external moderator 202 can partially or completely include a hydrogen-containing, solid, thermoplastic, material such as high-density polyethylene (HDPE). In at least some implementations, all or a portion of external moderator 202 can be disposed proximate thermal neutron detectors 102 forming fissile neutron detection system 200. In some implementations, no air gap or similar void can exist between the external moderator 202 and the exterior surface of the neutron detectors 102 forming fissile neutron detection system 200. In some implementations, an air gap or similar void space can exist between external moderator 202 and the exterior major surfaces of neutron detectors 102. As depicted in FIGS. 2A and 2B, each of neutron detectors 102 is separated by a neutron moderator 150 having a thickness 210. In some implementations, each of neutron detectors 102 can be separated by a neutron moderator 150 having a constant thickness 210. In some implementations, each of thermal neutron detectors 102 can be separated by a neutron moderator 150 having a variable thickness 210. In some implementations that include a plurality of neutron moderators 150A-150n in a stacked relationship (e.g., FIG. 2B), each of neutron moderators 150 can have the same or a different constant thickness 210. In some implementations that include a plurality of neutron moderators 150A-150n (e.g., FIG. 2B), each of neutron moderators 150 can have the same or a different variable thickness 210. FIG. 2C is a diagrammatic, partially cutaway illustration of an embodiment of a fissile neutron detection system generally indicated by the reference number 200″. System 200″ can be identical to system 200 of FIG. 2A with the exception that thickness 214 outward of thermal neutron detector 102B is greater than thickness 212 outward of thermal neutron detector 102A. Applicants recognize that this configuration results in what can be referred to as a unidirectional fissile neutron detection system wherein the sensitivity in first major receiving direction 152a is greater than the sensitivity in second major receiving direction 152b. Further, the sensitivity in first major receiving direction 152a can be greater than the fundamental receiving sensitivity of core fissile neutron detection system 100 in direction 152a, and also greater than the fundamental receiving sensitivity of system 200 in direction 152a. At the same time, the receiving sensitivity in second major receiving direction 152b can be less than (or attenuated) compared to the fundamental receiving sensitivity of core fissile neutron detection system 100 in direction 152b, as well as less than (or attenuated) compared to the fundamental receiving sensitivity of system 200 in direction 152b. In some instances, the sensitivity opposite the primary receiving direction in a unidirectional system can be effectively negligible such that the opposite receiving direction may not be shown in illustrations of embodiments of unidirectional fissile neutron detection systems discussed below. Of course, whether one or the other of the denominative first or second major receiving directions is more sensitive that the other is merely a matter of descriptive nomenclature. While thickness 212 is less than thickness 210 of the moderator, thickness 214 can be greater than thickness 210 of the moderator or at least greater than thickness 212. Applicants have discovered that for a fissile neutron source that is located to one side of a detector system, a unidirectional detection system that applies the teachings that have been brought to light with regard to FIG. 2C can be of benefit. As discussed previously, supplemental moderator layers, substantially thinner than moderator 150 in FIG. 2A, can provide for at least somewhat enhanced detection efficiency. In general, a thin supplemental moderator should be sufficiently thin such that only a negligible number of incoming fissile neutrons are scattered. In other words, incoming fissile neutrons that initially pass through the thin supplemental moderator have a low probability of experiencing scattering in the supplemental moderator. Indeed, no measurement enhancements due to immediate moderation of incoming fissile neutrons, from either of the major receiving directions, should be expected to arise from scattering in thin supplemental moderator layers. However, as discussed previously and as should be appreciated by a person of ordinary skill in the art having this disclosure in hand, for any fissile neutron detection system having no supplemental layers, a certain number of neutrons will escape detection, even after undergoing significant numbers of moderator collisions and despite having been scattered and hence slowed to some degree. Paths 194c, 194e and 194h of FIG. 1F represent a group of at least partially thermalized neutrons that evade detection in the absence of a thin supplemental moderator. As such, this is a representative group of scattered neutrons, scattered by the primary moderator in forward and backward directions, that can be further scattered in significant numbers by the supplemental moderators. In other words, significant numbers of neutrons that have been scattered by moderator 150, and that have nevertheless passed undetected through a thermal neutron detector, can be further scattered back toward that detector by the supplemental layer for subsequent detection. Path 194e′ when compared to corresponding path 194e of FIG. 1F, illustrates a neutron that is backscattered by supplemental moderator 202 and then detected. Path 194f′ in FIG. 2A is representative of other neutrons that are backscattered by supplemental moderator 202 and detected. These events can occur in measurably significant numbers given that these neutrons have already been significantly slowed by moderator 150. Applicants recognize that thin supplemental moderators are generally not intended for scattering incoming fissile neutrons as the fissile neutrons initially enter the fissile neutron detection system. It should be appreciated by a person of ordinary skill in the art, in view of this overall disclosure, that fissile neutron detection systems described herein generally pass undetected a significant number of fissile neutrons with little to no reduction in energy, and that a thin supplemental moderator layer generally will not result in a measurable increase in efficiency with respect to these undetected fissile neutrons that exit the system approximately as fissile neutrons. On the other hand, it is also the case that some neutrons exiting the moderator will pass undetected through the thermal neutron detectors and evade detection, even though these neutrons are subject to at least some thermalization. Applicants submit that the presence of a supplemental moderator does result in a measurable increase in detection with respect to these thermalized neutrons. FIG. 2D is a diagrammatic, partially cutaway illustration of an embodiment of a fissile neutron detection system generally indicated by the reference number 200′″. System 200′″ can be identical to system 200′ of FIG. 2B with the exception that thicknesses 212 and 214 are configured in accordance with embodiment 200″ of FIG. 2C to form a unidirectional detection system in a manner that is consistent with the descriptions above. FIG. 3A depicts simulated performance 300 of an illustrative fissile neutron detection system 200 such as depicted and described in detail with regard to FIG. 2A in which a first neutron detector 102A and a second neutron detector 102B are separated by a neutron moderator 150, in accordance with at least one embodiment of the present disclosure. FIG. 3B illustrates details with regard to a simulation system, generally indicated by the reference number 301, for generating the performance shown in FIG. 3A. The simulated system includes a thickness parameter X for supplemental moderators and a thickness parameter Y for the center moderator of the core structure (i.e., detection system 100). It should be noted that the geometry is not shown to scale and that moderator thickness dimensions X and Y were varied as part of a scan of simulations. Consistent with known modeling and testing techniques that are familiar with those of ordinary skill in the engineering practice of neutron detection, a neutron source 302 is illustrated in a configuration that is thought to correspond, at least within a reasonable approximation, to real world threat sources for emitting fissile neutrons 160. Neutron source 302 is 200 cm from moderator 150. As depicted in FIG. 3B, each of thermal neutron detectors 102 can include a stainless steel, hermetically sealed, chamber 105 that can have a thickness (i.e., a sidewall 120 extent or height) of from approximately 1 centimeters (cm) to approximately 5 cm. Simulations were performed for various 6Li active material 112 sheet thicknesses and neutron moderator 150 thicknesses. Map 310 shows the capture performance of two neutron detector fissile neutron detection system 200 depicted in FIG. 2A as a shaded map of 6Li active material sheet thickness 312 along the x-axis versus neutron moderator thickness Y 314 along the y-axis. In this embodiment, capture performance peaks in a region 320, at a 6Li active material sheet thickness of from approximately 80 micrometers (μm) to approximately 100 μm in combination with a neutron moderator thickness Y of from approximately 3 centimeters (cm) to approximately 4 cm. FIG. 3C is a plot, generally indicated by the reference number 315, in which detection efficiency is plotted against a ratio of thickness Y of the middle moderator to thickness X of the supplemental moderators based on the simulations relating to FIG. 3B. These results were generated by Applicants based on Monte Carlo techniques for device simulation that were first developed by experimental physicists as computational tools used in the course of fundamental research relating to neutron detection. The horizontal axis explicitly gives the values of Y and X in centimeters for each point on the plot. A leftmost point 322 in the plot corresponds to middle moderator thickness Y being zero for a ratio (0.0/2.6). In other words, there is no middle moderator. A rightmost point 324 in the plot corresponds to a zero thickness X of the supplemental moderators for a ratio (5.2/0.0). In other words, only the middle moderator is present. Based on the plot, a peak efficiency occurs at a point 326 where the thickness ratio is given as (4.0/0.6). In other words, moderator 150 is approximately 6.7 times more thick than the supplemental moderators outward of the thermal neutron detectors. Based on FIG. 3C, one of ordinary skill in the art can define a range for the thickness ratio in order to maintain a desired detection efficiency. It should be clear from this plot, however, that the supplemental moderators should be substantially or far thinner than middle moderator 150. At the same time, a point 328 demonstrates that one can configure the middle moderator as too thick which likewise results in reduced detection efficiency. FIG. 4A depicts simulated performance 400 of fissile neutron detection system 200′ such as depicted and described in detail with regard to FIG. 2B in which first neutron detector 102A, second neutron detector 102B, and a third neutron detector 102C are separated by respective neutron moderators 150A and 150B, in accordance with at least one embodiment of the present disclosure. FIG. 4B illustrates details with regard to a simulated system, generally indicated by the reference number 402, for generating the results of FIG. 4A. The simulated system includes a thickness parameter X for supplemental moderators and a thickness parameter Y for center moderators 150A and 150B. It should be noted that the geometry is not shown to scale and that moderator thickness dimensions X and Y were varied as part of a scan of simulations. For purposes of the simulations, honeycomb thickness is equal to 0.3 cm. The thermal neutron detector chamber thicknesses 102A-102C are 2.1 cm and the available moderator thickness of 2X+2Y is equal to 5.1 cm. As depicted in FIG. 4B, each of neutron detectors 102A-102C can include a stainless steel, hermetically sealed, chamber 105 (FIG. 1) that can have a thickness (i.e., a sidewall 120 extent or height, FIG. 1B) of from approximately 1 centimeters (cm) to approximately 2.5 cm. Simulations were performed for various 6Li active material 112 sheet thicknesses and neutron moderator 150 thicknesses. Map 400 shows the capture performance of fissile neutron detection system 200′ depicted in FIG. 2B as a shaded plot of 6Li active material sheet thickness 412 along the x-axis versus neutron moderator thickness Y along the y-axis. In this embodiment, capture performance peaks in a region 420, at a 6Li active material sheet thickness of from approximately 80 micrometers (μm) to approximately 100 μm in combination with a neutron moderator thickness Y of from approximately 1.5 centimeters (cm) to approximately 2.5 cm. FIGS. 5A, 5B, 5C, and 5D depict an embodiment of a thermal neutron detector, generally indicated by the reference number 600, and suitable for use with fissile neutron detection system 100 depicted in FIGS. 1A-1D, in accordance with one or more embodiments described herein. Thermal neutron detector 600 can include top cover 104, bottom cover 108, lithium-6 (6Li) foils 112A and 112B, a number of electrodes 116 (collectively “electrodes 616”), sidewall 120, seal 124, a front electronics board 632, and a back electronics board 636. In this embodiment, a single isolator feedthrough 128 isolates an input conductor 638 that is electrically coupled to an electrically conductive cross-member 640a which, in turn, is electrically connected to plurality of electrodes 116. This electrically conducting cross member may hereinafter be referred to as a “ganging board” and/or a “bus”. Front electronics board 632 and back electronics board 636 can include communicably coupled readout electronics or similar devices to produce a detection signal 638 on an output line 640. A first 6Li foil 112A can be disposed proximate the top cover 104 and a second 6Li foil 112B can be disposed proximate the bottom cover 108. Top cover 104 and bottom cover 108 can be attached or otherwise disposed proximate to sidewalls 120 as shown in FIG. 5A to form chamber 105 that can, in operation, contain a readout gas. Seal 124 provides a seal between top cover 104, sidewalls 120, and bottom cover 108 that can greatly reduce or even prevent the readout gas from escaping from chamber 105. Seal 124 can also greatly reduce or prevent the entry of gases or fluids from external to chamber 105. Electrodes 116 are fed through isolator 128 that can be located on any suitable sidewall 120. Electrodes 116 can be electrically conductively coupled to the front electronics board 632 and can be conductively coupled to the back electronics board 636. The readout electronics can provide a voltage bias between electrodes 116 and the 6Li foils 112A and 112B. In some embodiments, each of the 6Li foils 112A and 112B can have a thickness of approximately 100 micrometers (μm). Additionally, the readout electronics can decouple the signals received from electrodes 116, can amplify the signals received from electrodes 116, and host post-digitization and further computer and wireless interfacing to share information relating to the collected signals with one or more user applications. Upon exposure to fissile material such as plutonium and highly enriched uranium (HEU), neutrons and gamma rays can impinge upon fissile neutron detector system 100. Neutrons impinging upon fissile neutron detector system 100 (FIG. 1) can pass through one or more external moderators 202 and/or one or more neutron moderators 150 prior to impinging on the top plate 104 or bottom plate 108 of the neutron detector 600. External moderator 202 