Split (2)

train · 33.5k rows

patent_number stringlengths 0 9	section stringclasses 4 values	raw_text stringlengths 0 954k
	claims	1. A lower tie plate, comprising:an inlet nozzle; anda flow-directing structure that extends over the inlet nozzle, the flow-directing structure comprising:a closed surface that defines a damper that is configured to obstruct flow through the inlet nozzle; andan opening in the flow-directing structure that is configured to allow flow through the flow directing structure and into the inlet nozzle, wherein the opening is asymmetrical with respect to a longitudinal axis of the lower tie plate;wherein the lower tie plate is configured to be:positioned at a lower end of a fuel assembly; andat least partially received in a seating orifice of a fuel support member;wherein each of the closed surface and the opening are configured to be positioned in the seating orifice of the fuel support member when the lower tie plate is at least partially received in the seating orifice of the fuel support member; andwherein each of the closed surface and the opening has a rotational position in the seating orifice of the fuel support member, wherein the rotational position is based on rotation of the flow directing structure:around the longitudinal axis; andrelative to the seating orifice of the fuel support member. 2. The lower tie plate of claim 1, wherein the flow-directing structure is configured to rotate with rotation of the fuel assembly such that rotation of the fuel assembly is rotation of the flow-directing structure about the longitudinal axis. 3. The lower tie plate of claim 1, the flow-directing structure further comprising a bail. 4. The lower tie plate of claim 1, wherein the closed surface and the opening are each offset with respect to a longitudinal axis of the lower tie plate. 5. The lower tie plate of claim 4, wherein the closed surface is configured to divert flow from a linear flow path through the lower tie plate. 6. The lower tie plate of claim 1, wherein the closed surface is configured to obstruct flow through the inlet nozzle and into the fuel assembly. 7. The lower tie plate of claim 1, wherein the flow-directing structure is configured to be rotated independently of a body of the lower tie plate. 8. The lower tie plate of claim 1, wherein the flow-directing structure is configured to be received within the seating orifice of the fuel support member to couple the inlet nozzle to a channel of the fuel support member. 9. The lower tie plate of claim 1, wherein the inlet nozzle includes an opening to an enlarged volume within the lower tie plate. 10. The lower tie plate of claim 9, comprising a rod supporting grid that is located at an upper end of the enlarged volume, wherein the rod supporting grid is configured to house ends of fuel rods and to direct a flow of coolant from the enlarged volume into the fuel assembly and between the fuel rods. 11. The lower tie plate of claim 1, wherein the flow-directing structure includes a bail that is configured to facilitate directing a lower end of the lower tie plate so as to be received by the seating orifice of the fuel support member. 12. The lower tie plate of claim 1, wherein the flow-directing structure has a shape with a concave surface. 13. The lower tie plate of claim 1, wherein a ratio of a size of the closed surface to a size of the opening is at least one to two.
050193287	claims	1. A natural circulation type boiling light-water reactor comprising: a pressure vessel divided into a steam/water chamber and a steam chamber; a reactor core disposed within said steam/water chamber to generate main steam which contains radioactive isotope .sup.16 N, said reactor core including a plurality of fuel elements; a shroud disposed within said steam/water chamber encircling said reactor core; a steam dryer assembly through which the main steam generated from said shroud passes into said steam chamber to reduce a wetness fraction of said main steam; a chimney connected at one end thereof to said shroud and extending within said steam/water chamber toward said steam chamber, through which the main steam flows together with said radioactive isotope .sup.16 N, said chimney being filled with light water as coolant and having the other end thereof opened toward said steam dryer assembly; a steam outlet through which the main steam generated is drawn out of said pressure vessel, said steam outlet being provided in a wall of said pressure vessel; and steam passage means through which the main steam generated passes from said shroud to said steam outlet via said steam dryer assembly. naturally circulating light water within a pressure vessel through a reactor core so as to generate main steam containing radioactive isotope .sup.16 N; making said main steam generated travel in said pressure vessel with taking a time period twice or more longer than a half-life of said radioactive isotope .sup.16 N; and drawing said main steam generated out of said pressure vessel. 2. A reactor according to claim 1, wherein it takes said main steam a time period twice or more longer than a half-life of said radioactive isotope .sup.16 N to flow from said shroud to said steam outlet. 3. A reactor according to claim 2, wherein said reactor further comprises means for prolonging said steam passage means. 4. A reactor according to claim 3, wherein said prolonging means includes a steam guide connected to one end of said steam dryer assembly and extending within said steam chamber, and wherein said steam outlet is provided in a portion of said wall of said pressure vessel located below an outlet of said steam guide. 5. A reactor according to claim 3, wherein said prolonging means is constituted by increasing axial dimension of said chimney. 6. A reactor according to claim 4, wherein said prolonging means is also constituted by increasing axial dimension of said chimney. 7. A reactor according to claim 2, wherein said reactor further comprises means for reducing a travelling speed of said main steam in said steam passage means. 8. A reactor according to claim 7, wherein said travelling speed reducing means is constituted by increasing radial inside dimension of said chimney beyond the extend of said shroud. 9. A process for drawing main steam containing radioactive isotope .sup.16 N out of a natural circulation type boiling light-water reactor, comprising the steps of:
055219519	claims	1. A method for repairing a reactor core shroud having a circumferential crack at a predetermined elevation, comprising the steps of: forming a curved bracket with first and second circular holes; machining a first circular hole in a first shroud section located at an elevation higher than said predetermined elevation, said first circular hole being located to align with said first circular hole formed in said bracket when said bracket is placed in a predetermined position relative to said shroud; machining a second circular hole in a second shroud section located at an elevation lower than said predetermined elevation, said second circular hole being located to align with said second circular hole formed in said bracket when said bracket is in said predetermined position; placing said bracket in said predetermined position outside said shroud with said respective first and second circular holes in alignment; blindly installing first and second pin assemblies into said respective first and second circular holes of said bracket and said shroud by remote manipulation. inserting first and second pin assemblies into said respective first and second circular holes of said bracket and thereafter into said respective first and second circular holes of said shroud, leaving a portion of said first and second pin assemblies protruding outside said shroud; and remotely manipulating each of said portions of said first and second pin assemblies protruding outside said shroud to securely fasten said bracket to said shroud. 2. The method as defined in claim 1, wherein said blind installation comprises the steps of: 3. The method as defined in claim 2, wherein each of said remotely manipulated first and second pin assemblies is tensioned to exert a radially outwardly directed contact load on the cylindrical surface of said first and second circular holes respectively formed in said shroud.
054815789	abstract	A fuel bundle and lower tie plate assembly for a nuclear reactor includes a plurality of fuel rods supported between an upper tie plate and a lower tie plate assembly, the lower tie plate assembly including an upper grid portion and a lower body portion, the upper grid portion having a plurality of fuel rod supporting bosses interconnected by a plurality of webs thus forming flow openings between the bosses. The body portion includes an inlet nozzle and a peripheral wall extending between the bottom nozzle and the upper grid portion to define a flow volume therein. A debris catcher, including a plurality of perforated tubes, is incorporated into the lower tie plate assembly, such that one of the tubes is in abutment with a respective lowermost end of each of the plurality of fuel rod supporting bosses. The tubes are attached to a plate formed with openings corresponding to the open lower ends of the tubes so that coolant is forced to flow through the tubes in order to pass through the grid portion of the lower tie plate. The openings in the tubes cause the coolant to change direction, thus trapping debris in the tubes and effectively preventing debris from passing through the lower tie plate assembly.
047602660	summary	The invention relates to a device for the generation of thermal neutrons, including a housing containing a transportable neutron source and a moderator and having a neutron outlet opening. German Patent No. DE-PS 30 31 107 describes such a device. Normally, cost-intensive basic materials are needed for the production of neutron sources. Therefore, the need to improve the ratio of the source size to the number of thermal neutrons exiting the source housing has existed for a long time. However, the solutions to this problem suggested thus far by specialists continue to be unsatisfactory. It is accordingly an object of the invention to provide a device for the generation of thermal neutrons, which overcomes the hereinafore-mentioned disadvantages of the heretofore-known devices of this general type and which improves the ratio of the source capacity to the thermal neutrons produced With the foregoing and other objects in view, there is provided, in accordance with the invention, a device for the generation of thermal neutrons, comprising a gas-filled housing having a wall with inner surfaces and a neutron outlet opening, a spherical moderator housing disposed in the gas-filled housing, a transportable neutron source disposed in the spherical moderator housing, and cups formed of moderating material disposed between the moderator housing and the inner surface of the wall of the gas-filled housing, the cups being spaced at a given distance from the inner surface of the wall of the housing, from the moderator housing and from each other defining spaces between the cups, and the spaces between the cups defining openings directed toward the neutron outlet opening. The moderator cups serve both for the reflection of the neutrons primarily thermalized in the moderator housing and for the secondary moderation of a not inconsequential number of higher energy neutrons (E>E.sub.therm) exiting the moderator housing. In accordance with another feature of the invention, the cups are coaxial to the moderator housing. This improves the reflection characteristic. In accordance with an added feature of the invention, the housing wall is in the form of another outer cup having free ends forming the neutron outlet opening. This improves the effect of the secondary moderation. In accordance with an additional feature of the invention, there is provided neutron-permeable material enclosing the moderating material. This is done if a liquid is used as moderator. In accordance with a further feature of the invention, there are provided brackets fastened to the cups and extended between the moderator housing and the inner surface of the wall of the gas-filled housing. This maintains the specified spacings between the cups. In order to form a neutron holder, in accordance with yet another feature of the invention, the spherical moderator housing or ball has a step-shaped canal formed in the center thereof defining at least one shoulder and ending at opposite sides of the gas-filled housing, the neutron source being supported on the at least one shoulder. Due to the fact that the canal passes through from one side of the housing to the other, on one hand it is easy to introduce the neutron source into the moderator sphere or to remove it therefrom and on the other hand, if the housing is disposed in a water seal of purifier, there is assurance that water will flow through the canal. In accordance with yet a further feature of the invention, the moderator housing has a center, the neutron outlet opening is disposed in a given plane and the cups are in the form of half shells having circular, ring-shaped openings in a plane running parallel to the given plane of the neutron outlet opening and intersecting the center of the moderator housing. The neutrons secondarily moderated or reflected in the vicinity of the cups thus reach the neutron outlet opening along the shortest path. In accordance with a concomitant feature of the invention, the gas-filled housing includes a cylindrical part, and including a collimator with a collimator inlet side and at least one collimation path, the collimator inlet side having a plastic or synthetic plating, coating or lining disposed thereon defining a free, open collimation path, and a plastic or synthetic plating, coating or lining disposed on the cylindrical part of the housing extending in a continuous taper to the collimator inlet side and forming part of the plating on the collimator inlet side. It is therefore seen that it is possible to use the device in combination with a collimator. With this embodiment, the higher energy neutrons still flowing in the direction toward the neutron outlet opening after their primary moderation, are also moderated secondarily, thus contributing to a further improvement of the efficiency of the neutron source. 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 device for the generation of thermal neutrons, 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.