and/or neutron moderators 150 can reduce the energy level of incident fissile neutron 160 (e.g., 100 keV to 10 MeV) to the energy level of a thermal neutron 162 (e.g., 0.025 eV). Thermal neutron(s) 162 can be captured by one of the 6Li atoms in the 6Li foils 112A and 112B. The capture of thermal neutron(s) by the 6Li atom results in a lithium 7 atom that decays into two daughter particles, an alpha particle 166 and a triton 168. Triton 168 and alpha particle 166 travel in opposite directions, and dissipate energy as they travel through 6Li foil 112A and 112B. Referring to FIGS. 1A and 1B, in conjunction with FIG. 5A, upon exiting the 6Li foil, triton 168 and/or alpha particle 166 can have sufficient kinetic energy to ionize atoms present in the readout gas within chamber 105. Electrons 172 liberated during the ionization of readout gas can drift in the direction of electrodes 116 and the ions produced during the ionization of readout gas atoms can drift in the direction of 6Li foils 112A and 112B. Electrons 172 that drift within a predetermined distance of roughly 5 times the radius of electrode 116 (i.e., the Townsend avalanche region) can encounter an electric field that accelerates electrons 172 at a rate sufficient to cause further ionization of the readout gas. The further ionization of the readout gas liberates additional electrons 172, which can drift toward electrodes 116 and cause even further ionization of the readout gas. This process that occurs within Townsend avalanche region 176 is called gas multiplication. Ionized readout gas atoms within the Townsend avalanche region that move towards 6Li foils 112A and 112B cause a movement of charge along electrodes 116. The charge moving along each electrode 116 is collected by readout electronics, on board 632 and/or 636, and amplified with a pulse-mode, charge-sensitive preamplifier to produce a voltage output signal 192. Pulse height discrimination circuitry included with in the readout electronics then compares the voltage output signal to a first predetermined threshold and determines if a fissile neutron event has been detected (e.g., for a gas multiplication of roughly 100, and an amplification circuitry gain of lfC/mV, pulse heights greater than 250 keV can indicate the occurrence of a fissile neutron event). In some embodiments, the false positive detection rate of fissile neutrons 160 based on the first predetermined threshold can be less than 1×10−5 for a gamma ray exposure rate of 100 mR/hr. A second predetermined threshold, lower than the first predetermined threshold, can also be established. Voltage output signals below the second predetermined threshold can be deemed attributable to very low ionizing gamma ray events or movements of charge in thermal neutron detector 600 induced by one or more other sources (e.g., thermal heat, radio frequency electromagnetic radiation, and changes in the relative position of electrodes 116 and the 6Li foils 112A and 112B. Voltage output signals below the first predetermined threshold and above the second predetermined threshold are indicative of gamma ray events. The detected rate of fissile neutrons 160 and gamma rays impinging upon the detector can be used in radiation detection methodologies (e.g., to detect the presence of plutonium or highly enriched uranium). The composition of the readout gas can advantageously remain relatively constant over time to avoid deterioration of the gamma ray and neutron detection process. Changes greater than 1% in the composition of the readout gas can affect the Townsend avalanche process. For example, nitrogen, oxygen, or water molecules that leak into the chamber do not ionize as well as the argon gas in the amplification region near electrodes 116, and therefore can reduce the Townsend avalanche process near electrodes 116 when introduced into the readout gas. This reduces the ability of the readout electronics to distinguish between noise, gamma ray, and fissile neutron events, thereby decreasing the efficiency of neutron detector 600. In some embodiments, seal 124 can be formed from one or more elastomeric materials, such as polyisobutylene, to maintain the readout gas composition within the chamber 105 over extended periods of time (e.g., 30 years). Seal 124 can conform to the region between top cover 104 or bottom cover 108 and sidewalls 120, filling any gaps due to imperfections in the surface quality of top cover 104, bottom cover 108, and sidewalls 120. In some embodiments, the surface quality of top cover 104, bottom cover 108, and sidewalls 120 can be selected to generate uniform electric fields near electrodes 116 of neutron detector 600, with no regard for sealing of top cover 104, bottom cover 108, and sidewalls 120, since the elastomeric nature of seal 124 can accommodate such fluctuations. It is noted that in a given thermal neutron detector, there is no requirement that an integral layer of active sheet material 112 must be used. In other words, an overall layer of active sheet material can be made up, for example, of a plurality of “tiles” of active sheet material or a patchwork of such tiles. FIGS. 6A and 6B depict another illustrative thermal neutron detector, generally indicated by the reference number 700, suitable for use with fissile neutron detection system 100 depicted in FIGS. 1A-1D, in accordance with one or more embodiments described herein. FIG. 6A is a diagrammatic perspective view whereas FIG. 6B is a diagrammatic view, in elevation, taken to show the right edge of the thermal neutron detection in the view of FIG. 6A. Thermal neutron detector 700 includes similar elements as thermal neutron detector 600 depicted in FIGS. 5A-5D. Thermal neutron detector 700 can include an array of vertically elongated structural members 750 that can be posts, for example, having upper and lower feet that serve as bonding pads. Elongated structural members 750 can include a top pad 764, a bottom pad 768, and a post 772. In embodiments, elongated structural members 750 can extend between top cover 104 and bottom cover 108. In some embodiments, elongated structural members 750 can provide structural support that can reduce undesirable mechanical vibrations within thermal neutron detector 700. For example, a tungsten wire electrode 116 having a length of approximately 100 centimeters (cm) and a diameter of approximately 30 micrometer (μm), placed under approximately 250 g of tension has a first vibrational frequency of approximately 200 Hz, which corresponds to significant vibrations typically generated by motor vehicles. Placing a single structural support 750 near the middle of the tungsten wire electrode 116, increases the first vibrational frequency to approximately 420 Hz, thereby reducing vibrations induced by vehicular movement by a factor of 100 or more. Reducing vibrations in thermal neutron detectors with surface areas greater than 1000 square centimeters (cm2) is) advantageous because, as the surface area of neutron detector 700 and the length of electrodes 116 is increased, the increased dimensions can lead to vibrations that can cause changes in the relative position of electrodes 116 and/or active materials 112 disposed in chamber 105. Relative changes in position between electrodes 116 and active materials 112 can cause movement of charge within fissile neutron detection system 100. Such charge displacement within fissile neutron detection system 100 can generate voltage output signals 192 that are indistinguishable from gamma ray or neutron signals. Elongated structural members 750 can reduce mechanical vibrations of top cover 104 and bottom cover 108 by providing a mechanical connection therebetween. For example, adding elongated structural support 750 at the center of a 1 square meter (m2) thermal neutron detector 700 can increase the resonance frequency of electrodes 116 in neutron detector 700 by a factor of two or greater and can reduce the amplitude of the vibrations by a factor of two compared to when top cover 104 and/or bottom cover 108 are supported only at the edges by sidewall 120. The shape of elongated structural members 750 can be selected to minimize vibrations between top cover 104 and bottom cover 108 (e.g., the cross section of the elongate structural members 250 can be a “T”, an “I”, an “L”, an “X”, or a “C”). In some embodiments, the cross section of the elongated structural members 750 can be rectangular. In embodiments, each of the electrodes 116 can pass through a slot or similar aperture that penetrates at least a portion of elongated structural member 750 to reduce the vibration of the electrodes 116. The slots or apertures can provide mechanical support for electrodes 116. In some embodiments, the slot or aperture can be located near a side or edge of elongated structural member 750. Electrode 116 traversing the chamber 105 can pass through multiple slots or apertures. Elongated structural members 750 can be positioned within chamber 105 such that the slots or apertures alternate sides of elongated structural members 750 as electrode 116 traverses chamber 105. In some embodiments, some or all of the elongated structural members 750 can be fabricated using an electrically non-conductive material such as, for example, plastic. In some embodiments, the slot or aperture can be positioned to provide an upward or downward lateral force on electrode 116. In some embodiments, electrodes 116 can be supported by a structural member that attaches to the top cover 104 or the bottom cover 108, but not both. In some embodiments, elongated structural members 750 can contact top cover 104 and bottom cover 108, but do not include a slot or aperture and are displaced from electrodes 116 so as to not cause a mechanical interference. FIG. 6C illustrates a modified form of thermal neutron detector 700, generally indicated by the reference number 700′ and taken from the same viewpoint as FIG. 6B. The structure is the same as that of FIG. 6B with the exception that a modified support member 750′ is used. Support member 750′ can be of any suitable shape including a mounting pad 780 from which a post 782 extends. A cap 784 that secures to 782 is positioned on top of post 782 to capture electrode 116. Like support member 750, support member 750′ can be formed from an electrically insulative material such as plastic. FIG. 7 is a flow diagram of an embodiment of a method, generally indicated by the reference number 800, for detecting fissile neutrons 160 using fissile neutron detector 100 that includes at least one thermal neutron detector 102, and at least one neutron moderator 150 disposed proximate thermal neutron detector 102, in accordance with at least one embodiment of the present disclosure. High-energy fissile neutrons such as those emitted by plutonium and highly enriched uranium (HEU) provide a tell-tale indicator of the presence of such materials. Fissile neutrons can have energy levels that exceed 100 keV. At such energy levels, a large percentage of fissile neutrons can pass undetected through active material 112 typically found in thermal neutron detectors 102, 600, 700 and 700′. The presence of a neutron moderator, such as neutron moderator 150, can beneficially reduce the energy level of fissile neutrons to the energy level of thermal neutrons, approximately 0.025 eV. Such a reduction in energy level can be at least partially attributable to collisions between fissile neutrons 160 and hydrogen nuclei in moderator 150. Consequently, moderators having a high percentage of hydrogen by weight can be preferable. Thermal neutrons 162 can impact active material 112, causing the spontaneous formation of charge-carrying daughter particles, such as alpha-particles 166 and tritons 168. Triton 168 can ionize a gas within chamber 105 of neutron detector 102, 600, 700 and 700′. The presence of the ionized gas and dissociated electrons 172 within chamber 105 can induce a current flow on an electrode in the chamber that is maintained at a potential. The current flow can be proportional to the number or rate at which fissile neutrons 160 are impacting active material 112 within chamber 105. Method 800 commences at 802. At 804, the energy level of at least a portion of fissile neutrons 160 incident upon fissile neutron detection system 100 is reduced to the energy level of thermal neutron 162. In some implementations, this reduction in energy level is accomplished using moderator(s) 150. Such neutron moderators 150 can include a number of interstitial neutron moderators 150 that are positioned proximate first thermal neutron detector 102A and second thermal neutron detector 102B and within a space formed between the between first thermal neutron detector 102A and second thermal neutron detector 102B. Such neutron moderators 150 can include one or more neutron moderators 150 having an exterior side and in which at least a portion of the exterior side is disposed proximate thermal neutron detector(s) 102. Fissile neutron detection systems can include one or more external neutron moderators 202 positioned proximate an exterior surface of first thermal neutron detector 102A, second thermal neutron detector 102B, and/or intermediate moderators 150A-150n. Neutron moderator(s) 150 which reduce the energy level of incident fissile neutrons 160 can include one or more solids, liquids, and/or compressed gases capable of reducing the energy level of at least some of incident fissile neutrons 160. In some implementations, neutron moderator(s) 150 can include materials, compounds, or substances having a significant hydrogen concentration—greater than approximately 10 percent atoms per cm3 hydrogen. The impact between incident fissile neutrons 160 and the hydrogen nuclei within the neutron moderator(s) 150 can reduce the energy level of incident fissile neutron 160 to that of thermal neutron 162 which then exits neutron moderator(s) 150. Due to the random nature of the collisions within neutron moderator(s), a portion of incident fissile neutrons 160 can flow as neutrons having an energy level at or above that of thermal neutron 162 from neutron moderator(s) 150 in a direction that precludes impingement on a thermal neutron detector 102, 600, 700, 700′ disposed proximate at least a portion of the exterior side of neutron moderator 150. For example, incident fissile neutron 160 can flow from the “side” or “edge” of neutron moderator(s) 150 in a direction along a vector pointing away from a thermal neutron detector 102, 600, 700, 700′ that is proximate at least a portion of the side of neutron moderator(s) 150. The geometry of fissile neutron detection system 100, the geometry and composition of neutron moderator 150, the geometry and composition of external neutron moderator 202, and the construction and geometry of the thermal neutron detector 102, 600, 700 and 700′ all play a role in determining the capture rate of incident fissile neutrons 160. For example, planar neutron detectors such as neutron detector 102 depicted in FIGS. 1A-1E, neutron detector 600 depicted in FIGS. 5A-5D, neutron detector 700 depicted in FIGS. 6A-6B and neutron detector 700′ of FIG. 6C all present a significantly increased cross-sectional capture area which provides a marked advantage and improvement in fissile neutron detection performance and accuracy over straw-type neutron detectors in which detector “straws” can be positioned within a block of neutron moderator. At 806, at least some of thermal neutrons 162 exiting neutron moderator(s) 150 can pass through top cover 104 or bottom cover 108 of first thermal neutron detector 102A and enter chamber 105A or pass through top cover 104 or bottom cover 108 of second thermal neutron detector 102B and enter chamber 105B. Once inside of chamber 105, thermal neutron 162 can impinge on one or more active materials 112 disposed therein. Active material 112 can include any substance, isotope, element, compound, or mixture capable of generating charge-carrying particles upon exposure to thermal neutrons 162. Non-limiting examples of such active materials include, but are not limited to, lithium-6 (6Li); boron-10 (10B); and helium-3 (3He). Such active materials 112 can be present in chamber 105 in one or more forms. For example, in some implementations, 6Li in the form of thin (50 μm to 150 μm) sheets can provide all or a portion of active material 112 that is disposed either at one or more intermediate points (e.g., thermal neutron detector 102) or proximate one or more interior surfaces (e.g., thermal neutron detector 600) of chamber 105. In some implementations, 10B in the form of a thin layer disposed on at least a portion of the interior surface of chamber 105 can provide all or a portion of active material 112. In some implementations, 3He in the form of a gas disposed in chamber 105 can provide all or a portion of active material 112. The charge-carrying particle(s) emitted by active material 112 in response to the impact of thermal neutron 162 can travel into readout gas 170 disposed within chamber 105. The charge-carrying particles, such as triton 168, can ionize a portion of readout gas 170, creating a positively charged readout gas ion and electron 172. At 808, neutron detector 102, 600, 700, 700′ in response to the charged particles generated by the impact of thermal neutron 162 on active material 112, generates a current indicative of a number of thermal neutrons 162 that impact active material 112 or a rate at which thermal neutrons 162 impact active material 112 in the respective neutron detector 102, 600, 700, 700′. The method 800 concludes at 810. FIG. 8 is a flow diagram of an embodiment of a method, generally indicated by the reference number 900, for generating a current in thermal neutron detector 102, 600, 700, 700′ in response to the impact of thermal neutron 162 on an active material such as 6Li or 10B disposed in chamber 105 of thermal neutron detector 102, 600, 700, 700′ in accordance with at least one embodiment of the present disclosure. The interaction of the charged particles, such as triton 168, generated by the impact of thermal neutron 162 on active material 112, with readout gas 170 disposed in chamber 105 can cause a current to flow on electrode 116 placed in chamber 105. Electrode 116 can be maintained at a potential that differs from the potential of active material 112. Method 900 commences at 902. At 904, one or more charged particles can be generated by the capture of thermal neutron 162 by active material 112. In implementations using 6Li, these charged particles can include alpha-particle 166 (two protons and two neutrons) and triton 168 (one proton and two neutrons). In implementations, triton 168 can travel a distance of up to 135 μm through a 6Li sheet of active material 112. Thus, within thermal neutron detectors 102, 600, 700, 700′ using 6Li active materials, the thickness of a 6Li sheet of active material 112 can be maintained at less than 135 μm to increase the probability that triton 168 will escape active sheet material 112. In implementations using 10B, these charged particles can include an alpha particle and a 7Li ion. Approximately 78% of the time either of the alpha particle or the 7Li ion can escape a 1 μm thick layer of 10B. Thus, within thermal neutron detectors 102, 600, 700, 700′ using 10B active materials, the 10B is typically applied as a coating or layer to all or a portion of the interior surfaces of chamber 105. At 906, the charged particles escaping active material 112 ionize at least a portion of readout gas 170 disposed within chamber 105. In some implementations, readout gas 170 can include an elemental gas, a gas mixture, a gas combination, a gas compound, or any other combination of gases. In some implementations, readout gas 170 can include one or more noble gases, such as argon (Ar). In 6Li implementations, at least a portion of alpha particles 166 and/or at least a portion of tritons 168 can ionize a portion of readout gas 170, generating drift electrons 172 and a positively charged readout gas ion. In 10B implementations, at least a portion of alpha particles 166 and/or at least a portion of the 7Li particles can ionize a portion of readout gas 170, generating drift electrons 172 and a positively charged readout gas ion. At 908, electrode 116 placed in chamber 105 can be maintained at a potential that differs from the potential of active material 112. In some instances, electrode 116 can be maintained at a potential that is positive (e.g., +1100 V) measured with respect to the potential of active material 112 (e.g., grounded or 0 V). The electric field created within chamber 105 can cause drift electrons 172 to drift or travel towards electrode 116. The electric field created within chamber 105 can also cause the positively charged readout gas ions to drift or travel towards active material 112. As drift electrons 172 travel and/or accelerate toward electrode 116, additional ionization of readout gas 170 can occur. This “chain reaction” of ionization of readout gas 170 can, in turn, cause an avalanche of electrons 174 within amplification region 176 about electrode 116. At 910, the combined flow of positively charged readout gas ions toward active material 112 and the flow of drift electrons 172 toward electrode 116 causes an overall charge flow within chamber 105. This flow of charges within chamber 105 can induce a current in electrode 116. In some instances, the magnitude of the current in the electrode can be correlated with the number of thermal neutrons 162 that impact active material 112 and/or the rate at which thermal neutrons 162 impact active material 112. Method 900 concludes at 912. FIG. 9 is a flow diagram of an embodiment of a method, generally indicated by the reference number 1000, for generating a current in thermal neutron detector 102, 600, 700, 700′ in response to the impact of thermal neutron(s) 162 on an active material such as 3He or boron trifluoride (BF3) disposed in chamber 105 of thermal neutron detector 102, 600, 700, 700′ in accordance with at least one embodiment of the present disclosure. The interaction of the charged particles generated by the interaction between thermal neutron 162 and a gaseous active material 112 such as 3He disposed in the chamber 105 can cause a current to flow on electrode 116 disposed within chamber 105. In some instances, electrode 116 can be maintained at a potential that differs from the potential elsewhere in chamber 105 and different from the potential of gaseous active material 112. Method 1000 commences at 1002. At 1004, one or more charged particles can be generated by the capture of thermal neutron 162 by active material 112. In implementations using 3He, these charged particles can include a proton and a triton. At 1006, electrode 116 placed in chamber 105 can be maintained at a potential that differs from the potential of active material 112. In some instances, electrode 116 can be maintained at a potential that is positive (e.g., +1100V) measured with respect to the potential of active material 112 (e.g., grounded or 0 V). The electric field created within chamber 105 can cause the charged particles to drift or travel towards electrode 116. At 1008, the flow of charged particles toward electrode 116 causes an overall charge flow within chamber 105. This flow of charges within chamber 105 can induce a current in electrode 116. In some instances, the magnitude of the current in the electrode can be correlated with the number of thermal neutrons 162 that impact active material 112 and/or the rate at which thermal neutrons 162 impact active material 112. Method 1000 concludes at 1010. FIG. 10A depicts a diagrammatic, partially cutaway view, in elevation, of an embodiment of a fissile neutron detection system, generally indicated by the reference number 1100a, including thermal neutron detector 102 and neutron moderator 150. In the embodiment depicted in FIG. 10A, thermal neutron detector 102 is formed into a ring-like structure that at least partially surrounds neutron moderator 150. In such an arrangement, incident high-energy fissile neutrons 160 most often pass through thermal neutron detector 102 and impinge upon neutron moderator 150. Within neutron moderator 150, the energy level of at least some of fissile neutrons 160 can be reduced to an energy level of thermal neutron 162. At least a portion of thermal neutrons 162 can exit the neutron moderator 150 and enter the neutron detector 102. In accordance with the teachings of this disclosure, the embodiment of the thermal neutron detector arrangement of FIG. 10A can be configured to include an active sheet layer arrangement 112 that surrounds moderator 150 such that a majority of the incident fissile neutrons pass through the active sheet layer arrangement prior to impinging on moderator 150, and (ii) a majority of thermal neutrons impinge on the active sheet layer arrangement after exiting the moderator arrangement. As will be further discussed and as also shown in subsequent figures, thermal neutron detectors, in accordance with the present disclosure, can include a first set of electrodes 116A on one side of the active sheet layer arrangement and another set of electrodes 116B on an opposite side of the active sheet layer arrangement. FIG. 10B is a diagrammatic, partially cutaway view, in elevation, of an embodiment of a fissile neutron detection system, generally indicated by the reference number 1100b, including thermal neutron detector 102 and neutron moderator 150. In the embodiment depicted in FIG. 10B, a number of thermal neutron detectors 102A-102D at least partially surround an exterior periphery or peripheral outline of neutron moderator 150. In such an arrangement, incident high-energy fissile neutrons 160 can pass through any one of the number of neutron detectors 102A-102D and thereafter impinge upon neutron moderator 150. Within neutron moderator 150, the energy level of at least some of fissile neutrons 160 can be reduced to an energy level of thermal neutron 162. At least a portion of thermal neutrons 162 can exit neutron moderator 150 and enter one of thermal neutron detectors 102A-102D. FIG. 10C is a diagrammatic, partially cutaway view, in elevation, of an embodiment of a fissile neutron detection system, generally indicated by the reference number 1100c, including thermal neutron detector 102 and neutron moderator 150. In the embodiment depicted in FIG. 10C, a number of thermal neutron detectors 102A-102B at least substantially surround an exterior periphery or peripheral outline of neutron moderator 150. Although depicted as a planar body, in such an arrangement, each of neutron detectors 102 can include a planar body, a curved body, or an angular body. As depicted in FIG. 10C, first thermal neutron detector 102A can be spaced a first distance 1102 from at least a portion of an exterior side (i.e., major surface) of neutron moderator 150 and a second thermal neutron detector 102B can be spaced a second distance 1104 from at least a portion of an opposing exterior side (i.e., major surface) of neutron moderator 150. In accordance with the general teachings of this disclosure, each thermal neutron detector arrangement of FIG. 10C can be configured to include an active sheet layer arrangement 112 that spans at least a majority of the lateral extents of that detector. FIG. 10D is a diagrammatic, partially cutaway view, in elevation, of an embodiment of a fissile neutron detection system, generally indicated by the reference number 1100d, including thermal neutron detector 102 and neutron moderator 150. In the embodiment depicted in FIG. 10D, neutron moderator 150 at least partially surrounds at least a portion of an exterior surface of first thermal neutron detector 102A and an exterior surface of second thermal neutron detector 102B. Gaps or voids are present around each thermal neutron detector. These gaps can contain, for example, air, or materials that secure the thermal neutron detectors to the moderator. Horizontal gaps 1106 above and below each major surface of the thermal neutron detectors are sufficiently limited in vertical extents so as to have a limited effect on detection performance. Vertical gaps 1108 are wider than horizontal gaps but are limited in volume and are less critical than the horizontal gaps due to positioning outward of the thermal neutron detectors so as to be less interactive with active material 112 within each of the thermal neutron detectors. FIG. 10E is a diagrammatic, partially cutaway view, in elevation, of an embodiment of a fissile neutron detection system, generally indicated by the reference number 1100e, including thermal neutron detector 102 and neutron moderator 150. In the embodiment depicted in FIG. 10E, first thermal neutron detector 102A and second thermal neutron detector 102B are disposed in an alternating or “sandwich” arrangement with first neutron moderator 150A, second, center thermal neutron moderator 150B and a third, thermal neutron moderator 150C. As depicted in FIG. 10E, moderators 150A and 150C are substantially thinner than moderator 150B for purposes of enhancing detection performance. FIG. 10F is a diagrammatic, partially cutaway view, in elevation, of an embodiment of a fissile neutron detection system, generally indicated by the reference number 1100f, including thermal neutron detector 102 and neutron moderator 150. In the embodiment depicted in FIG. 10F, first thermal neutron detector 102A and second thermal neutron detector 102B are disposed in a moderator region defined between the thermal neutron detectors having a limited gap between each thermal neutron detector and the moderator such that the moderator is at least substantially surrounded by the arrangement of detectors. It is noted that the exposed ends and/or sides (or even the entire peripheral edge configuration) of moderator 150B can be considered as gaps in an overall arrangement of thermal neutron detectors. These gaps, however, have a limited effect on detection performance at least due to their locations outward of the active material in the thermal neutron detectors and their limited extents as part of the periphery of the moderator region defined by the cooperation of thermal neutron detectors 102A and 102B. FIG. 10G is a diagrammatic, partially cutaway view, in elevation, of an embodiment of a fissile neutron detection system, generally indicated by the reference number 1100g, including thermal neutron detector 102 and neutron moderator 150. In the embodiment depicted, a unidirectional fissile neutron detector is formed wherein reception from the top side including, major receiving direction 152a, is enhanced with respect to an opposite set of receiving directions. The enhanced sensitivity is due to one thickness 1110 of moderator 150 above thermal neutron detector 102A, in the view of the figure, being less than another thickness 1112 of the moderator below thermal neutron detector 102B. FIG. 10H is a diagrammatic, partially cut-away view, in elevation, of an embodiment of an illustrative fissile neutron detection system, generally indicated by the reference number 1100h, including neutron moderator 150 and thermal neutron detector 102. In the embodiment depicted in FIG. 10H, first thermal neutron detector 102A and second thermal neutron detector 102B are disposed in an alternating or “sandwich” arrangement with first neutron moderator 150A and second neutron moderator 150B. At least a portion of the surface of first neutron detector 102A is exposed. In this way, a unidirectional fissile neutron detection system is formed at least with a relatively higher detection sensitivity in receiving directions from the side of the detection arrangement on which first major receiving direction 152a is shown. In embodiments, moderator 150B can be up to 10 cm thick. FIG. 10I is a diagrammatic, partially