043449123	description	DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS In the thorium fuel cycle protactinium-231 and uranium-232 are produced primarily by the following chain of neutron reactions: EQU Th.sup.232 (n,2n)TH.sup.231 .fwdarw..beta..fwdarw.Pa.sup.231 EQU Pa.sup.231 (n,.gamma.)Pa.sup.232 .fwdarw..beta..fwdarw.U.sup.232 The pertinent decay data for these nuclides is as follows: ______________________________________ Decay Mode Half Life ______________________________________ Th.sup.231 .beta. 26 hours Pa.sup.231 .alpha. 3 .times. 10.sup.4 years Pa.sup.232 .beta. 1.3 days U.sup.232 .alpha. 72 years ______________________________________ The uranium-232 decay chain contains eight radioactive daughter nuclides. Six alpha particles are emitted in the declay from uranium-232 to stable lead-208. The alphas emitted from uranium-232 have energies of about 5.3 MeV. However, the total energy in the complete decay chain, from uranium-232 to stable lead-208, is about 42 MeV. This includes about 35 MeV for the six alpha particles and about 7 MeV for the beta particles and gamma rays which are also emitted. A portion of the above enumerated 7 MeV comes from two daughters of uranium-232, namely, thallium-208 and bismuth-212, which emit high energy gamma rays the energy range being from about 1.6 to 2.6 MeV. Therefore, the presence of uranium-232 in nuclear fuel can serve as a deterrent to misappropriation both because of the high energy gamma rays emitted from the daughter products and because of the heat produced by the decay chain. The deterrent effect of high energy gamma rays associated with uranium-232 has long been recognized. However, the recycling of protactinium-231 in order to increase the production of uranium-232 has not been considered. Further, the deterrent effect of the heat produced by a high concentration of uranium-232 has not been heretofore taken into account. This heating effect is similar to the deterrent effect of heating produced by the presence of plutonium-238 in plutonium fuel as noted in the previously mentioned Nuclear Energy article. The thorium fuel cycle may be used in light water reactors, heavy water reactors, high temperature gas cooled reactors, and fast breeder reactors. Recycling or recovery of protactinium-231 is theoretically possible for any of these reactors. Irradiation of protactinium recovered from the thorium fuel cycle may also be accomplished in reactors which operate on the uranium or plutonium fuel cycle. However, this disclosure presents calculations only for light water reactors. Two types of light water reactors are considered. The first type is called a pre-breeder, and is fueled with thorium and moderately enriched natural uranium. The second type is called a breeder and is fueled with thorium and uranium-233. Pre-breeder reactors are operated to produce heat for the generation of electricity, and also to produce uranium-233 for the initial fuel charge for breeder reactors. Breeder reactors are self-sustaining, that is, once a breeder reactor is operating to produce heat for the generation of electricity, it also produces as much uranium-233 as it consumes. Preliminary depletion calculations have been made on conceptual light water pre-breeder and breeder cores to establish the potential for the production of both protactinium-231 and uranium-232. The calculations assume a reactor with a thermal power rating of 3000 MW thermal (equivalent to 1000 MW electrical) and operating at an 80% load factor. The important features of these reactors are presented in Table 1 below. ______________________________________ Fraction Fuel.sup.(a) of Core Volume Core Thorium Refueled Number Coolant Volume Loading Each of Volume (cm3) (MT) Year Modules ______________________________________ Pre- breeder .504 2.85 .times. 10.sup.7 75 1/3 88 Breeder 1.86 6.36 .times. 10.sup.7 280 1/3 157 ______________________________________ .sup.(a) Fuel volume comprises only fuel pellets. The results of the calculations are as follows: Pre-breeders would produce about 2.3 kg per year of protactinium-231. However, breeders would produce about 6.1 kg per year. Neutron irradiation of separated protactinium-231 could be conducted efficiently to convert protactinium-231 to uranium-232. For instance, a one year irradiation of protactinium-231 in special rods in a pre-breeder reactor would convert about one third of the protactinium-231 to uranium-232. However, about 10% of this produced uranium-232 would be destroyed by neutron capture within the reactor. Allowing for reasonable reprocessing and fabrication losses of about 2% at each step results in the conclusion that about 80% of the protactinium-231 produced in pre-breeders or breeders is made available as uranium-232 for adding to an uranium fuel. Fuel for conventional light water reactors could be mixed or "spiked" with uranium-232 in order to discourage attempts to bring the enrichment up to weapons grade level. Assuming that the enrichment is approximately 3.2% and the "spike" is 1000 ppm uranium-232 in uranium-235, the pre-breeder and breeder annual outputs of protactinium-231 would provide enough uranium-232 to "spike" from 2 to 5 annual fuel reloadings of a conventional 1000 MW electrical reactor. Alternatively, all or part of this protactinium-231 can be recycled into pre-breeder or breeder fuel in order to increase the concentration of uranium-232 in the uranium-233. If all of the protactinium-231 produced by a pre-breeder were recycled into pre-breeder fuel, the concentration of uranium-232 in uranium-233 would increase to approximately 12,000 ppm or more. Approximately the same concentration would result if all of the protactinium-231 produced by a breeder were recycled into breeder fuel. The effect on fissile inventor ratio of the breeder would be insignificant. It should be noted that the heating rate due to alpha particles from the uranium-232 decay is supplemented by alpha particles, betas and gammas which result from the decay of the daughter products. The ratio of decay energy from uranium-232 plus the energy arising from the decay of the daughters to the decay energy of uranium-232 alone is given in Table 2 below. ______________________________________ ##STR1## Years After Separation of U-232 From Daughter Products Ratio ______________________________________ 0.5 2.2 1.0 3.1 1.5 3.9 2.0 4.6 3.0 5.6 4.0 6.3 ______________________________________ For example, one year after separation from the daughter products, this ratio is approximately 3 to 1. Since the half life of uranium-232 is shorter than that of plutonium-238, one year after the separation of the uranium-232 from the daughter products, the heating rate due to 1% uranium-232 in uranium would be equivalent to the heating rate due to 3.6% of plutonium-238 in plutonium in the process described previously. If 100% of the protactinium-231 produced by a breeder reactor were recycled into the reactor fuel, the radiation level for a typical fresh breeder module would be approximately 200 roentgens per hour at a distance of one meter. By contrast, the radiation level for a typical fresh breeder module without recycling of protactinium-231 would be approximately 48 roentgens per hour at a distance of one meter. This increased level of radiation would therefore act as a deterrent to the theft or concealment of nuclear fuel. The process disclosed herein for adding uranium-232 to nuclear fuel could be combined with other known methods for making nuclear fuels highly radioactive. For instance, in the Nuclear Industry trade journal article referenced above, the addition of neptunium-237 to nuclear fuel was discussed. "Spiking" fuel with a combination of neptunium-237 and uranium-232 would produce a fuel element with excellent diversion resistance. Spiking fuel with uranium-232 could also be combined with the known CIVEX process to make fuel with improved long term misappropriation resistance. The CIVEX process has as an input the stored spent fuel from light water reactors. The output of this process is radioactively hot refabricated fuel ready for insertion into reactors. In the processing stage, highly radioactive elements of the spent fuel are added to the replenished new fuel for the reactors. Thus, the fresh fuel contains a fraction of the fission products in a mixture of uranium and plutonium, and such fresh fuel can be remotely fabricated as the final stage of the CIVEX plant. Combining this process with the addition of uranium-232 into the fuel could overcome some of the criticisms of the CIVEX process, the main criticism being that, although the CIVEX process produces fuel containing certain radioactive fission products, these fission products have relatively short half lives such that, within one year and a half to four years after fabrication of the fuel, the fuel loses its proliferation resistance. On the other hand, the concentration of radioactive daughter products of uranium-232 in the process disclosed herein increases within this time frame, and the uranium-232 itself has a half life of 72 years. Thus, in fresh fuel containing a combination of the CIVEX fission products and uranium-232, the decrease in the radioactivity from the CIVEX fission products is compensated by the buildup of the radioactivity from the uranium-232 daughter product. In cases where protactinium-231 is neutron irradiated separately to produce uranium-232 for use as a "spike," thorium-228 produced by decay of the uranium-232 may be recovered along with the uranium-232 and used as part of a "spike." Modification of the invention may be possible without departing from the spirit and the scope of the appended claims.