cut-away view, in elevation, of an embodiment of an illustrative fissile neutron detection system, generally indicated by the reference number 1100i, including neutron moderator 150 and thermal neutron detector 102. In the embodiment depicted in FIG. 10I, first thermal neutron detector 102A and second thermal neutron detector 102B are disposed in an alternating or “sandwich” arrangement with first neutron moderator 150A, second neutron moderator 150B and third neutron moderator 150C. Based on the foregoing discussions, it should be appreciated that a unidirectional fissile neutron detection system is formed at least with a relatively higher detection sensitivity in receiving directions from the side of the detection arrangement on which first major receiving direction 152a is shown, at least for the reason that moderator 150a is thinner than moderator 150C. In embodiments, first supplemental moderator 150A can have a thickness that is in a range from 0.1 cm to 1 cm, inclusively, and second supplemental moderator third supplemental moderator thickness 150C can be in a range from 1.1 cm to 10 cm, inclusively. FIG. 10J is a diagrammatic, partially cut-away view, in elevation, of an embodiment of illustrative fissile neutron detection system, generally indicated by the reference number 1100j, including neutron detector 102 and neutron moderator 150. In the embodiment depicted in FIG. 10J, two neutron detectors 102A-102B are disposed to at least substantially surround an exterior peripheral outline of a non-planar neutron moderator 150A. In such an arrangement, some or all of neutron detectors 102 can have a planar body disposed about at least a portion of the exterior peripheral outline of non-planar neutron moderator 150, as will be further described. FIG. 10K is a diagrammatic, partially cut-away view, in cross elevation, of an embodiment of an illustrative fissile neutron detection system, generally indicated by the reference number 1100k, including thermal neutron detector 102 and neutron moderator 150, in accordance with at least one embodiment of the present disclosure. In the embodiment of FIG. 10K, six thermal neutron detectors 102A-102F are disposed to at least substantially surround an exterior peripheral outline of a non-planar neutron moderator 150. In such an arrangement, some or all of neutron detectors 102 can have a contoured body or shell configured to closely approximate a surface contour of at least a portion of the exterior periphery of non-planar neutron moderator 150. Stated in another way, the detector arrangement can define a moderator region that is at least generally complementary to the outer peripheral outline of moderator 150A. Based on this configuration, interstitial gaps or voids between the neutron moderator and the thermal neutron detectors can be reduced or even eliminated, even though the neutron moderator is essentially elliptical in configuration, as shown. It should be appreciated that any suitable number of thermal neutron detectors can be used such that the view of FIG. 10J will be obtained in any plane that bisects the detection system. In some embodiments, moderator can include a suitable closed shape such as circular or ovoid in a plan view and thermal neutron detectors 102 can be pie or wedge shaped. FIG. 10L is a diagrammatic, partially cut-away view, in elevation of an embodiment of a fissile neutron detection system, generally indicated by the reference number 11001, including an arrangement of thermal neutron detectors 102 and an arrangement of neutron moderators 150, in accordance with at least one embodiment of the present disclosure. In the embodiment of FIG. 10L, thermal neutron detectors 102A and 102B at least partially surround an exterior peripheral outline of a plurality of neutron moderators 150A-150D. Although each thermal neutron detector is shown as a planar body, in such an arrangement, each of thermal neutron detectors 102 can include a planar body, a curved body, an angular body or some other suitable shape. Moderators 150A-150D are shown in a side-by-side relationship in physical contact with one another, although physical contact is not a requirement. FIG. 10M is a diagrammatic, partially cut-away view, in elevation, of an embodiment of a fissile neutron detection system, generally indicated by the reference number 1100m, including an arrangement of thermal neutron detectors 102 and an arrangement of neutron moderators 150, in accordance with at least one embodiment of the present disclosure. In the embodiment depicted in FIG. 10M, a center moderator arrangement 1118, made of a plurality of individual moderators 150A-150D, is thicker than one either of a first supplemental moderator arrangement 1120 and a second supplemental moderator arrangement 1122. It is noted that any suitable number of one or more moderators can make up the center moderator arrangement. First supplemental moderator arrangement is made up of moderators 150E-150 H while second supplemental moderator arrangement 1122 is made up of moderators 150I-150L, although either supplemental moderator arrangement can be made up or any suitable number of one or more moderators. Moreover, the moderators of each supplemental moderator arrangement are shown in a side-by-side relationship in physical contact with one another, although physical contact is not a requirement. It should be appreciated that the structure of detection system 1100m provides for bidirectional reception based on two major opposing receiving directions. FIG. 10N is a diagrammatic, partially cut-away view, in elevation, of an embodiment of a fissile neutron detection system, generally indicated by the reference number 1100n. The structure of system 1100n reflects the structure of system 1100m of FIG. 10M with the exception that a first supplemental moderator arrangement 1120′ is relatively thinner than a second supplemental moderator arrangement 1122′ such that a unidirectional fissile neutron detection system is formed with a relatively higher detection sensitivity in receiving directions from the side of the detection arrangement on which first major receiving direction 152a is shown. FIG. 10P is a diagrammatic, partially cut-away view, in elevation, of an embodiment of a fissile neutron detection system, generally indicated by the reference number 1100p. The structure of system 1100p reflects the structure of system 1100f of FIG. 10F with the exception that a first group of thermal neutron detectors 102A-102D replace thermal neutron detector 102A and a second group of thermal neutron detectors 102E-102H replace thermal neutron detector 102B of fissile neutron detection system 1100f. Applicants recognize that the use of a plurality of smaller surface area thermal neutron detectors to replace what would otherwise be a relatively large surface area thermal neutron detector provides benefits in terms of maintaining detection performance and avoiding damage that can result from a pressure differential between ambient pressure and the lower pressure within the sealed chamber that is defined by each detector. In this regard, any suitable number of side-by-side thermal neutron detectors can be used and there is no requirement that the same number must be used in each group of thermal neutron detectors, as is likewise the case with respect to the figures which follow. Moreover, a group of moderators can likewise be used in place of moderator 150A. Given that each thermal neutron detector 102 includes an active material layer 112, as is the case in this embodiment as well as all embodiments described herein that use multiple thermal neutron detectors, it should be appreciated that the active material layers of the thermal neutron detectors cooperate to substantially surround the moderator or moderator arrangement at least for the reason that the thermal neutron detectors themselves substantially surround the moderator or moderator arrangement. Thus, the active material layers of the thermal neutron detectors cooperate to span at least a majority of the lateral extents of the moderator arrangement. In this regard, the lateral extents of the active sheet material of each thermal neutron detector is nearly coextensive with the lateral extents of the thermal neutron detector itself, which limits escape avenues for thermal neutrons to evade detection. FIG. 10Q is a diagrammatic, partially cut-away view, in elevation, of an embodiment of a fissile neutron detection system, generally indicated by the reference number 1100q. The structure of system 1100q reflects the structure of system 1100e of FIG. 10E with the exception that a first group of thermal neutron detectors 102A-102D replace thermal neutron detector 102A and a second group of thermal neutron detectors 102E-102H replace thermal neutron detector 102B of fissile neutron detection system 1100e. Thus, the descriptions herein with respect to the use of a group of thermal neutron detectors in place of a single thermal neutron detector are equally applicable here. FIG. 10R is a diagrammatic, partially cut-away view, in elevation, of an embodiment of a fissile neutron detection system, generally indicated by the reference number 1100r. The structure of system 1100r reflects the structure of system 1100i of FIG. 10I with the exception that a first group of thermal neutron detectors 102A-102D replace thermal neutron detector 102A and a second group of thermal neutron detectors 102E-102H replace thermal neutron detector 102B of fissile neutron detection system 1100i. Thus, the descriptions herein with respect to the use of a group of thermal neutron detectors in place of a single thermal neutron detector are equally applicable here. FIG. 10S is a diagrammatic, partially cut-away view, in elevation, of an embodiment of a fissile neutron detection system, generally indicated by the reference number 1100s. The structure of system 1100s reflects the structure of system 1100p of FIG. 10P with the exception that side moderators 1130A-1130F are positioned between adjacent ones of thermal neutron detectors 102A-102H. It should be appreciated that side moderators can be employed in any embodiment described herein that includes side-by-side thermal neutron detectors. The side moderators can be formed from the same material as center moderator 150A, although this is not a requirement. Further, the side moderators can be integrally formed with center moderator 150A or individually formed. While gaps 1132 are illustrated between center moderator 150A and the thermal neutron detectors and between center moderator 150A and side moderators 1130, it should be noted that such gaps may not have an appreciable effect on detection efficiency if the gaps are sufficiently narrow, but in any case are not required. Sufficiently narrow gaps can also be present between the side moderators and adjacent thermal neutron moderators depending on desired detection efficiency, but are not required. In embodiments, the side moderators are no more than 5 cm thick between adjacent ones of the thermal neutron detectors. In some embodiments, the thickness of side moderators can be in a range from 1 cm to 5 cm, inclusively. As is illustrated in FIGS. 10C, 10D, 10E, 10F, 10G, 10H, 10I, 10J, 10K, 10L, 10M, 10N, 10P, 10Q, 10R and 10S, in accordance with the general teachings of this disclosure, each thermal neutron detector arrangement of these embodiments can be configured to include an active sheet layer arrangement that spans at least a majority of the lateral extents of that detector. As illustrated in FIGS. 10A-10N and 10P-10S, in accordance with descriptions above, the major, opposing sides of the moderator arrangements can be in direct physical contact with one of the major surfaces of each thermal neutron detectors. In some implementations, in a manner consistent with the subject illustrations, an air gap (or some other form of void space disposed between the moderator and the detector arrangement) can exist between the moderator arrangement and the neutron detectors. Applicants recognize that, in the context of the described embodiments, the distance between the moderator or moderator arrangement and the thermal neutron detectors can generally be minimized or at least reduced in order to ensure that thermal neutrons do not escape from the detection system after exiting the moderator. Stated in another way, reducing the gap between the moderator and detector arrangement can provide for a minimum or at least reduced number of thermal neutron detectors to minimize or at least reduce the probability of escape of thermal neutrons from the detection system after exiting the moderator. In a manner consistent with minimizing these gaps, embodiments that are within the scope of the present disclosure can at least substantially fill the volume of a moderator region (i.e., greater than 50 percent) that is defined between the thermal neutron detectors with moderating material, as is illustrated in the subject figures. In one embodiment, at least 60 percent of the volume of the moderator region is filled by moderating material. In this regard, a moderating arrangement can include a single member or multiple members of moderating material. Interstitial gaps between multiple members, for example those of FIG. 10S in which side moderators are disposed, do not contribute to the filled volume. As mentioned previously, with this disclosure in hand it should be clear that excessive thermal neutron detector-to-moderator spacing will generally reduce overall efficiency. While not intending to be bound by theory, Applicants submit that excessive spacing (i.e., an excessive moderator detector gap) can result in excess space, between detector arrangements, that is not filled by moderating material and that this provides what may be referred to as sideways or grazing escape paths for thermal neutrons to exit from moderating material without thereafter entering a thermal neutron detector to thereby evade detection. For at least this reason, a person of ordinary skill in the art, having this disclosure in hand, should appreciate that for the embodiments described herein detection efficiency is typically enhanced when the volume or region defined by the detector arrangement is at least substantially filled by the moderator, as opposed to cases where moderator-detector spacing results in significant void or otherwise unfilled space within the envelope defined by the inner periphery of the thermal neutron detector arrangement. It is noted that all of the embodiments of the subject figures are illustrated in a manner that is consistent with small moderator detector gaps for filling the majority of the envelope defined by the inner periphery of the thermal neutron detector arrangement. In FIGS. 10A-10N and 10P-10S, the neutron moderator arrangement is substantially surrounded by the detector arrangement at least for the reason that a majority of the major opposing sides of neutron moderator 150 are in a direct confronting relationship with the surrounding thermal neutron detectors and, more particularly, in such a confronting relationship with surrounding active sheet arrangement 112. In this regard, it is submitted that a person of ordinary skill in the art, having this disclosure in hand, will appreciate that the active sheet layer of each thermal neutron detector tends to be a particularly consequential feature. While other thermal neutron detector features can provide for crucial functions, including but not limited to supporting the active sheet layer and sealing the detection gas, the active sheet layer serves a particularly noteworthy function of actively initiating the fundamental physical processes necessary for detection. It should be further appreciated that the detector arrangements described herein tend to be thin, at least somewhat planar and are laterally spanned by active sheet material. Therefore, much of the descriptive terminology directed herein towards thermal neutron detectors can at least generally be considered as applicable with respect to the active sheet layer(s), such that various aspects of the description, directed towards thermal neutron detector arrangements, are at least generally applicable with respect to the active sheet layer(s) when considered independent of the remaining physical structure of the thermal neutron detectors. As one example, the advantages associated with “surrounding” the moderator arrangement with the thermal detector arrangement can just as accurately be regarded as surrounding of the moderator with active sheet material. In other words, the active sheet material layer(s) surround the moderator arrangement. As another example, close spacing between the moderator and the thermal neutron detectors can be regarded as facilitating close spacing between the moderator and the active sheet layer(s) of the thermal neutron detector(s) at least for the reason that the active sheet layer(s) can be regarded as actively initiating physical processes necessary for detection. While there is a benefit associated with the active sheet layer arrangement closely surrounding a moderator, in terms of detecting thermal neutrons that might otherwise evade detection, for example, based on grazing angles, Applicants recognize that it can be advantageous for the electrode arrangement to include first and second sets of electrodes in a spaced apart confronting relationship, with the active sheet layer arrangement disposed therebetween, such that both sets of electrodes are distributed across at least a majority of the active sheet arrangement. Applicants submit that the detection efficiency experiences an overall enhancement despite the introduction of an additional set of electrodes between the active sheet material layer of each thermal neutron detector and the moderator. Without intending to be bound by theory, Applicants believe that the enhanced detection capability is a result of allowing for generating electrical current on both sides of the active sheet material responsive to ionization of the readout gas disposed within chamber 105. Moreover, the relative increase in spacing that is needed in the direction transverse to the major surfaces of the active sheet material layer to accommodate the second electrode set is small since the active sheet material already must be supported in a spaced apart relationship from the housing of the thermal neutron detector. In some embodiments, for example FIGS. 10B-10J, 10L, and 10N the fissile neutron detection system includes first and second thermal neutron detectors that each support an active sheet layer that spans at least a majority of the lateral extents of the moderator arrangement, and each of the thermal neutron detectors includes first and second electrode arrangements in a spaced apart confronting relationship with the active sheet layer disposed therebetween such that the electrodes of the first electrode arrangement are laterally spaced apart proximate to one of a pair of opposing major surfaces of the active sheet layer and the electrodes of the second electrode arrangement are laterally spaced apart proximate to the other, opposite one of the opposing pair of major surfaces with a projection of each of the first and second arrangements of electrodes onto the active sheet material layer defining an area that substantially covers one of the major surfaces of the active sheet material layer. In other embodiments, for example FIGS. 10K, 10P, 10Q, 10R, and 10S, the fissile neutron detection system can include at least one group of thermal neutron detectors with the thermal neutron detectors of each group in a side-by-side relationship. In these embodiments, each one of the thermal neutron detectors of the group can include first and second sets of electrodes in a spaced apart confronting relationship and an active sheet material layer disposed therebetween with the electrodes of the first electrode set laterally spaced apart proximate to one of a pair of opposing major surfaces of the active sheet layer and the electrodes of the second electrode set laterally spaced apart proximate to the other, opposite one of the opposing pair of major surfaces such that a projection of each of the first and second sets of electrodes onto the active sheet material layer defines an area that substantially covers one of the major surfaces of the active sheet material layer. It is noted that the configurations depicted in FIGS. 10A-10N and 10P-10S employ a thermal neutron detector arrangement in which thermal neutron detector 102, 600, 700, 700′ generally surrounds neutron moderator 150 can provide significant advantages over prior fissile neutron detector designs. Such prior fissile neutron detector designs can be classified as either TYPE I arrangements in which the moderator surrounds the thermal neutron detector and TYPE II arrangements in which the moderator is interspersed among an array of detectors that are separated from each other by substantial amounts along various directions. With regard to TYPE I arrangements, back scattered neutrons that initially strike the moderator can be directed away from the inner detector such that the backscattered neutrons are lost to detection. With respect to the configurations disclosed herein, it is noted that the great majority of such backscattered neutrons can be collected by thermal neutron detectors 102, 600, 700, 700′ as a consequence of the manner in which the detector arrangement can directly surround the moderator in almost all directions. With regard to TYPE II arrangements, it is noted that the disclosed embodiments advantageously reduce the distance that backscattered neutrons travel prior to impacting a neutron detector. In contrast, even optimized TYPE II arrangements can be handicapped by the distances backscattered neutrons must travel prior to detection. While TYPE II systems can allow for an increasing number of neutron detectors as a way of increasing overall system efficiency, such additional detectors generally push the system as a whole toward a heavier, bulkier, and more expensive solution than the embodiments described herein. The system geometries disclosed herein offer significant improvements in space, weight and cost when compared to traditional TYPE II systems. In reference to FIGS. 10A-10N and 10P-10S, for purposes of descriptive clarity, attention is directed to several aspects of performance as delineated from prior technical discussions. The fissile neutron detection systems of the present disclosure are arranged such that a neutron moderator 150 is surrounded by a thermal neutron detector 102, 600, 700, 700′ arrangement such that incoming fissile neutrons 160 generally pass through neutron detector 102, 600, 700, 700′ before striking the neutron moderator 150, and the vast majority (>60%) of all thermal neutrons 162 scattered by neutron moderator 150 (including back scattered as well as forward scattered thermal neutrons) will be collected by neutron detector 102, 600, 700, 700′. Furthermore, in any of the described embodiments, the moderator can, in some implementations, define a generally planar geometry (not necessarily a flat plane) having a large area (anywhere from 0.5 m2 to 10 m2) and a thickness that is small compared with any given lateral extent thereof, and the at least one neutron detector 102, 600, 700, 700′ surrounds the neutron moderator 150 in close proximity especially over the major planar sides. Thus fissile neutron detection system 100 can be considered as a layered arrangement that provides for advantages over prior fissile neutron detection systems. Both forward and backward scattered neutrons travel only short distances before impinging on thermal neutron detectors 102, 600, 700, 700′ under the constraint that the moderator arrangement at least substantially fills the volume of the moderator region that is defined by the thermal neutron detector arrangement, for minimizing or at least reducing the moderator-to-thermal neutron detector spacing at least along the major receiving directions for which the thermal neutron detector is configured. This detector characteristic—that of short scattering to detection paths—helps insure that forward and backwards scattered thermal neutrons 162 tend not to be absorbed by intervening materials or evade detection on a grazing path and thus be entirely lost to detection. Furthermore, at least one further benefit of the short path is that it takes up less linear space than a long path would require. Applicants appreciate that at least in the cases of forward and backward scattering of detected fissile neutrons, the short path between scattering and detection provides for fissile neutron detection systems that have lower extent (at least in the direction of initial neutron trajectory) as compared to conventional detectors. It is to be noted that the terms forward and backward scattering as employed herein can be considered as any scattering event where the scattered neutron deviates from its initial incoming trajectory by more than approximately 60 degrees. While the benefits of the described approaches are clearly not limited to planar embodiment, Applicants are unaware of any conventional system that uses solid neutron conversion materials that achieves as comparably high efficiency (15% detection of fissile neutrons) within thickness ranges that are so short as compared to lateral extent, without increasing by 25% or more the amount of lithium or boron that is used in the system. In other words, conventional approaches require detector systems of significantly greater thickness or neutron conversion material (such as lithium or boron) as compared to those disclosed herein. In the context of commercial applications wherein size, weight and cost are paramount, this advantage represents a significant improvement. Summarizing with respect to overall operation of various embodiment described above, a moderator arrangement, composed of a moderator material, can be surrounded by a detector arrangement such that at least 60% scattered neutrons that exit the moderator travel only a short distance before they strike an active area (for example lithium foil) of the detector arrangement. In some embodiments, the moderator defines a generally planar shape having a thickness that is short compared with any lateral extent thereof. For irregularly shaped lateral extents, the thickness can be short compared to any dimension across the lateral extents that bisects the lateral extents to form two equal areas. It is noted that in the context of this disclosure there is no requirement that these planar geometries be flat, and it should be appreciated that the planar geometries described herein can be curved in a variety of ways just as any piece of sheet metal or paper can be curved and bent in a variety of ways and yet still regarded as being generally planar. At least in the case of generally planar moderator and detector geometries, the overall collection efficiency tends to exceed that which can be obtained in conventional systems such as the TYPE I and TYPE II systems for the same amount of conversion material used in the detector. In addition to higher absolute efficiencies, the relative efficiency, and reduction of weight, cost, and/or thickness tends to exceed that of conventional detector systems such as the TYPE I and TYPE II conventional systems described above. This aspect can be readily appreciated by comparing the described planar embodiment with conventional detectors constrained to occupy and be contained within the same or similar spatial envelope as Applicant's systems. For example a detector system such as that of FIG. 1 occupies a spatial area of 1 m2 and a thickness of 0.15 meters. A conventional TYPE I detector of similar shape could readily lose between 25% and 40% of efficiency as compared to a similar sized unit that is constructed based on this disclosure. Similarly, a TYPE II system of similar area can require a thickness of 0.2 meters or more and need 200% or more lithium or boron material and would thus be at least 25% heavier and more expensive. With continued focus on generally planar embodiments, Applicants note that certain ones of the above configurations can include an outer moderator (proximate to and not surrounded by the thermal neutron detector arrangement) and an inner moderator that is almost entirely surrounded by the detector arrangement. In this context, Applicants consider a distinguishing feature of some of the embodiments disclosed herein that only incoming fissile neutrons entering from extremely shallow sideways/grazing angles (for example in some embodiments only neutrons entering sideways with less than 20 degrees from plane defined by the planar modulator) can strike the inner moderator without first passing through a thermal neutron detector. (Since thermal neutron detectors are generally not intended to have high efficiency for sideways incident thermal neutrons, these thermal neutrons can be of little consequence at least in many intended applications.) While this feature by itself does not directly result in the dislodged efficiency improvements, it is to be noted that insofar as all or most impinging high energy neutrons cannot enter the inner moderator without passing through the detector, it is conversely the case that all or most scattered low energy neutrons cannot exit and pass away from the moderator without passing through the detectors. This latter consideration clearly results sweeping advantages compared to conventional systems, including but not limited to Type I and Type II systems, and to whatever extent the former consideration results in and/or is related to these benefits, it is considered by Applicants at the very least to be of general interest. Of the many benefits of the disclosed systems, it is again of particular interest that scattered neutrons can be collected with a relatively small amount of intervening structure One implication of this unusual feature is that, in many embodiments, the designer is free to surround the moderator in very close proximity. For example, for a moderator of a given thickness, in many cases the disclosed embodiments allow for the detector to surround the moderator with gap spacings therebetween that are much smaller than the moderator thickness. This is of benefit at least for the reason that such close moderator detector spacing, over the great majority of the moderator sides, affords very little opportunity of escape for scattered neutrons. In other words the close moderator detector spacing, over almost the entirety of the moderator, prevents most scattered neutrons from escaping the detector system without impinging on some part of the detector. Stated in another way, the moderator arrangement at least substantially fills the moderator region or volume that is defined by the detector arrangement. Applicants are unaware of any conventional systems that can reasonably be regarded as sharing this important feature. Summarizing with respect to the foregoing paragraph, many of the described detector arrangements at least substantially surround their associated inner moderator with moderator-detector spacing that is small as compared to moderator thickness. FIG. 11A is a diagrammatic illustration, in an exploded perspective view, of an embodiment of a thermal neutron detector, generally indicated by the reference number 1200, that can be used alone or in combination in fissile neutron detection systems 100, 600, 700, 700′ in accordance with at least one embodiment described herein. FIG. 11B is a partially assembled view of thermal neutron detector 1200, in perspective. Thermal neutron detector 1200 uses a modular assembly 1201 in which components such as a first ground plate 1202A, a first set of electrodes 116A, active material 112 (and