	description	This application is a division of U.S. application Ser. No. 13/861,004 filed on Apr. 11, 2013, now U.S. Pat. No. 10,446,280, which claims the benefit of U.S. Provisional Application No. 61/625,740, filed Apr. 18, 2012, the disclosures of which are hereby incorporated by reference in their entirety. The following relates to the nuclear reactor arts, nuclear power generation arts, nuclear reactor control arts, nuclear reactor human-machine interface (HMI) arts, and related arts. Nuclear power plants are highly complex and include numerous systems to ensure safe operation. By way of illustrative example, a typical nuclear power plant employing a pressurized water reactor (PWR) includes: the nuclear reactor containing a nuclear reactor core comprising fissile material (e.g. 235U) immersed in primary coolant water and ancillary components such as a pressurizer and reactor coolant pumps (RCPs); a control rod drive system including control rods, control rod drive mechanisms (CRDMS) and ancillary components designed to insert neutron-absorbing control rods into the nuclear reactor core to extinguish the nuclear chain reaction (either during normal shutdown, e.g. for refueling, or in response to an abnormal condition, i.e. a scram); a steam generator in which primary coolant heats secondary coolant to generate steam; a turbine driven by the steam; an electric generator turned by the turbine to generate electricity; a complex switchyard providing the circuitry to couple the output of the generator to an external electric grid; a condenser for condensing the steam; piping with valving and ancillary components for conducting feedwater and steam between the various components; one or (typically) more house electrical systems for providing electrical power to the RCPs and other electrically driven components; backup power sources (typically diesel generators and batteries); an emergency core cooling system (EGGS) to dissipate residual heat still generated by the nuclear reactor core after shutdown of the chain reaction; ancillary cooling water systems supplying components such as the condenser; and so forth. A boiling water reactor (BWR) is similar, except that in a BWR primary coolant boils in the pressure vessel and directly drives the turbine. These numerous systems interact with one another. A malfunction of one component may trigger responses by other systems, and/or may call for the operator to perform certain operations in response to the malfunction. Existing control rooms for nuclear power plants typically include a control panel for each component, sub-system, or other operational unit. The resulting layout is unwieldy, including numerous control panels with typically dozens of video display units (VDUs) along with additional indicator lights, and various operator controls such as touch-screen VDU interfaces along with switches, buttons, and so forth. The control panels are arranged to form a horseshoe-shaped arc of about 90° or larger, and inside of this arc further control panels are installed as bench boards. These vertical and bench-mounted control panels include readout displays, indicators, and controls for all components, valves, electrical switches, circuit breakers, piping, and so forth. The arced configuration enables an operator at the controls (OATC) to view all controls simultaneously or with a small turn to the left or right. Substantial effort has been expended in optimizing control room ergonomics, for example by placing the most critical and/or frequently used control panels near the center of the arc. The VDUs are typically designated as safety- or non-safety related, with usually around a dozen safety-related VDUs near the center of the arc or at centrally located bench boards, and the two dozen or more non-safety related VDUs distributed around the periphery. Nonetheless, the control room is complex. A staff of five or more human operators is usually required around the clock. Response to a given situation may require accessing several control panels, which may be located at different places along the vertical arc and/or at different bench boards. When an abnormal situation arises, it typically results in numerous alarms being set off at various control panels associated with the various components affected by the abnormal situation. One (or possibly more) alarm indicates the “root cause” of the abnormal situation, while the other alarms indicate various automated responses to the root cause, consequent operational deviations, or additional problems triggered by the root cause. For example, a failure of the condenser will cause automated shutdown of the turbine, interrupts the steam flow, trips the reactor and brings the EGOS online; and, as further consequences reactor pressure and temperature likely will rise and various electrical systems may also react. Each of these events is unusual and generates an alarm, and this cascade of alarms occurs over a relatively short time interval, with some alarms activating almost simultaneously from the operators' point of view. The on-site human operators then confer to decipher the sequence of events that have led to these alarms, and agree upon appropriate remedial action. In making the diagnosis, operators may need to move around the control room to review various control panels. Yet, operator response should be swift to alleviate the abnormal situation. Any error in diagnosing the root cause may result in incorrect remedial action which can delay resolution of the root cause and may possibly introduce further problems. Disclosed herein are improvements that provide various benefits that will become apparent to the skilled artisan upon reading the following. In accordance with one aspect, a reactor control interface comprises a home screen video display unit (VDU) configured to display: blocks representing functional components of a nuclear power plant including at least (i) blocks representing functional components of a normal heat sinking path of the nuclear power plant and (ii) blocks representing functional components of at least one remedial heat sinking path of the nuclear power plant, and connecting arrows of a first type connecting blocks that are providing the current heat sinking path wherein directions of the connecting arrows of the first type represent the direction of heat flow along the current heat sinking path. In accordance with another aspect, a method operates in conjunction with video display units (VDUs) of a reactor control interface wherein the VDUs include a group of safety VDUs and an additional VDU that is not a safety VDU. The method comprises: detecting a malfunctioning safety VDU, the remaining safety VDUs being functioning safety VDUs; shifting the displays of the functioning safety VDUs to free up one of the functioning safety VDUs wherein the shifting transfers the display of one of the functioning safety VDUs to the additional VDU that is not a safety VDU; and transferring the display of the malfunctioning safety VDU to the functioning safety VDU freed up by the shifting. In accordance with another aspect, a non-transitory storage medium stores instructions executable by an electronic data processing device in communication with a video display unit (VDU) to perform a method comprising: displaying a home screen representing a nuclear power plant, the home screen including blocks representing functional components of the nuclear power plant including at least (i) blocks representing functional components of a normal heat sinking path of the nuclear power plant and (ii) blocks representing functional components of at least one remedial heat sinking path of the nuclear power plant, and connecting arrows of a first type connecting blocks that are providing the current heat sinking path wherein directions of the connecting arrows of the first type represent the direction of heat flow along the current heat sinking path; and in response to the nuclear power plant transitioning to a different heat sinking path, updating the connecting arrows of the first type by deleting and adding connecting arrows of the first type so that the updated connecting arrows of the first type represent the different heat sinking path. Disclosed herein are improved control room designs that substantially enhance the effectiveness of the nuclear power plant operators. In existing control rooms for nuclear power plants, a large number of VDUs (e.g. 30, 40, or even more VDUs) are employed in order to ensure that all relevant data are displayed at all times. However, it is recognized herein that the large number of VDUs can actually reduce operator effectiveness because it is not possible for the operator (or even a crew of five, six, or more operators) to monitor all VDUs simultaneously. Moreover, the large area over which this large number of VDUs must be distributed requires operators to move about the control room in order to view the various VDUs. In control room embodiments disclosed herein, this large multiplicity of VDUs is replaced by a smaller number of VDUs, e.g. about 5-7 VDUs. To accomplish this, it is necessary to employ hidden windows. In other words, not all the information of the conventional 30, 40, or more VDUs can be displayed on the 5-7 VDUs of the disclosed control room embodiments. Nonetheless, all vital information must be displayed so that it is guaranteed that there is no possibility that the operator at the controls (OATC) will miss a safety-related event. To achieve this fail-safe display of all vital information, it is disclosed herein to provide a main display that focuses operator attention on the overriding concern of maintaining a safe heat sinking path for the nuclear reactor core. It is recognized herein that this single aspect of nuclear power plant operation captures all possible safety-related events. In normal operation, the heat sinking path for a pressurized water reactor (PWR) is the following steam cycle (where “RCS” is “reactor coolant system”, “PC” is “primary coolant”, and “SC” is “secondary coolant”): Nuclear core→RCS→SC feedwater→SC steam→turbine→condenserwhere the condenser converts the secondary coolant steam back to secondary coolant feedwater while rejecting heat to circulating water. Heat is also rejected to the electric generator by action of the turbine—a portion of this heat is converted to electricity while the remainder is converted to heat in the generator. A small portion of heat is also rejected in the turbine itself, resulting in some steam condensation inside the turbine, and the condensate is also fed back to the secondary coolant feedwater system. The steam cycle of a boiling water reactor (BWR) is similar, except that there is no for steam generator and primary coolant boiled in the pressure vessel directly drives the turbine: Nuclear core→RCS→PC steam→turbine→condenser In any deviation from normal operation, a safe heat sinking path must be maintained. For example, if the primary coolant exceeds a safe threshold, the reactor scrams and the emergency core cooling system (EGGS) takes over to reject residual heat from the shut-down nuclear reactor to an ultimate heat sink (UHS) in the form of a large body of water, cooling tower, or so forth. Here the safe heat sinking path (for both PWR and BWR) is: Nuclear core→RCS→ECCS→UHSNote that here the heat being generated in the reactor core is not due to an operating nuclear chain reaction (that having been extinguished by the scram and possibly by other measures such as injection of soluble boron neutron poison), but rather is due to residual decay heat produced as short half-life reaction byproducts decay. As another example, in the case of a loss of coolant accident (LOCA) the reactor again scrams, and the safe heat sinking path for the residual decay heat is: Nuclear core→RCS→Containment→ . . . →(UHS or ambient)In this situation, the LOCA vents primary coolant steam into the containment. The containment prevents any radiological release. Some type of containment cooling system (indicated by the ellipsis “ . . . ” in the heat sinking path) transfers heat from containment to either the ultimate heat sink or to the ambient air (or both). This heat sinking path may operate in parallel with the heat sinking path through the ECCS. In one nuclear reactor design currently under development (the B&W mPower™ small modular reactor) another contemplated safe heat sinking path employs an auxiliary condenser (“AUX”): Nuclear core→RCS→Steam generator→AUX→ambient In this design, the auxiliary condenser is located outside containment (e.g., a roof-mounted condenser) and is air-cooled by battery-operated fans. The