any support 106—see FIG. 1C) a second set of electrodes 116B, and a second ground plate 1202B can be preassembled prior to disposal in a housing 1208 that includes the top cover 104, the bottom cover 108 and at least a portion of the sidewalls 120. In some embodiments, first ground plate 1202A, first set of electrodes 116A, active material 112, second set of electrodes 116B, and second ground plate 1202B can be preassembled using a number of internal spacers to provide clearance between electrodes 116, active material 112, and ground plates 1202. The internal spacers can include a number of side spacers 1204A-1204B (collectively, “side spacers 1204”) and a number of end spacers 1206A-1206D (collectively, “end spacers 1206”) that, when assembled, provide sufficient clearance and electrical isolation of the various components within the modular assembly. As used herein, terms such as “side” and “end” denote locations relative to each other and do not represent absolute references. Thus, an “end object” can function as a “side object” when the object is rotated through an angle such as 90 degrees. Similarly, a “side object” can function as an “end object” when the object is rotated through an angle such as 90 degrees. Ground plates 1202 can include one or more electrically conductive materials. Such materials can include one or more suitable metals such as aluminum, copper, or alloys containing various quantities of aluminum or copper. In some embodiments, ground plates 1202 can include a conductive mesh material to permit the passage of readout gas 170 through all or a portion of ground plates 1202. In some implementations, ground plates 1202 can include one or more electrically conductive materials disposed on all or a portion of the exterior surface of ground plate 1202 proximate a housing 1208. Side spacers 1204 can include any number or combination of devices or components capable of maintaining a desired separation between active material 112 and ground plate 1202. Side spacers 1204 can have any suitable shape, and thus although shown as straight members in FIG. 11A, side spacers 1204 can be curved, arced, angular or any other shape needed to maintain the desired separation or distance between active material 112 and ground plate 1202. Electrodes 116 are terminated on a number of ganging boards or buses 1210A-1210D (collectively, “buses 1210”). Buses 1210 advantageously provide distribution of electric power and collection of current signals via a limited number of penetrations through the neutron detector 1200. For example, as depicted in FIG. 11A, buses 1210 permit the ganging of electrode power and beneficially route all electrical connections through one or more couplers 1212. In some instances, couplers 1212 can include a modular plug or similar device that simplifies and speeds electrical connection of the respective thermal neutron detector 1200. In embodiments, couplers 1212 can include a number of conductors for powering electrodes 116 within the fissile neutron detection system 1200. In embodiments, the one or more couplers 1212 can include a number of signal conductors for communicating thermal neutron detection signals to one or more external devices, such as a count readout device and/or alarm device. The use of one or more couplers 1212 can greatly reduce the number of penetrations through the housing 1208. Reducing the number of penetrations through the neutron detector housing reduces the likelihood of egress of the readout gas 170 from the chamber 105 and also reduces the likelihood of ingress of environmental contaminants into the chamber 105. Housing 1208 can include all or a portion of top cover 104, bottom cover 108, and at least a portion of one or more sidewalls 120. Advantageously, housing 1208 can be cast, extruded or similarly formed using a single component, thereby limiting the number of joints in thermal neutron detector 1200. Minimizing the number of joints within thermal neutron detector 1200 beneficially reduces the likelihood of egress of readout gas 170 from chamber 105 and also reduces the likelihood of ingress of environmental contaminants into chamber 105. In some implementations, end plates 120B and 120C can be attached to the housing 1208 using one or more joints having a sealant 124 disposed therein. In some implementations, the end plates 120B and 120C may be attached to the housing 1208 via welding or brazing. In other embodiments, the end plates 120B and 120C may be attached to the housing 1208 via one or more fasteners, such as one or more screws or similar. It should be appreciated that a projection 1209 (shown as a dashed line) of each of the first and second sets of electrodes onto active sheet material layer 112 defines an area that substantially covers one of the major surfaces of the active sheet material layer. That is, projection 1209 covers more than 50 percent of the major surface area of active sheet material that each electrode set faces. FIG. 11C is a detail drawing depicting, in a diagrammatic elevational cutaway view, an electrode connection device 1250 for use with thermal neutron detector 1200 depicted in FIGS. 11A and 11B, in accordance with at least one embodiment of the present disclosure. In at least some implementations, electrodes 116 can electrically conductively couple to a bus ganging structure or bus 1210. One or more conductors or pins 1256 can pass through a sealing plate 1258 that is affixed to housing 1208, for example to top cover 104 of housing 1208 and/or end plate 120C (FIG. 11B). One or more seals 1260, for example one or more polyisobutylene seals can be disposed between sealing plate 1258 and cover 104 or end plate 120C to provide hermetically sealed chamber 105. In some instances, a member or standoff 1252 and sealing plate 1258 can separate and, in some instances, electrically isolate bus 1210 from cover 104 or endplate 120C. In the embodiment depicted in FIG. 11C, a conductive bolt 1254 penetrates sealing plate 1258. A second polyisobutylene seal 1262 can be disposed between conductive bolt 1254 and sealing plate 1258. FIG. 11D is a detail drawing, in a diagrammatic elevational cutaway view, depicting another electrode connection device 1250 for use with the illustrative neutron detector 1200 depicted in FIGS. 11A and 11B, in accordance with at least one embodiment of the present disclosure. In at least some implementations, some or all of electrodes 116 can electrically conductively couple to bus ganging structure or bus 1210. One or more conductors or pins 1256 pass through sealing plate 1258 that is affixed to housing 1208, for example to top cover 104 of housing 1208 or end plate 120C (FIG. 11B). One or more seals 1260, for example one or more polyisobutylene seals 1260 can be disposed between sealing plate 1258 and cover 104 or end plate 120C to provide hermetically sealed chamber 105. In some instances, a member or standoff 1252 and sealing plate 1258 can separate and, in some instances, electrically isolate bus 1210 from cover 104 and/or endplate 120C. In the embodiment depicted in FIG. 11D, a sealing member 1270, such as an epoxy sealing member, penetrates sealing plate 1258 and at least partially surrounds pin 1256. FIG. 11E is a detail drawing depicting, in a diagrammatic elevational cutaway view, another electrode connection device 1250 for use with thermal neutron detector 1200 depicted in FIGS. 11A and 11B, in accordance with at least one embodiment of the present disclosure. In at least some implementations, some or all of electrodes 116 can electrically conductively couple to bus ganging structure or bus 1210. One or more conductors or pins 1256 pass through an aperture in housing 1208, for example in top cover 104 and/or endplate 120C of housing 1208. One or more weld or O-ring seals can be disposed to seal at least a portion of the aperture in housing 1208. In some instances, a plug 1286 such as a metal or polymeric plug can be disposed proximate the aperture in housing 1208. A glass or epoxy seal 1284 can at least partially surround pins 1256 extending from bus 1210. A member or standoff 1252 and glass or epoxy seal 1284 can separate and, in some instances, electrically isolate bus 1210 and pin 1256 from cover 104. In the embodiment depicted in FIG. 11E, the combination of weld or O-ring seal 1280, metal or polymeric plug 1286, and glass or epoxy seal 1284 at least partially surrounds pin 1256 and provides a hermetic seal for chamber 105. FIG. 11F is a close up diagrammatic perspective view of electrode connection device 1250 depicted in FIG. 11C, in accordance with at least one embodiment of the present disclosure. FIG. 11G is a close up diagrammatic plan view of electrode connection device 1250 depicted in FIG. 11C, in accordance with at least one embodiment of the present disclosure. FIGS. 11H-11K are diagrammatic views, in elevation, showing various thermal neutron detector (TND) embodiments that Applicants employed for purposes of simulations to determine the efficiency of each TND embodiment. FIG. 11H illustrates what is referred to as a Double Outward TND-Double Foil embodiment, generally indicated by the reference number 1286, and including a detection chamber 105 defined between first and second active sheet material layers ASM1 and ASM2, respectively, such as 6Li with one set of electrodes 116 in chamber 105. In the present figures, it is to be understood that chambers 105 sealingly contain readout gas in a manner that is consistent with the descriptions above. FIG. 11I illustrates what is referred to as a Double Inward TND-Single Foil embodiment, generally indicated by the reference number 1288, and including detection chamber 105 with a single active sheet material layer ASM1 centered in the detection chamber between first and second sets of electrodes 116A and 116B. FIG. 11J illustrates what is referred to as a Triple Inward TND-Double Foil embodiment, generally indicated by the reference number 1290, and including detection chamber 105 with active sheet material layers ASM1 and ASM2 supported in the detection chamber between first, second and third sets of electrodes 116A-116C. FIG. 11K illustrates what is referred to as a Quadruple Inward TND-Triple Foil embodiment, generally indicated by the reference number 1292, and including detection chamber 105 with active sheet material layers ASM1-ASM3 supported in the detection chamber between four sets of electrodes 116A-116D, as shown. FIG. 11L is a plot of thermal neutron detection efficiency per lithium foil 1296 versus Lithium foil thickness 1298 in microns wherein the legend in the figure illustrates the simulation efficiency for a corresponding embodiment 1286, 1288, 1290 and 1292 of FIGS. 11H-11K, respectively. In this regard, the horizontal plot of lithium foil thickness shows the thickness of one layer. In this regard, the horizontal plot of lithium foil thickness shows the thickness of one layer. It is clear from these plots that the double inward configuration of embodiment 1288 of FIG. 11I presents the greatest efficiency per foil layer. Given that the double inward embodiment requires only a single layer of foil, Applicants recognize that adding foil layers produces diminished returns at increased cost. A person of ordinary skill in the art, having this disclosure in hand, should readily appreciate that one way to increase thermal neutron detector efficiency is to simply add detection (i.e., active) layers, for example by stacking thermal neutron detectors on top of one another along a given receiving direction. However, Applicants appreciate that such stacking is not necessarily appropriate especially in cases where the designer seeks to optimize performance with respect to one or more of cost, size and weight of an overall detector system. In other words, Applicants respectfully submit that diminishing benefits are likely just through the practice of blindly stacking thermal neutron detectors. It should be appreciated by a person of ordinary skill in the art, having this disclosure in hand, that an aspect of these descriptions is to teach techniques for enhancing efficiency of devices that utilize as few layers as possible given a particular set of performance goals. Indeed, one aspect of this disclosure is to teach techniques for maximizing the efficiency per unit cost of a given detector by utilizing only two layers of lithium foil, each one in a TND, on opposite sides of one moderator layer. FIG. 12A is a diagrammatic exploded view, in perspective of an embodiment of a fissile neutron detection system, generally indicated by the reference number 1300, that uses three neutron moderators 150A-150C and eight thermal neutron detectors 1200 such as depicted, for example, in FIGS. 11A and 11B, in accordance with at least one embodiment of the present disclosure. FIG. 12B is a diagrammatic, assembled view, in perspective, of fissile neutron detection system 1300 of FIG. 12A. The modular construction of the thermal neutron detector 1200 beneficially permits positioning of any number of thermal neutron detectors 1200 in a variety of configurations. Such can, for example, facilitate the use of a single thermal neutron detector 1200 within a relatively compact, portable, handheld device and the combination of a number of thermal neutron detectors 1200 into a stationary roadside monitoring array. As depicted in FIGS. 12A and 12B, neutron moderator 150A is disposed between eight thermal neutron detectors 1200A-1200H arranged in two rows of four detectors. The majority of the surface area of neutron moderator 150A is therefore disposed proximate one or more thermal neutron detectors 1200. External neutron moderators 150B and 150C can be disposed proximate the surfaces of the thermal neutron detectors 1200A-1200H that are opposite neutron moderator 150A. It is noted that the external moderators are not a requirement. As depicted in FIGS. 12A and 12B thermal neutron detection systems 1200 can be arranged such that one or more couplers 1212A-1212H for each of thermal neutron detectors 1200A-1200H, respectively, exits the assembly from a single end. Such an arrangement can facilitate the connection of each of thermal neutron detector 1200 to a communications bus, a power distribution bus, or any combination thereof. In embodiments, chamber 105 of thermal neutron detector 1200 can have a length, measured along a first axis, of approximately 100 centimeters (cm). In embodiments, chamber 105 can have a thickness, measured along a second axis orthogonal to the first axis, of approximately 3.5 centimeters (cm). In embodiments, chamber 105 can have a width, measured along a third axis orthogonal to the first axis and the second axis, of approximately 20 centimeters (cm). Although not depicted in FIGS. 12A and 12B, fissile neutron detection system 1300 can be disposed partially or completely within an external housing. Such can facilitate the installation of system in an outdoor environment such as a checkpoint, port-of-entry, or similar locations where screening for fissile nuclear material can be beneficial. Attention is now directed to FIGS. 13A-13C, each of which diagrammatically illustrates the lateral extents of a moderator arrangement as part of a fissile neutron detection system in a plan view. It is noted that the moderator arrangement in each of these figures can be made up of one or more individual parts or pieces of moderating material such as, for example, HDPE. In a manner that is consistent with the descriptions above, interstitial gaps can be present within the moderator arrangement without significant adverse effect on its moderating properties, although such gaps are not required. Any suitable shape can be used with the shapes used in the present figures serving as examples that are not intended as limiting. FIG. 13A illustrates an embodiment of lateral extents generally indicated by the reference number 1320. The lateral extents