auxiliary condenser is connected with the steam generator, which is internal to the pressure vessel in the mPower™ design (i.e., an integral PWR), so that it provides passive cooling using secondary coolant trapped in the steam generator when main feedwater and steam line valves are shut. In some event scenarios it is contemplated to employ this heat sinking path without scram. It is also contemplated to employ this heat sinking path in combination with heat sinking via the ECCS. The disclosed control room embodiments employ a main or “home” display that is always maintained on a designated VDU. The home display is a functional display of the heat sinking path. The home display does not attempt to show individual valves or other details (although it is contemplated in some embodiments to include one or more principal valves, e.g. main steam and feedwater valves), but rather represents functional blocks. By way of illustrative example, the turbine system is suitably represented as a single block labeled “Turbine” (or another intuitive label). Similarly, the steam system (piping, valves, et cetera) conveying steam from the steam generator to the turbine is represented by a functional block labeled “Steam”, without attempting to display individual pipes or valves. Any noteworthy excursion of the heat sinking path away from its normal operational envelope is highlighted on the home display by a distinctive color and/or another attention-grabbing visual effect (e.g., flashing, boldface, et cetera). This highlighting identifies the functional component that is in an abnormal condition. Components that perform a normal remedial response are highlighted in a different color (and/or other different visual effect) to emphasize that they have responded. In this way, the operator at the controls can immediately identify the root cause of the operational excursion, and can also readily recognize components that are responding normally to the excursion. Additional VDUs of the disclosed control room embodiments provide additional information. In the illustrative embodiments, these additional VDUs provide alarm displays and trend displays. Further VDUs of the disclosed control room embodiments provide control capability. In the illustrative embodiments, these include a procedures/components display and a system mimic display. The procedures/components display enables operations at the procedure-level or component system level, and displays only those procedures that can be performed given the current operational state of the nuclear power plant. The system mimic display provides lower-level control of individual valves and so forth. These VDUs are optionally touch-sensitive or include a pointer-based user input device (e.g. mouse, trackpad, et cetera) and operatively interconnected so that, for example, by touching (or selecting via a mouse) the “Turbine” block on the VDU displaying home screen the turbine mimic is brought up on the mimic display. Optionally, one or more further VDUs provide human-machine interfacing for non-safety related components and systems. In one embodiment, a “non-safety related” component or system is one in which any event occurring in that component or system cannot result in a safety-related operational excursion for at least one hour. Because the disclosed control room embodiments rely upon only a few VDUs, failure of a VDU can be problematic. In some disclosed embodiments, this is addressed using a VDU-shifting scheme by which the display of the failed VDU is shifted to another VDU. Starting with reference to FIG. 1, some illustrative embodiments are described. An illustrative nuclear reactor 1 is of the pressurized water reactor (PWR) type, and includes a pressure vessel 2 comprising an upper vessel and a lower vessel joined by a mid-flange. The pressure vessel 2 houses a nuclear reactor core 4 comprising fissile material, e.g. 235U immersed in primary coolant water. Reactivity control is provided by a control rods system that includes control rod drive mechanisms (CRDMs) 6 and control rod guide frame supports 8. The illustrative CRDMs 6 are internal CRDMs disposed inside the pressure vessel and including CRDM motors 6m disposed inside the pressure vessel; however, external CRDMs with motors mounted above the pressure vessel and connected via tubular pressure boundary extensions are also contemplated. The pressure vessel of the operating PWR contains circulating primary coolant water that flows upward through the nuclear reactor core 4 and through a cylindrical central riser 10, discharges at the top of the central riser 10 and flows back downward through a downcomer annulus 12 defined between the pressure vessel and the central riser to complete the primary coolant circuit. In the illustrative PWR, primary coolant circulation is driven by reactor coolant pumps (RCPs) 14 which may be located where illustrated in FIG. 1 or elsewhere; moreover, natural circulation or the use of internal RCPs disposed inside the pressure vessel is also contemplated. Pressure inside the pressure vessel of the illustrative PWR is maintained by heating or cooling a steam bubble disposed in an integral pressurizer volume 16 of an integral pressurizer 17; alternatively, an external pressurizer can be connected with the pressure vessel by piping. The illustrative PWR is an integral PWR in which a steam generator (or plurality of steam generators) 18 is disposed inside the pressure vessel, and specifically in the downcomer annulus 12 in the illustrative PWR; alternatively, an external steam generator can be employed. In the illustrative integral PWR, secondary coolant in the form of feedwater is input to the steam generator 18 via a feedwater inlet 20, and secondary coolant in the form of generated steam exits via a steam outlet 21. In the alternative case of an external steam generator, the ports 20, 21 would be replaced by primary coolant inlet and outlet ports feeding the external steam generator. The PWR 1 is disposed inside a primary containment 22, which is suitably a steel structure, steel-reinforced concrete structure, or the like. With continuing reference to FIG. 1, the steam outlet 21 of the nuclear reactor delivers steam to a steam line 24 that drives a turbine 26 that turns an electric generator 28 so as to generate electricity that is delivered to an electrical switchyard 30 that feeds an electrical grid (not shown). Steam flows from the turbine 26 into a condenser 32 that condenses the steam to generate feedwater that is delivered by a feedwater line 34 to the feedwater inlet 20 of the steam generator 18 of the integral PWR so as to complete the steam cycle. Condensate generated inside the turbine 26 is also recaptured and added to the feedwater, as indicated by an arrow running from the turbine 26 to the feedwater line 34. The turbine 26, electric generator 28, and condenser 32 are typically housed inside a turbine building 36 (although in some embodiments the condenser may be mounted on top of the turbine building, and other variants are contemplated). In addition to feeding the switchyard 30, the electric generator 28 also delivers house electricity for running pumps, monitors, and other components of the nuclear reactor plant. In the diagrammatically illustrated BOP, the generator 28 feeds a medium voltage a.c. power system 40 which in turn powers a low voltage a.c. power system 42, which in turn powers a d.c. power system 44 that drives a vital a.c. power system 46. It is to be understood that the illustrative nuclear power plant of FIG. 1 is an illustrative example. The disclosed nuclear power plant control room designs are suitably employed in conjunction with an integral PWR-based plant (as illustrated), or with a PWR-based plant employing an external generator, or with a boiling water reactor (BWR) based plant. In the case of a PWR with an external steam generator, the steam generator is typically housed inside containment with the pressure vessel so that the steam line 24 and contents of the turbine building 36 remain as illustrated. In the case of a BWR, there is no steam generator; instead, primary coolant boils inside the pressure vessel and is ported out the steam line. In the case of a BWR, the turbine and other steam-handling components are constructed to accommodate potential radioactive contaminants in the steam, which is primary coolant water in the BWR case. With continuing reference to FIG. 1, the nuclear power plant is controlled via a control room 50. FIG. 1 is diagrammatic and does not show the actual physical layout of the nuclear power plant; however, in a typical embodiment a reactor building (not shown) houses the containment 22 (which in turn houses the PWR 1) and the control room 50, while the turbine building 36 is spatially separated by some distance, e.g. a few meters to a few tens or hundreds of meters. As the steam and feedwater lines 24, 34 run between containment 22 and the turbine building 36, keeping the separation relatively short reduces thermal losses in these lines. In the control room, an operator at the controls (OATC) is a human operator who performs control functions via a control station that includes six video display units VDU1, VDU2, VDU3, VDU4, VDU5, VDU6. The six video display units VDU1, VDU2, VDU3, VDU4, VDU5, VDU6 are suitably disposed on an arced table 52 or other arced support that partially encircles the OATC, so that the OATC has ready access to any of the six units. VDU5 shows the home screen providing a functional diagram of the nuclear power plant that highlights the heat sinking path and operational status of functional blocks. VDU3 and VDU4 are control units that enable the operator to control systems of the power plant. VDU3 is the system mimic display and enables low level control of individual components, while VDU4 is a procedures and components display that enables initiation of procedures performed by systems or groups of systems. The procedures available to be performed are stored in a procedures database 54, and the procedures and components display shows only those available procedures that can be safely performed given the current operational state of the nuclear power plant. VDU2 shows data trends. VDU1 is an alarm display, and in some embodiments sorts alarms by both time-of-occurrence and by priority. VDU6 is an optional unit that displays non-safety related subject matter. In some multiple-reactor nuclear power plants, VDU6 displays common control functions that are shared by both reactors. The subject matter displayed on VDU6 may be under control of someone other than the OATC; additionally or alternatively, if the OATC does control subject matter shown on VDU6 then this is lower priority subject matter. With reference to FIG. 2, the home display shown in VDU5 is presented. Each functional system of the illustrative nuclear power plant of FIG. 1 is represented by a block or icon, e.g. a box with rounded corners in the illustrative home screen of FIG. 2. Thus (and comparing with FIG. 1), in the illustrative example: a block labeled “Fuel” represents the nuclear reactor core 1. A block labeled “Nuclear instrumentation” represents the in-core instruments (not shown in FIG. 1). A block labeled “Control rod drives” represents the complete control rod drives system including the illustrated CRDMs 6 with their motors 6m and the control rods and connecting elements, e.g., spiders, connecting rods (not shown in FIG. 1). A block labeled “Reactor coolant system” represents the reactor coolant system which includes the primary coolant water and its containing pressure vessel 2 along with ancillary components such as the RCPs 14 and the pressurizer 16, 17 that control flow and pressure of the primary coolant. A block labeled “Containment” represents the function of the containment 22. For mnemonic purposes, the containment 22 is also diagrammatically indicated in the home display, but this is optional. The block labeled “Containment” represents the containment in the functional sense, for example its role in the heat sinking path Nuclear core→RCS→Containment→ . . . →(UHS or ambient). Further, a block labeled “Reactor coolant inventory” represents the Reactor coolant inventory and purification system (RCIPS) as a functional unit. A block labeled “Component cooling water” represents the functional system that provides component cooling water to the RCIPS and other components. A block labeled “Chilled water” represents the chilled water supply. A block labeled “Emergency Core Cooling” represents the emergency core cooling system (EGGS). (None of these components are shown in FIG. 1.) With continuing reference to FIG. 2 and compared with FIG. 1, a block labeled “Turbine” represents the turbine 26 as a system. A block labeled “Steam” represents the functional system that generates and conveys steam from the nuclear reactor to the turbine. Thus, the block labeled “Steam” encompasses the steam generator 18, the steam pipe 24, and ancillary valves. A block labeled “Generator” represents the electrical generator 28. A block labeled “Condenser” represents the condenser 32. A block labeled “Switchyard” represents the switchyard 30. The electrical systems 40, 42, 44, 46 diagrammatically indicated in FIG. 1 are represented by corresponding blocks labeled “Medium voltage a.c. power”, “Low voltage a.c. power”, “d.c. power”, and “vital power”, respectively. Additionally, the home screen of FIG. 2 includes a block labeled “Auxiliary a.c. power’ that represents the diesel generators and/or batteries that provide emergency power if the generator 28 is not operating. The home screen of FIG. 2 further includes blocks labeled “circulating water” that represents circulating water that provides the cold water flow for the condenser 32, and a “Turbine control” block representing control systems that control the turbine 26 and generator 28. The home screen of FIG. 2 also includes a block labeled “Auxiliary condenser” representing the auxiliary generator (AUX) of the proposed mPower™ small modular reactor, including the condenser itself and associated cooling fans and control circuitry. (None of these components are shown in FIG. 1.) It should be noted that the illustrative blocks of FIG. 2, which employ textual labels, could be otherwise labeled. For example, in some embodiments a system of three-letter acronyms is employed to label blocks of the home screen, e.g. “CND”=“Condenser”, “RCS”=“Reactor coolant system”, and so forth. It is also contemplated to employ representative symbolic icons, either instead of or in addition to textual or acronym labels. The home screen displayed by VDU5 is a functional block diagram including the blocks representing functional systems as just described, along with arrows selectively connecting blocks. In the illustrative home screen, there are two types of connecting arrows: solid arrows and dotted arrows. The solid arrows represent the heat sinking path of the nuclear power plant in its current operational state. That is, the solid connecting arrows interconnect the displayed blocks that are providing the current heat sinking path, and the directions of the solid connecting arrows represent the direction of heat flow along the current heat sinking path. The dotted arrows are optional, and if included indicate other connections between the displayed functional blocks. FIG. 2 shows the home screen during normal operation of the nuclear power plant of FIG. 1. More generally, connecting arrows of a first type, e.g. solid connecting arrows, represent the current heat sinking path, and arrows of a second type (or of second and third types, et cetera), e.g. the dotted connecting arrows, connect blocks to represent other functional associations between functional blocks but do not represent the current heat sinking path. The normal operational heat sinking path in the form of the steam cycle: Nuclear core→RCS→SC feedwater→SC steam→turbine→condenseris represented by solid arrows in FIG. 2. Specifically, solid arrows from “Nuclear instrumentation” to “Control rod drives” and from “Control rod drives” to “Reactor coolant system” represents the path portion Nuclear core→RCS. Explicit inclusion of “Nuclear instrumentation” and “Control rod drives” in this path portion allows for the home screen to highlight abnormal operation of the reactor core, as indicated by the in-core instruments, or of the control rod drives which control reactivity of the core. In the home screen of FIG. 2, a solid arrow from “Reactor coolant system” to “Steam” represents the path portion RCS→SC feedwater→SC steam in which heat from the reactor coolant system boils secondary coolant feedwater in the steam generator 18. A solid arrow from “Steam” to “Turbine” and from “Turbine” to “Generator” represents the path portion SC steam→turbine in which the generated steam flows from the nuclear reactor 1 to the turbine 26 via the steam pipe 24. (The arrow from “Turbine” to “Generator” specifically denotes the rejection of heat to the generator 28 in this path portion). A solid arrow from “Turbine” to “Condenser” represents the path portion turbine→condenser in which the steam flows from the turbine 26 to the condenser 32 where it is condensed back to form feedwater. An additional solid arrow in the home screen of FIG. 2 running directly from “Turbine” to “Feedwater” represents portion of steam that condense in the turbine 26 and is returned to the feedwater system. With continuing reference to FIG. 2, the dotted connecting arrows indicate other operative connections between functional components that are not directly part of the heat sinking path. For example, the dotted arrows from “Generator” to “Switchyard” and from “Generator” to “Medium voltage a.c. power” denote distribution of the electricity produced by the electric generator 28. These functional connections are important and hence are shown on the home screen to inform the OATC that these connections are in effect, but they do not directly impact the heat sinking. As also seen in FIG. 2, certain functional blocks include numeric annotations. For example, the “Reactor coolant system” block is annotated “2064 PSIG” indicating measured pressure of the primary coolant water in the pressure vessel 2. The “Steam” block includes the annotation “840 PSIG” indicating the measured steam pressure. The “Turbine” block is annotated “100%” indicating the turbine is presently running at 100% capacity. The “Generator” block is annotated with the present electrical power output level “158 MWe”. The “Feedwater” block is annotated with the measured feedwater temperature “325° F.”. The “Medium voltage a.c. power” and “Low voltage a.c. power” blocks are annotated with the current rms voltage levels “4176 VAC” and “483 VAC”, respectively. By providing these annotations on the home screen, the OATC is immediately aware of these parameters which are indicative of the current state of the corresponding annotated functional blocks. With reference to FIG. 3, the home screen shown on VDUS is presented after a failure of the condenser 32 and a consequential trip of the turbine 26 and shutoff of the electrical generator 28. The condenser is the root cause of this abnormal operating condition, and so the “Condenser” block is highlighted by a first highlighting format indicated in FIG. 3 by double-crosshatching. In practice, VDUS is preferably a color display and the “Condenser” block is preferably highlighted in red, as red is an attention-grabbing color, although other colors and/or a flashing display are also contemplated. Thus, the OATC immediately knows that the root cause of the abnormal condition relates to the condenser 32, although the specific mechanism of the condenser failure is not apparent from the home screen. The “Turbine” block is shown with a different highlighting format, represented in FIG. 3 by single-crosshatching. This highlighting, which may in practice be a different color (e.g. green) indicates to the OATC that this component (the turbine 26) is in an abnormal operating condition, but that this abnormal operating condition was caused by something outside of the turbine 26 (namely, caused by the condenser failure in this example). Additionally, the illustrative reactor responds to this condition by bringing the auxiliary condenser online—accordingly, the “Auxiliary condenser” block is highlighted by yet another highlighting format (indicated by wide single-crosshatching in FIG. 3, but in practice preferably by yet another color, e.g. yellow). This third highlighting format informs the OATC that the component is performing a remedial action in accordance with its intended operation. The auxiliary condenser is not in an abnormal operating state, but the fact that it is operating is associated with an abnormal state. The “Auxiliary a.c. power” block is also highlighted by wide single-crosshatching, indicating powering of the fans of the auxiliary condenser system by auxiliary a.c. power (e.g. diesel generators and/or batteries). This highlighting informs the OATC that auxiliary a.c. power is active in accordance with its intended operation. Moreover, the solid arrows have changed to indicate the new heatsinking path, namely Nuclear core→RCS→Steam generator→AUX→ambient. The solid arrows connecting to the “Turbine”, “Condenser”, and “Feedwater” lines are removed as these components are no longer part of the heat sinking path. The solid arrow connecting “Reactor coolant system” to “Steam” remains so as to indicate the RCS→Steam generator path portion which continues to operate, and new solid arrows are shown connecting the “Steam” block to the “Auxiliary condenser” block and connecting the “Auxiliary condenser” block to the “Reactor coolant system” block. These new arrows represent steam flow from the steam generator to the auxiliary condenser (where heat is rejected to atmosphere) and from the auxiliary condenser back to the steam generator (where it is reheated by the RCS). The home screen of FIG. 3 informs the OATC that the condenser has failed (shown by double-crosshatching, e.g. red color, highlighting), and that the turbine has tripped (shown by single-crosshatching, e.g. green color, highlighting), and that the auxiliary condenser has been brought online (shown by wide single-crosshatching, e.g. yellow color, highlighting of both “Auxiliary condenser” and “Auxiliary a.c. power” blocks). Furthermore, the updated solid connecting arrows inform the OATC that a (new) safe heat sinking path is in operation, namely through the auxiliary condenser. For simplicity, FIG. 3 does not include the block annotations shown in FIG. 2; however, they generally remain visible during abnormal operation. In the state shown in FIG. 3, if the auxiliary condenser is unable to provide adequate heat sinking then the pressure annotation of the “Reactor coolant system” block will begin rising reflecting a rising primary coolant pressure. With reference to FIG. 4, the home screen is shown after the primary coolant pressure has risen above a first threshold. This pressure violation is indicated by applying the first highlighting format (double-crosshatching, e.g. red) to the “Reactor coolant system” block. Although this pressure violation is not technically a “root cause” of an abnormal state (the condenser failure is the root cause), it is not an expected consequence of the condenser failure. Rather, in some instances the auxiliary condenser will provide adequate heat sinking and the pressure violation will not occur. The fact that the pressure violation has occurred can therefore be thought of as a new or supplemental root cause—it leads to the expected response of scramming the reactor, i.e. dropping the control rods to extinguish the nuclear chain reaction. This is indicated in the home screen by coloring the “Control rod drives” block with the second highlighting effect (single crosshatching, e.g. green). In an alternative embodiment, the “Control rod drives” block is colored with the third highlighting (wide single-crosshatching, e.g. yellow) since the scram is a remedial action performed in accordance with its intended operation. However, since scram is something that it is desired that the OATC immediately notices, using the more aggressive second highlighting effect, as illustrated in FIG. 4, is advantageous. In the illustrated response sequence, the scram does not immediately lead to bringing the ECCS online. In the illustrative reactor, it is hoped that by scramming and hence extinguishing the nuclear chain reaction, the auxiliary condenser may thereafter be able to handle rejection of the residual decay heat, so that the ECCS may not need to be brought online. However, if the auxiliary condenser is not able to keep up with the residual decay heat, then the primary coolant pressure will continue to rise in the state shown in FIG. 4. With reference to FIG. 5, the home screen is shown after the continually rising primary coolant pressure has risen above a second threshold that is higher than the first threshold. This pressure violation is “supplemental” to the violation of the first threshold, so the “Reactor coolant system” block merely remains with the first highlighting format (double-crosshatching, e.g. red). The ECCS is brought online responsive to violation of the second pressure