of any moderator arrangement disclosed herein can form opposing major sides that can be planar although curved and other suitable shapes can be used. Lateral extents 1320 form a rectangular peripheral outline or edge configuration in the present figure and can be bisected or divided into two areas of equal size, for example, by an arbitrary dimension 1322 to form equal areas A1 and A2. An unlimited number of such bisecting dimensions can be defined. In moderator embodiments described within the scope of the present application, any given dimension that bisects the lateral extents is greater than any thickness of the moderator arrangement normal to the plane of the figure. Lateral extents 1320, of the present example, further define a widthwise major dimension 1324 and a lengthwise major dimension 1326. In this instance, the lengthwise and widthwise major dimensions are each greater than any thickness of the moderator arrangement, FIG. 13B illustrates another embodiment of lateral extents, generally indicated by the reference number 1330, forming an elliptical peripheral outline that is shown as bisected by an arbitrary given dimension 1332 by way of example. Further, a major widthwise dimension 1334 as well as a major lengthwise dimension 1336 are shown, each of which is greater than any thickness of the moderator arrangement in a direction normal to the plane of the figure. While is recognized that multiple embodiments described herein employ a rectangular peripheral outline or edge configuration, the disclosure is not to be construed as being limited in this regard. On one hand, Applicants appreciate that rectangular outlines can often facilitate ease of manufacturing based on a variety of practical considerations. For example, detectors with rectangular peripheral outlines can often be readily assembled using combinations of rectilinear components and/or subsystems. Structural members composed of metal or other materials are often widely and inexpensively available as bar stock, which, in many cases, can lend to the ease of construction of rectangular geometries. Oftentimes rectangular plates can be produced with relative ease and minimal waste as compared with more complex shapes. Applicants further appreciate that rectangular shapes can facilitate design and construction of robust and easily assembled thermal neutron detectors for a variety of practical reasons that should be apparent to a person of ordinary skill in the art having this disclosure in hand. For example, techniques and/or assemblies intended to facilitate high precision positioning of complex structures including but not limited to electrode wires and lithium foil, can in some cases be more straightforward to implement as compared to systems with complex non-rectangular peripheral outlines. Applicants recognize, however, that in some applications, end-use specifications may include requirements for more complex shapes. For example, a fissile neutron detector system could be specified in such a way that it is required to cover a large fraction of surface area of some external encasement that itself defines an irregular non-rectangular shape. In such instances, the teachings and descriptions herein can be readily applied by one having overall skill in the art in such a way as to capture the sweeping benefits brought to bear thereby. FIG. 13C illustrates still another embodiment of lateral extents, generally indicated by the reference number 1340, having an irregular peripheral outline or edge configuration. In particular, a major rectangular portion 1342 can be the same dimensionally as the rectangular shape of lateral extents 1320 in FIG. 13A, however, a thin rectangular shape 1344 has been appended to rectangular portion 1342 to form the overall irregular lateral extents. It is noted that thin rectangle 1344 can have a width 1346 that is less than any thickness of the moderator arrangement. Nevertheless, any given dimension that bisects lateral extents 1340 such as, for example, dimension 1322′ includes a length that is greater than any thickness of the moderator arrangement. In this case, area A2′ includes a contribution from thin rectangular region such that A2′ is greater than A2 of FIG. 13C causing dimension 1322′ to shift by an amount that causes A1′ to increase in area by an amount that is equal to the area of thin rectangle 1344 such that A1′ is equal to A2′. Thus, irregular shapes such as shown in FIG. 13C are submitted to be within the scope of the present disclosure. With this disclosure in hand, one of ordinary skill in the art will appreciate that virtually an unlimited number of irregular shapes fall within the purview of FIG. 13C and the teachings that have been brought to light herein. Reference is now made to the graph of FIG. 14, generally indicated by the reference number 1400 and which is taken from the prior art, which illustrates distance traveled by a neutron (Rrms in cm) in a hydrogenous (i.e., moderating) material versus incident energy of the neutron (E1 in MeV). Four plots 1402, 1404 and 1406 correspond to resultant energies Ef of 1/40 eV, 40 eV and 4 keV, respectively. Plot 1402 clearly represents a thermal neutron with a resultant energy of 0.025 eV. Applicants recognize in view of FIG. 14 that fissile neutrons around 1 MeV will thermalized after traveling somewhat less than 10 cm in the moderating material, and thus an effective thickness for a layer of moderating material is less than 10 cm. In this regard, a moderating layer that is too thin will fail to moderate incident fissile neutrons to detectable thermal neutrons. Attention is now directed to the graph of FIG. 15, generally indicated by the reference number 1500, taken from the prior art, illustrating normalized neutron flux ϕ/ϕinc on the vertical axis against the thickness of a moderating slab that is formed from a hydrogen rich material. This particular study used ordinary concrete, which is approximately 10 percent atoms per cm3 of hydrogen, but the result will be similar with other hydrogen rich materials such as water or HDPE, which are approximately 75 percent atoms per cm3 of hydrogen. Applicants recognize that, at least to a reasonable approximation, neutron loss increases rapidly as slab thickness increases. In this regard, at a thickness of 10 cm more than 70% of incident 1 MeV fissile neutrons are lost based on plot 1502. For a more hydrogen rich material such as water or HDPE, at a thickness of 10 cm more than 90% of incident 1 MeV fissile neutrons are lost. Applicants recognize that this level of neutron loss is problematic such that a moderator of 10 cm is too thick. In light of the two results shown in FIG. 14 and FIG. 15, a moderating layer 150 that is too thin will fail to moderate incident fissile neutrons to detectable thermal neutrons while a moderating layer that is too thick will result in the loss of an excessive number of incident neutrons in the moderating material such that these lost neutrons are likewise undetectable. In order to maintain, neutron loss in a range that is acceptable such as, for example, 20 percent to 40 percent, moderator thickness can be in the range of at least approximately 1 cm to 5 cm. With this disclosure in hand and the familiarity of one of ordinary skill in the art with simulation techniques that are applicable to fissile neutron detection such as, for example, Monte Carlo simulations, it is submitted that one of ordinary skill in the art can identify an optimal moderator thickness within the subject thickness range. The following examples pertain to embodiments that employ some or all of the described fissile neutron detection apparatuses, systems, and methods described herein. The enclosed examples should not be considered exhaustive, nor should the enclosed examples be construed to exclude other combinations of the systems, methods, and apparatuses disclosed herein and which are not specifically enumerated herein. a. According to Example 1 there is provided a fissile neutron detection system. The fissile neutron detection system can include at least one neutron detector. Each neutron detector can further include a body having a length, a width, and an extent defining a closed chamber, the length and the width of the chamber greater than the thickness of the chamber. Each neutron detector can further include at least one active material that emits at least one ionizing particle upon exposure to thermal neutrons, the active material disposed within the chamber; and at least one electrode. The fissile neutron detection system also includes at least one neutron moderator disposed proximate the at least one thermal neutron detector, the at least one neutron moderator including a material that transitions at least a portion of high-energy incident fissile neutrons to low-energy thermal neutrons, wherein at least 60% of the low-energy thermal neutrons exiting the moderator enter the thermal neutron detector(s). Example 2 can include elements of example 1 where the chamber formed by the body of each thermal neutron detector can include a single, continuous, chamber. Example 3 can include elements of example 2 where the chamber formed by the body of each thermal neutron detector can include a hermetically sealed chamber. Example 4 can include elements of example 1 where the at least one thermal neutron detector can include a plurality of thermal neutron detectors. Example 5 can include elements of example 1 where the at least one neutron moderator can include a plurality of neutron moderators. Example 6 can include elements of example 1 where the at least one neutron moderator can include a material that includes a minimum of 40 weight percent hydrogen. Example 7 can include elements of example 6 where the at least one neutron moderator can include high-density polyethylene (HDPE) member. Example 8 can include elements of example 7 where the at least one neutron moderator can include a HDPE member having a uniform thickness of from approximately 1 centimeter (cm) to approximately 5 cm. Example 9 can include elements of example 1 where the at least one neutron moderator comprises a number of members, each of the members having a uniform thickness. Example 10 can include elements of example 1 and can additionally include a voltage source conductively coupled to the at least one electrode in the at least one thermal neutron detector. Example 11 can include elements of example 1 and can additionally include a number of support members disposed at intervals along at least a portion of a length of the at least one electrode. Example 12 can include elements of example 1 where the at least one thermal neutron detector can include an exterior surface having a top cover and a bottom cover separated by a sidewall having an extent or height. Example 13 can include elements of example 12 where the sidewall comprises a multi-piece sidewall. Example 14 can include elements of example 12 where the at least one thermal neutron detector can include a first thermal neutron detector and a second thermal neutron detector; where the at least one neutron moderator can be disposed proximate at least a portion of the exterior surface of the first thermal neutron detector and at least a portion of the exterior surface of the second thermal neutron detector; and where at least a portion of the at least one neutron moderator can be disposed in a space bordered by the portion of the exterior surface of the first thermal neutron detector and the portion of the exterior surface of the second thermal neutron detector. Example 15 can include elements of example 14 where the first thermal neutron detector can include a planar body having a planar top cover and a planar bottom cover; where the second thermal neutron detector can include a planar body having a planar top cover and a planar bottom cover; and where the neutron moderator can include a planar member disposed proximate the top cover of the first thermal neutron detector and the top cover of the second thermal neutron detector. Example 16 can include elements of example 15 where the planar top cover of the first thermal neutron detector can have a length of approximately 100 centimeters (cm) and a width of from approximately 20 cm; where the planar bottom cover of the first thermal neutron detector can have length of from approximately 100 and a width of approximately 20 cm; where the sidewall of the first thermal neutron detector can have a thickness of from approximately 0.5 cm to approximately 5 cm; where the planar top cover of the second thermal neutron detector can have a length of approximately 100 centimeters (cm) and a width of approximately 20 cm; where the planar bottom cover of the second thermal neutron detector can have length of from approximately 100 cm and a width of from approximately 20 cm; and where the sidewall of the second thermal neutron detector can have a thickness of approximately 3.5 cm. Example 17 can include elements of example 14 where the first thermal neutron detector can include an arcuate body having an arcuate top cover and an arcuate bottom cover; where the second thermal neutron detector can include an arcuate body having an arcuate top cover and an arcuate bottom cover; and where the neutron moderator can include a constant thickness planar member disposed proximate the top cover of the first thermal neutron detector and the top cover of the second thermal neutron detector. Example 18 can include elements of example 14 where the first thermal neutron detector can include an angular body having an angular top cover and an angular bottom cover; where the second thermal neutron detector can include an angular body having an angular top cover and an angular bottom cover; and where the neutron moderator can include a constant thickness planar member disposed proximate the top cover of the first thermal neutron detector and the top cover of the second thermal neutron detector. Example 19 can include elements of example 1 where the at least one neutron moderator can include at least one member having an exterior surface; and where the at least one thermal neutron detector can be disposed proximate at least a portion of the exterior surface of the member of the at least one neutron moderator. Example 20 can include elements of example 1 where the at least one thermal neutron detector body can include a body having an exterior surface; and where the at least one neutron moderator can include at least one external neutron moderator disposed proximate at least a portion of the exterior surface of the body of the at least one thermal neutron detector. Example 21 can include elements of example 3 where the at least one thermal neutron detector can include an ionizable readout gas disposed within the hermetically sealed chamber. Example 22 can include elements of example 21 where the ionizable readout gas can include at least one noble gas. Example 23 can include elements of example 22 where the at least one noble gas can include argon (Ar). Example 24 can include elements of any of examples 1 through 23 where the at least one active material can include at least one sheet of solid active material. Example 25 can include elements of example 24 where the at least one sheet of active material can include at least one lithium 6 (6Li) sheet. Example 26 can include elements of example 25 where each 6Li sheet can include a 6Li sheet having a thickness of from approximately 50 micrometers (μm) to approximately 120 μm. Example 27 can include elements of example 26 where each 6Li sheet can include a 6Li sheet having a length and a width that exceed the thickness of the 6Li sheet. Example 28 can include elements of example 27 and can additionally include a support structure disposed proximate each 6Li sheet, the support structure disposed at an intermediate location within the chamber. Example 29 can include elements of example 27 where the at least one 