threshold, and this is indicated in FIG. 5 by coloring the “Emergency Core Cooling” block with the second highlighting effect (single crosshatching, e.g. green). Again, in an alternative embodiment, the third highlighting (wide single-crosshatching, e.g. yellow) could instead be used since the ECCS is performing a remedial action in accordance with its intended operation. Additionally, a new solid connecting arrow is added, running from the “Reactor coolant system” block to the “Emergency Core Cooling” block. This solid arrow indicates activation of another heat sinking pathway: Nuclear core→RCS→ECCS→UHS. Note that the illustrative home screen does not include a functional block representing the UHS (i.e. ultimate heat sink). However, it is contemplated to include such a functional block, in which case a further solid connecting arrow would suitably run from the “Emergency Core Cooling” block to the UHS block. In the illustrative example, the auxiliary condenser remains online after the ECCS is brought online, and so the solid connecting arrows indicating the heat sinking path involving the auxiliary condenser remain in FIG. 5. Alternatively, if the auxiliary condenser is taken offline concurrently with bringing the ECCS online, then these arrows for the auxiliary condenser heat sinking path would be turned off in FIG. 5. The sequence of FIGS. 2-5 illustrates how the home screen provides the OATC with a rapid and accurate assessment of the root cause of the problem and its consequences. With reference to FIG. 6, the alarm register display on VDU1 is shown for the system in the state shown in FIG. 5. In other words, the alarm register display of FIG. 6 is displayed on VDU1 concurrently with the display of the home screen of FIG. 5 on VDUS. The illustrative alarm register includes two sortable alarms lists: the list in the left window shows alarms listed in reverse chronological order, that is, by reverse time sequence (with the most recent alarm on top; alternatively, the list can be in chronological order, i.e. with the oldest alarm on top). The list in the right window shows the alarms ordered by priority. The alarm register uses the same highlighting formats as are used in the home screen. Thus, for example, the alarm indicating the condenser is offline is in the first highlight format, e.g. in red color, as this is the highest priority alarm. The alarm indicating turbine trip is in the second highlight format, e.g. in green color. The alarm indicating auxiliary condenser online is in the third highlight format, e.g. in yellow color. And so forth. The (left-hand) list in reverse chronological order is advantageous in tracing the sequence of events, while the (right-hand) list sorted by priority allows the OATC to identify the most urgent alarms. To assist in tracing the alarm history it is contemplated to label the alarms by time-of-occurrence in the left hand reverse chronological view (time stamps not shown in FIG. 6). It is noted that the (left-hand) list in chronological order includes two RCS overpressure alarms—the first occurred when the RCS pressure exceeded the lower first threshold (triggering scram), and the second occurred when the RCS pressure exceeded the higher second threshold (triggering placement of the EGGS online). In the (right-hand) list by priority, only the second alarm (RCS pressure exceeding the second threshold) is listed, since this alarm subsumes the alarm for RCS pressure exceeding the first threshold. In some embodiments, alarms are removed from the (right-hand) priority list when the underlying condition is remediated. It will be appreciated that the order of the lists can be reversed, i.e. the priority list can be on the left and the chronologically ordered list on the right. It is also contemplated to provide operator controls (not shown) to allow the OATC to sort the alarms shown in the right-hand window by various sorting criteria. VDU1 has its screen split vertically into two alarm registries which display the same information, but in different formats. The left side of the display shows alarms chronologically organized, e.g. listed in reverse chronological order with the most recent alarm on top, and optionally including time-stamps. In this example, sorting, filtering, and other visual manipulations disabled in the left-hand window, so that the OATC must view all alarms. The right side of the display shows alarms sorted by priority, with the highest priority alarms at the top. Optionally, the OATC has the ability to sort, filter, or re-arrange alarms in the right-hand window in order to display meaningful data to the current task. With reference to FIG. 7, an illustrative configuration for the multi-trend display on VDU2 is shown. The illustrative configuration employs “hidden” windows that are operator-selectable using selection tabs at the bottom of the view (suitably selected by touch if VDU2 is a touch-sensitive screen, or using a mouse pointer, or so forth). The illustrative selection tabs include: “PWR”; “LOW PWR”; “EOP”; “SOP”; “REFUEL”; “START-UP”; and “SHUT-DOWN”. Additional or other tabs are also contemplated for different situations. The illustrative multi-trend view includes a relatively larger central window surrounded by relatively smaller peripheral windows. For each view (corresponding to a selected tab) the trends displayed in the various peripheral windows are in a fixed arrangement. Thus, in the illustrative example, the “PWR” tab is selected and “Trend 4” is displayed in the upper right peripheral window. This is then done consistently—in the “PWR” view the upper right peripheral window always displays “Trend 4”, and the operator cannot reorder the peripheral windows (e.g., using a drag-and-drop process). In this way, it is ensured that for a given tab (e.g. the “PWR” tab) the OATC always sees the same arrangement of trends in the multi-trend display on VDU2. In this way, the OATC can gain familiarity with this layout and, with experience, immediately knows that the upper right peripheral window is displaying “Trend 4”. The relatively larger central window, on the other hand, displays an operator-selected trend. For example, at the instant shown in FIG. 7 the larger central window is displaying “Trend 8”. Selection of the view to display in the central window is suitably done by touch (for a touch-screen) or mouse selection of the peripheral view. Thus, by clicking the mouse cursor on the peripheral window showing “Trend 8” the OATC can display “Trend 8” in the central window (as shown). This allows the OATC to select a particular trend for inspection in the central window, while still seeing all of the other trends of that view in the peripheral windows. Note that in order to maintain the fixed pattern of peripheral windows, if no data is available for a given trend the corresponding peripheral window continues to be dedicated to that unavailable trend, as is the illustrative case for “Trend 9” in the lower left peripheral window of FIG. 7. In the illustrative example of FIG. 7, VDU2 can show up to twelve real-time graphs in the peripheral window based on the current plant state (additional or alternative to being based on an OATC-selected tab as in FIG. 7; also note that in the view shown in FIG. 7 only ten of the possible twelve peripheral windows are being used to show trends with the bottom rightmost two available peripheral window slots being unused in the illustrative “PWR” view). Graphs are arranged around the perimeter of the screen with a blank center area, and the OATC can select a graph to display in the center blank area. When a graph is displayed in the center, it is enlarged (while maintaining the aspect ratio) to enhance visibility for the operator Graphs may contain one or more trends. Each graph can zoom, pan, pause, display historical data, or so forth. The OATC optionally may choose to ‘stack’ multiple graphs in the center area, and stacked graphs are aligned by the x-axis (time) so that trends may be compared with respect to time. Tabs or buttons are optionally displayed horizontally across the bottom of the screen (as per FIG. 7) to display the trends relative to that plant state. The multi-trend display suitably defaults to the tab that corresponds with the current plant state and display the graphs associated with that tab. VDU3 shows a system mimic display. This display provides low-level control (e.g. of individual valves, switches, or so forth) for a given system. VDU3 employs “hidden” windows insofar as the OATC can select the system whose mimic is displayed. In some embodiments, this can be done by touching (or mouse-clicking) the corresponding system block in the home view of VDU5—for example, touching or mouse-clicking the “Turbine” block brings up a turbine control mimic on VDU3. To access lower-level components (e.g. a particular part of the turbine 26) a drill-down approach can be performed on VDU3, e.g. by clicking on a part of the turbine mimic an enlarged view of the selected area is shown. Other known graphical user interface (GUI) navigation techniques can additionally or alternatively be employed, such as having a set of tabs for different components. With reference to FIG. 8, an illustrative embodiment of VDU3, which displays the system mimic, is shown. This screen displays a mimic 60 of a current system (selected by the operator) in a simplified form. Mimics suitably consist of components such as piping, valves, pumps, heat exchangers, tanks, et cetera. Graphical components of a mimic are suitably drawn in diagrammatic form and extraneous information removed to increase salience of mission critical components. In one suitable configuration, the current system mimic is displayed in the center of the screen with narrow columns 62, 64 on far left side and right side, respectively, for navigation to interfacing systems, and navigation aids are displayed in color corresponding to the current system state. In some embodiments, a narrow row across the bottom of the screen contains navigation aids 66 to sub-systems that support the current system. These sub-system mimics provide more detailed information about a specific component or section of the mimic. VDU4 displays provides an interface via which the OATC can select to run various pre-defined procedures stored in the procedures database 54. Each procedure has a defined operational space of primary coolant pressure, valve settings, and so forth within which the procedure is allowed to run, and VDU4 preferably displays only that sub-set of procedures that are allowed to run for the current state of the nuclear power plant. In some embodiments, the list of procedures may be further refined by selecting a particular system by touching or mouse-clicking the block representing that system in the home view shown in VDU5. Other known GUI navigation techniques can additionally or alternatively be employed to select the procedure. In some embodiments VDU3 and VDU4 operate in concert, in that a given procedure that is running may stop to request that the OATC perform some low-level operation using VDU3. In such a case the executing procedure causes VDU3 to display the appropriate mimic via which the OATC can perform the low-level operation. Conversely, the procedure running on VDU4 may interlock VDU3 so that the OATC cannot perform a dangerous low-level operation via VDU3 during the procedure. With reference to FIG. 9, an illustrative embodiment of VDU4, which presents the components/procedures display, is shown. In this embodiment, the components/procedures display area is divided into three main sections: (1) a live video feed 70; (2) component data 72; and (3) computer-based procedure 74. The live video feed 70 is, in the illustrative embodiment of FIG. 9, located in the top right corner; and displays two live video feeds for the current system selected (other numbers of live video feeds are also contemplated, e.g. one feed, two feeds, three feeds, et cetera, and the number may be selectable by the OATC, who also has controls for audio, video, play, pause, rewind, rotate, tilt, zoom). The component data section 72 is suitably in the bottom right corner, and displays live data values for a selected component. Tabs 76 may be displayed horizontally across the bottom allow the OATC to select a different component and its associated data. Vertical tabs (not shown) inside the component live data view allow the OATC to select either a tabular display of live data values or live data trends. Vertical tabs aligned to the right of the data display allow the OATC to select either a tabular display of live data values, live data trends, or a component tag task. Optionally, the component data section also allows the OATC to electronically tag or untag components from this tag tab for tag-out, tag-caution, tag-test, and tag-maintenance. For example, when a component is tagged out, it is deemed unavailable by the control room. (For safety, such electronic tagging should be accompanied by physical tagging of the actual component. Also, to ensure accuracy, the tagging options are only displayed for the current component state). The computer-based procedure section 74 in the illustrative embodiment of FIG. 9 occupies the entire left side of the screen. A title at the top of the screen designates the currently selected system, and applicable tasks are listed for the current state of the system. The OATC can select a task to perform and view the task steps required. All task steps are disabled until the OATC acknowledges the component associated with the current step on the system mimic screen shown on VDU3 (e.g. FIG. 10) by touching or mouse-clicking on the component in VDU3. After acknowledgement, the task step is enabled and performed. The process is repeated with each step thereafter. The OATC has the option of reverting to the previous stable condition of the system once a task has been selected or begun. The OATC can also “auto-complete” a task in the event that attention is needed elsewhere. When a task is completed, the list of available system tasks reflects the new current state of the system. In another contemplated option, the OATC can touch or mous-click a component in the system mimic screen of VDU3 to filter the task list for only those which involve the selected component. The home screen shown in VDU5 has been described with reference to FIGS. 2-5, and provides high level indications of the plant status (except balance-of-plant systems). Each system is represented as a rounded rectangle or other diagrammatic block and is arranged on the home screen according to the functional relationships with other blocks. The functional system blocks indicate the state of the system through color coding, e.g. gray to indicate steady state, red to indicate alarm (i.e., the first highlighting format of the example of FIGS. 3-5), yellow to indicate caution (i.e. the third highlighting format of the example of FIGS. 3-5), and green to indicate expected responses (i.e. the second highlighting format of FIGS. 3-5). Relationships between the systems are designated by arrows, with arrowheads designating the direction of the relationship between the two systems connected (that is, input versus output). Input and output functional relationships between the systems are determined based on the state of the plant and vary as the plant state changes. While the example of FIGS. 2-5 employs textual labels for the blocks, in another embodiment each system block is labeled with a three letter acronym for the system. System blocks provide navigation by a touch or mouse-click for the OATC to quickly view the system-level mimic on the system mimic screen of VDU3. Navigation links are provided between home screen (VDU1), computer based Procedures screen (VDU4), and the system mimic screen (VDU3). The home screen (VDUS) is used as a primary starting point for system-system navigation and provides the corresponding system mimic on the system mimic screen (VDU3) and the applicable procedures and component data on the computer-based procedure screen (VDU4). In some embodiments, the computer-based procedure screen (VDU4) is an end-point navigation path (i.e., no navigation paths out of VDU4 are provided in the human-machine interface (HMI) design, only paths that drive information to be displayed on VDU4). The system mimic screen (VDU3) functions as a two-way navigation path from system-to-system as well as system-to-subsystem. The sortable alarm register screen (VDU1) and the multi-trend screen (VDU2) are each independent and provide no navigation to any other screen. System mimics (VDU3) reflect the actual response of the system or component from the action performed by the OATC. Control feedback that does not comply with the expected response of the component/system is indicated through an alarm/warning condition on VDU1 and VDUS. In further regard to navigation, and with brief returning reference to FIGS. 2-5, it will be noted that all functional blocks are shown in all illustrative home views of FIGS. 2-5. This is true even when the system corresponding to a functional block is not operative, e.g. the “Switchyard” block represents the electrical switchyard which is offline for the examples of FIGS. 3-5—nonetheless, the “Switchyard” block remains displayed (albeit with no connecting arrows). This is done because the home view is also a system selection tool. In the foregoing example, although the switchyard is offline, the OATC might want to view certain information about the switchyard, and can select to do so by touching or mouse-clicking the “Switchyard” block. Various sequential action guidance approaches are contemplated. Auto-complete can be used when the current task needs to be completed, but another task takes higher priority for the attention of the OATC. Preferably, each task provides an option for the OATC to “undo” the task steps completed at any point and return the system to the previous safe/stable state. The OATC also has the option of assuming manual control of a component through the component faceplate control in the system mimic screen (VDU3). Computer-based procedures are displayed on computer-based procedure screen (VDU4), and control is directly driven from the computer-based procedures. The available procedures are stored in the procedures database 54 (see FIG. 1), and only applicable procedures for the current selected system are displayed for the current plant mode and system status. A list of procedure titles is displayed as links to navigate to the procedure steps. The list of procedures is optionally filtered by touching of mouse-clicking on a component on the mimic screen (VDU3) to reduce procedure list to tasks that impact that component. In a suitable embodiment of the procedures section of VDU4, all steps of a procedure are visible from the time the procedure is selected until it is completed. Each step is inactive until the previous step is completed. A procedure step is disabled and cannot be performed until the OATC acknowledges the component receiving the action by touching or mouse-clicking on the component in the system mimic VDU3 (to improve situational awareness). When a procedure step is enabled by clicking the component in the system mimic, a checkbox or other selection (e.g. an “OK” button) beside the step on VDU4 is activated and the OATC is able to “check” the box by touch or mouse-click and the action is performed. When a procedure is completed, the final procedure step is to return to the system task menu. As already mentioned, only applicable procedures for the current selected system are displayed for the current plant mode and system status. A procedure is selected by touching or mouse-clicking on the procedure title, similar to selection of a hyperlink on a web page. When a procedure is completed, the list of available procedures will be updated to reflect the change in the system state from the previous procedure completion. Because the number of VDUs is relatively small, e.g. 5-7 VDUs in some preferred embodiments, and 6 VDUs in the illustrative example, it is advantageous to accommodate the possibility that a VDU may malfunction and become inoperative. One approach is to have redundant VDUs on hand; however, it would take time to switch out a defective monitor with a new monitor, and this may be unacceptable. With reference to FIGS. 10 and 11, an approach for addressing an inoperative VDU is illustrated. In illustrative FIG. 10, VDU4 has failed (as indicated by the large “X” crossing out VDU4). VDU4 ordinarily displays the components/procedures display—its unavailability would be a serious problem. To resolve this problem, the functions of the various VDUs shift, as shown in FIG. 11. Thus, VDU3 which formerly displayed the mimic display now displays the components/procedures display. Similarly, VDU2 which formerly displayed the multi-trend display now displays the mimic view. VDU1 which formerly displayed the alarms register now displays the multi-trend display. This leaves the alarms register, which has effectively “shifted off the end”. As seen in FIG. 11, this is accommodated by showing the alarms register on VDU6, which normally displays non-safety information or other “less important” information. To allow the OATC to still access that information, VDU6 also provides a command via which the OATC can temporarily switch VDU6 to show the non-safety information. In the illustrative example of FIG. 11, this is done by pressing the <F1> function key, and a suitable instruction is shown at the bottom of the alarms register displayed on VDU6 in FIG. 11. Because displaying the alarms register is generally more important than displaying the non-safety information, VDU6 is preferably programmed to “time out” the display of the non-safety information and return to the alarm register display if the OATC does not interact with the non-safety display for a certain time interval. By way of illustrative example, the time-out period may be one minute, i.e. when <F1> is pressed the non-safety screen replaces the alarms register on VDU6, and thereafter if no further action is taken VDU6 switches back to the alarm register display after one minute has passed. The defective monitor VDU4 is shown in FIG. 11 displaying the message “Display failure”. This (or a similarly informative) message is advantageously displayed on the defective VDU if the VDU is indeed capable of displaying a textual message. (Of course, if the defect of the defective VDU renders it incapable of displaying anything, then nothing is displayed on the defective VDU). By the disclosed approach of shifting the VDU screens as per illustrative FIGS. 8 and 9, the OATC continues to see something close to the usual arrangement of screens, with the exception that the alarms register is now on the rightmost VDU and VDU4 is blank. This is advantageous since it reduces likelihood of operator confusion. In order for the disclosed VDU shifting scheme to work, the VDUs should all have the user interfacing capability of the VDU with the most complex user interface. For example, VDU1 may not ordinarily need user input capability, since it ordinarily displays the alarms register (as in FIG. 8). However, when the VDU shift shown in FIG. 9 is executed, VDU1 then displays the mimic display, which does require user input (e.g., a touch screen, and/or a mouse, or so forth). Thus, all six VDUs should have the same user interfacing capacity, and indeed are preferably interchangeable. In the illustrative example with six VDUs, failure of more than one VDU cannot be accommodated by the shifting scheme. However, if a seventh monitor (e.g., a second non-safety related monitor) is added then up to two defective monitors can be accommodated. If an eighth monitor is added then up to three defective monitors can be accommodated. In some embodiments, the total number of VDUs is between 5 and 8. Additionally, it is contemplated to include a large (e.g. wall-mounted) overview display that is visible to the shift supervisor and other personnel in the control room, and/or the shift supervisor may have an additional monitoring VDU via which the supervisor can monitor the OATC. Moreover, it is to be appreciated that while the illustrative embodiment includes six distinct VDUs, it is alternatively contemplated to employ a single large-aspect ratio VDU spanning the display area of the illustrative six VDUs, with the functionality of the six individual monitors being provided by six windows displayed on the large-aspect ratio monitor. Said another way, there does not need to be physical separation between the display areas of VDU1-VDU6. The disclosed control room embodiments include a reactor control interface that includes the illustrative VDU1-VDU6 (or some other number of VDUs, e.g. in a range 5-8 VDUs) and a computer or other electronic data processing device (not shown) in communication with electronic data networks and with VDU1-VDU6 and programmed to generate the disclosed displays and to receive and process user inputs as described herein, and to send control signals to various components of the nuclear power plant (in accord with user inputs and/or in accord with automated procedures displayed on VDU4 and executed by the computer or other electronic data processing device). The