6Li sheet can be disposed proximate at least a portion of at least one surface forming an interior of the chamber. Example 30 can include elements of any of examples 1 through 23 where the at least one active material can include at least one layer of active material. Example 31 can include elements of example 30 where the at least one layer of active material can include at least one layer containing boron 10 (10B). Example 32 can include elements of example 31 where the at least one layer containing 10B can include at least one layer of 10B disposed on at least a portion of at least one interior surface of the chamber in the respective neutron detector. Example 33 can include elements of any of examples 1 through 20 where the at least one active material comprises an active gas disposed within the chamber. Example 34 can include elements of example 33 where the active gas disposed within the chamber can include at least one gas containing helium 3 (3He). According to example 35, there is provided a fissile neutron detection method. The fissile neutron detection method can include transitioning at least some incident fissile neutrons to thermal neutrons by passing the incident fissile neutrons through at least one neutron moderator disposed proximate at least one thermal neutron detector. The at least one thermal neutron detector can include: a body having a length, a width, and a thickness defining a closed chamber; the length and the width of the chamber greater than the thickness of the chamber; at least one active material that emits at least one ionizing particle upon exposure to thermal neutrons, the active material disposed within the chamber; and at least one electrode. The method can also include impinging at least 60% of the thermal neutrons exiting the neutron moderator on the at least one active material disposed in the chamber of the at least one thermal neutron detector. The method can further include generating, by the at least one thermal neutron detector, a current at the at least one electrode, the current proportional to the number of thermal neutrons impinging on the at least one active material in the at least one thermal neutron detector. Example 36 can include elements of example 35 where transitioning at least some incident fissile neutrons to thermal neutrons by passing the incident fissile neutrons through at least one neutron moderator disposed proximate at least one thermal neutron detector can include transitioning at least some incident fissile neutrons to thermal neutrons by passing the incident fissile neutrons through at least one neutron moderator proximate a plurality of thermal neutron detectors disposed proximate at least a portion on an exterior side of the at least one neutron moderator. Example 37 can include elements of example 35 where transitioning at least some incident fissile neutrons to thermal neutrons by passing the incident fissile neutrons through at least one neutron moderator disposed proximate at least one thermal neutron detector can include transitioning at least some incident fissile neutrons to thermal neutrons by passing the incident fissile neutrons through at least one of a plurality of thermal neutron moderators disposed proximate at least a portion on an exterior surface of the at least one thermal neutron detector. Example 38 can include elements of example 35 where impinging at least 60% of the thermal neutrons exiting the neutron moderator on the at least one active material disposed in the chamber of the at least one thermal neutron detector can include impinging at least 60% of the thermal neutrons exiting the at least one neutron moderator on at least one active material disposed in a hermetically sealed chamber of the at least one thermal neutron detector. Example 39 can include elements of example 35 and can additionally include generating, at least one signal proportional to at least one of: the ionization created by the interaction of the neutron and active material in the at least one thermal neutron detector, or the number of thermal neutrons impinging on the at least one active material in the at least one thermal neutron detector or the rate of thermal neutron impingements on the at least one active material in the at least one thermal neutron detector. Example 40 can include elements of example 35 where transitioning at least some incident fissile neutrons to thermal neutrons by passing the incident fissile neutrons through at least one neutron moderator disposed proximate at least one thermal neutron detector can include transitioning at least some incident fissile neutrons to thermal neutrons by passing the incident fissile neutrons through at least one neutron moderator that includes a minimum of 40 weight percent hydrogen. Example 41 can include elements of example 40 where transitioning at least some incident fissile neutrons to thermal neutrons by passing the incident fissile neutrons through at least one neutron moderator that includes a minimum of 10 percent atoms per cm3 of hydrogen can include transitioning at least some incident fissile neutrons to thermal neutrons by passing the incident fissile neutrons through at least one neutron moderator that includes a material containing a high density polyethylene (HDPE). Example 42 can include elements of example 41 where transitioning at least some incident fissile neutrons to thermal neutrons by passing the incident fissile neutrons through at least one neutron moderator that includes a material containing a high density polyethylene (HDPE) can include transitioning at least some incident fissile neutrons to thermal neutrons by passing the incident fissile neutrons through at least one neutron moderator that includes HDPE having a thickness of from approximately 1 centimeter (cm) to approximately 5 cm. Example 43 can include elements of example 35 and can additionally include at least partially encapsulating at least a portion of the at least one thermal neutron detector and at least a portion of the at least one neutron moderator in an external neutron moderator. Example 44 can include elements of example 43 where at least partially encapsulating at least a portion of the at least one thermal neutron detector and at least a portion of the at least one neutron moderator in an external neutron moderator can include at least partially encapsulating at least a portion of the at least one thermal neutron detector and at least a portion of the at least one neutron moderator in a material that includes a minimum of 10 percent atoms per cm3 of hydrogen. Example 45 can include elements of example 44 where at least partially encapsulating at least a portion of the at least one thermal neutron detector and at least a portion of the at least one neutron moderator in a material that includes a minimum of 10 percent atoms per cm3 of hydrogen can include at least partially encapsulating at least a portion of the at least one thermal neutron detector and at least a portion of the at least one neutron moderator in a material that includes high-density polyethylene (HDPE). Example 46 can include elements of any of examples 35 through 45, where generating, by the at least one thermal neutron detector, a current at the at least one electrode, the current correlated to the number of thermal neutrons impinging on the at least one active material in the at least one thermal neutron detector can include, for each thermal neutron impinging on the at least one active material in the at least one thermal neutron detector, generating at least one ionizing particle by at least one sheet of active material; ionizing, by the at least one ionizing particle, a readout gas disposed within the chamber of the at least one thermal neutron detector; maintaining the at least one electrode disposed in the chamber of the at least one thermal neutron detector at a voltage that differs from a voltage of the at least one sheet of active material; causing, by the ionized readout gas, a flow of charged particles away from the at least one electrode; and causing a current at the electrode by the flow of charged particles, the current correlated to the number of thermal neutrons impinging on the at least one sheet of active material disposed in the chamber of the at least one thermal neutron detector. Example 47 can include elements of example 46 where ionizing a readout gas disposed within the chamber of the at least one thermal neutron detector can include ionizing, by the at least one ionizing particle, a readout gas disposed within the chamber of the at least one thermal neutron detector. Example 48 can include elements of example 46 where maintaining the at least one electrode disposed in the chamber of the at least one neutron detector at a voltage that differs from a voltage of the at least one sheet of active material can include biasing the at least one electrode to a potential of at least +1100 volts (V) measured with respect to the potential of the at least one sheet of active material. Example 49 can include elements of example 46 where generating at least one ionizing particle by at least one sheet of active material can include generating the at least one ionizing particle by at least one solid sheet of active material disposed within the chamber of the at least one thermal neutron detector. Example 50 can include elements of example 49 where generating the at least one ionizing particle by at least one solid sheet of active material disposed within the chamber of the at least one neutron detector can include generating at least one ionizing particle by at least one solid sheet of active material comprising at least one lithium 6 (6Li) sheet disposed within the chamber of the at least one neutron detector. Example 51 can include elements of example 50 where generating at least one ionizing particle by at least one solid sheet of active material comprising at least one lithium 6 (6Li) sheet disposed within the chamber of the at least one thermal neutron detector can include generating at least one ionizing particle by at least one solid sheet of active material comprising at least one 6Li sheet having a thickness of from approximately 50 micrometers to approximately 120 micrometers disposed within the chamber of the at least one thermal neutron detector. Example 52 can include elements of example 51 where generating at least one ionizing particle by at least one solid sheet of active material comprising at least one 6Li sheet having a thickness of from approximately 50 micrometers to approximately 120 micrometers disposed within the chamber of the at least one thermal neutron detector can include generating at least one ionizing particle by at least one solid sheet of active material comprising at least one 6Li sheet disposed within the chamber of the at least one thermal neutron detector, the at least one sheet of 6Li comprising at least one of: a single 6Li sheet proximate a support structure and positioned at an intermediate point within the chamber of the at least one thermal neutron detector; at least one 6Li sheet disposed proximate at least a portion of at least one wall forming at least a portion of the chamber of the at least one thermal neutron detector; or a number of spaced 6Li sheets proximate a support structure and positioned at an intermediate point within the chamber of the at least one thermal neutron detector. Example 53 can include elements of any of examples 35 through 45 where generating, by the at least one thermal neutron detector, a current at the at least one electrode, the current correlated to the number of thermal neutrons impinging on the at least one active material in the at least one thermal neutron detector can include, for each thermal neutron impinging on the at least one active material in the at least one thermal neutron detector, generating at least one ionizing particle by at least one layer of active material disposed within the chamber of the at least one thermal neutron detector; ionizing, by the at least one ionizing particle, a readout gas disposed within the chamber of the at least one thermal neutron detector; maintaining the at least one electrode disposed in the chamber of the at least one thermal neutron detector at a voltage that differs from a voltage of the at least one layer of active material; causing, by the ionized readout gas, a flow of charged particles away from the at least one electrode; and causing a current at the electrode by the flow of charged particles, the current correlated to the number of thermal neutrons impinging on the at least one layer of active material disposed in the chamber of the at least one thermal neutron detector. Example 54 can include elements of example 53 where generating at least one ionizing particle by at least one layer of active material disposed within the chamber of the at least one thermal neutron detector can include generating at least one ionizing particle by at least one layer of active material comprising at least one layer containing 10B disposed within the chamber of the at least one thermal neutron detector. Example 55 can include elements of example 54 where generating at least one ionizing particle by at least one layer of active material comprising at least one layer containing 10B disposed within the chamber of the at least one thermal neutron detector can include generating at least one ionizing particle by at least one layer of active material comprising at least one layer containing 10B disposed proximate at least a portion of at least one wall forming at least a portion of the chamber of the at least one thermal neutron detector. Example 56 can include elements of example 53 where ionizing a readout gas disposed within the chamber of the at least one thermal neutron detector can include ionizing, by the at least one ionizing particle, a noble readout gas disposed within the chamber of the at least one thermal neutron detector. Example 57 can include elements of examples 35 through 45 where generating, by the at least one neutron detector, a current at the at least one electrode, the current correlated to the number of thermal neutrons impinging on the at least one active material in the at least one thermal neutron detector can include, for each thermal neutron impinging on the at least one active material in the at least one thermal neutron detector, generating at least one ionizing particle by at least one active gas disposed within the chamber of the at least one thermal neutron detector; maintaining the at least one electrode disposed in the chamber of the at least one thermal neutron detector at a potential greater than the at least one active gas; causing, by the ionized readout gas, a flow of charged particles away from the at least one electrode; and causing a current at the electrode by the flow of charged particles, the current correlated to the number of thermal neutrons impinging on the at least one layer of active material disposed in the chamber of the at least one thermal neutron detector. Example 58 can include elements of example 57 where generating at least one ionizing particle by at least one active gas disposed within the chamber of the at least one thermal neutron detector can include generating the at least one ionizing particle by at least one active gas that includes helium 3 (3He), the at least one active gas disposed within the chamber of the at least one thermal neutron detector. The terms and expressions which have been employed herein are used as terms of description and not of limitation, and there is no intention, in the use of such terms and expressions, of excluding any equivalents of the features shown and described (or portions thereof), and it is recognized that various modifications are possible within the scope of the claims. Accordingly, the claims are intended to cover all such equivalents. |
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