computer or other electronic data processing device suitably includes or has access to a hard drive or other electronic storage medium that stores the procedures database 54 (see FIG. 1). The computer or other electronic data processing device optionally comprises an interconnected plurality of computers or other electronic data processing devices. For example, in one contemplated embodiment each of VDU1-VDU6 comprises a desktop computer running software implementing the control room. In this approach, the six desktop computers (in the illustrative case of six VDUs) are interconnected via the electronic data network in order to perform intercommunication between the VDUs as described herein. For example, the desktop computer implementing VDU5 suitably communicates selection of a functional block to VDU3 and VDU4 and in response the desktop computers implementing those VDUs display the appropriate component mimic and procedures list, respectively. From the monitor shift example described with reference to FIGS. 10-11, it is apparent that the desktop computer normally implementing VDU3 (the system mimic) must also include software to implement VDU4 (the procedures/components display), and so forth for the other desktop computers. To achieve maximum redundancy in this embodiment, it is advantageous for each desktop computer to include the entirety of the control room software so that the monitor shift described with reference to FIGS. 10-11 can be performed. This also allows swap-out of desktop computers to permanently replace a defective VDU. Indeed, in one implementation of this approach, each desktop computer includes a VDU_type or other indicator as to which VDU the desktop computer implements, and the VDU shift of FIGS. 10-11 then amounts to updating the VDU_type values for the (illustrative six) desktop computers. In another approach, the control room software executes on a central computer not particularly associated with any of VDU1-VDU6, and that central computer generates and transmits the displays to the six VDUs which in this embodiment are “dumb” terminals. In either illustrative embodiment (i.e., the embodiment employing six interconnected desktop computers; or, the embodiment employing a central computer connected with six dumb terminals), the control room computer or interconnected computers are preferably connected with an electronic data network with suitable security provisions. For example, the electronic data network is preferably an isolated network that is connected with the various components of the nuclear power plant in order to receive alarm signals, send control signals, and so forth, but that is preferably not (at least directly) connected with the Internet or other wider area network. If required by the applicable nuclear regulatory agency, the electronic data network may be an entirely wired network; alternatively, if permissible under local nuclear regulations it is contemplated to employ a wireless network or a hybrid wired/wireless network. The disclosed control room embodiments may also be embodied as a non-transitory storage medium storing instructions that are executable by the VDUs comprising a central computer controlling dumb terminals, or alternatively comprising a set of interconnected desktop computers, or alternatively comprising another suitable configuration of display devices and electronic data processing devices, to perform the disclosed control room operations including displaying the various screens (e.g. the home screen, alarms register display, et cetera) and receiving user inputs as described. The non-transitory storage medium may, for example, comprise a hard disk, RAID disk array, or other magnetic storage medium, an optical disk or other optical storage medium, a FLASH memory or other electronic storage medium, various combinations thereof, or so forth. Still further, it is to be appreciated that various disclosed aspects of the illustrative embodiments can be implemented without other disclosed aspects. For example, the disclosed home screen of VDU5 may be implemented as described in the illustrative embodiments (or variants thereof) while the control interfacing may be implemented using techniques other than the disclosed operation of VDU3 and VDU4. Similarly, the disclosed home screen of VDU5 may be implemented as described in the illustrative embodiments (or variants thereof) while the alarm register and/or data trends are/is shown using a format different from that employed in described VDU1 and/or VDU2. As yet another example, the disclosed control room screens (i.e., VDU1-VDU6) can be implemented without the VDU-switching capability described with reference to FIGS. 10-11. Conversely, the VDU-switching capability described with reference to FIGS. 10-11 may be employed with a set of VDUs displaying control room subject matter formatted differently than that described for VDU1-VDU6. The preferred embodiments have been illustrated and described. Obviously, modifications and alterations will occur to others upon reading and understanding the preceding detailed description. It is intended that the invention be construed as including all such modifications and alterations insofar as they come within the scope of the appended claims or the equivalents thereof.
	summary
	description	The present application claims the benefit of priority to U.S. Provisional Application No. 62/463,319 filed Feb. 24, 2017, the entirety of which is incorporated herein by reference. The present invention generally relates to storage of nuclear fuel, and more particularly to an improved seismic-resistant nuclear fuel storage rack system for a fuel pool in a nuclear generation plant. A conventional high density nuclear fuel storage rack is a cellular structure supported on a set of pedestals, as shown in FIG. 1. The bottom extremity of each fuel storage cell is welded to a common baseplate which serves to provide the support surface for the upwardly extending storage cells and stored nuclear fuel therein. The cellular region comprises of a set of narrow prismatic cavities formed by the cells which are each sized to accept a single nuclear fuel assembly comprising either new or spent fuel. The term “active fuel region” denotes the vertical space above the baseplate where the enriched uranium is located. A principal safety function of the fuel rack is to protect the geometry of the “active fuel region” from being adversely affected under any credible accident event, the most severe of them being the plant's postulated earthquake events. A conventional rack has four or more pedestals (see, e.g. FIG. 1). Under an earthquake event, the rack behaves as a cantilever beam exerting significant stresses in the bottom pedestals. The standard practice of fastening the pedestals to the fuel pool's bottom concrete slab has the serious drawback of making the removal of the racks at a future date, submerged in about a 40 feet deep pool of water, extremely onerous. The consideration of convenient decommissioning with minimum dose to the plant personnel and the ability to “rerack” (if necessary), has led the industry to install racks in a “free-standing” configuration. The free-standing rack design configuration has become the dominant method over the past 30 years for installing wet storage capacity for used nuclear fuel in plants around the world. As would be expected, the rack modules are made as large as possible, limited only by the constraint of shipping them from the manufacturing facility to the plant and handling them within the plant for in-pool installation. They are also placed as close to each other as possible in the so-called “high density configuration” to maximize the in-pool fuel storage capacity. The inter-module gap between adjacent fuel racks can be as little as 2 inches in some installations. Free standing fuel racks resist seismic loads primarily by the reactive friction at the pedestal to pool surface interface and the so-called fluid coupling effect. In a conventional free-standing fuel rack, the pedestals are supported on the fuel pool bottom base slab on some type of bearing pad as shown in FIG. 1. However, if the earthquake is strong, then the interface friction may not be adequate to prevent lateral sliding movement or tipping/twisting of the racks, causing them to collide and creating a risk of damaging the cells and compromising the physical integrity of the stored nuclear fuel. An improved earth-quake resistant nuclear fuel rack storage system is desired. Embodiments of the present invention provide a seismic-resistant nuclear fuel rack stabilization system for a fuel pool that seeks to limit the kinematics of the racks and prevent damage to their active fuel region within their cellular structure during severe earthquakes. The present system is thus intended for use in high seismic scenarios, for example if the “zero period acceleration” (ZPA) of any of the earthquake's components exceeds 0.5 g. Features of embodiments of the present seismic-resistant design is that the rack modules are not fastened to the pool slab providing “free standing” fuel racks, but advantageously are substantially restrained against lateral horizontal movement during earthquakes and further provide the hardest location in the body of the module—their baseplates—to serve as the bumper to absorb impact loadings from other adjacent racks under earthquakes. In one aspect, a seismic-resistant nuclear fuel storage system includes: a fuel pool comprising a base slab and plurality of vertical sidewalls collectively defining a cavity configured for wet storage of nuclear fuel; a fuel rack comprising a plurality of vertically elongated tubular cells each defining a prismatic cavity configured for storing nuclear fuel therein, the cells attached to a common baseplate; a plurality of pedestals protruding downwardly from the baseplate; a plurality of spaced apart embedment plates fixedly anchored to the base slab, each embedment plate comprising an upwardly open receptacle having receptacle walls defining a receptacle depth, each receptacle receiving and entrapping one of the pedestals of the fuel rack therein; wherein the embedment plate receptacles are configured such that lateral movement of the fuel rack along the base slab in the event of a seismic event is constrained by engagement between the receptacle walls of each receptacle and the pedestals. In another aspect, a fuel rack stabilization system for seismic-resistant storage of nuclear fuel includes: a fuel pool comprising a base slab and plurality of vertical sidewalls collectively defining a cavity configured for submerged wet storage of nuclear fuel; a plurality of fuel racks supported on the base slab, each fuel rack comprising a plurality of vertically elongated tubes each defining a prismatic cavity configured for storing nuclear fuel therein, the tubes attached to a common baseplate; each fuel rack comprising a plurality of spaced apart pedestals protruding downwardly from the baseplate; a plurality of spaced apart embedment plates fixedly anchored to the base slab, each embedment plate comprising at least one upwardly open embedment cavity having cavity walls, the cavities each receiving and entrapping a respective one of the pedestals of the fuel racks therein; a pool liner secured to the base slab of the fuel pool, the pool liner extending between the plurality of spaced apart embedment plates and having a thickness less than the embedment plate; wherein a perimeter of the embedment plates is hermetically seal welded to the pool liner around all lateral sides to form an impervious barrier to outward leakage of pool water from the fuel pool; wherein the embedment plate cavities are configured such that lateral movement of the fuel rack along the base slab caused by a seismic event is restricted by engagement between the cavity walls of each cavity and the pedestal such that laterally acting seismic forces are not transmitted to the pool liner. In another aspect, a method for seismic-resistant storage of nuclear fuel in a fuel pool comprises: staging first and second fuels racks in a nuclear facility, each fuel rack comprising a plurality of tubes each defining a prismatic cavity configured for storing nuclear fuel therein, the tubes supported on a common baseplate comprising a plurality of pedestals protruding downwardly from the baseplate; lowering the first fuel rack into a water-filled fuel pool comprising a base slab and a metal pool liner secured to base slab; and insertably engaging each of the pedestals of the first fuel rack with corresponding upwardly open receptacles formed in a plurality of spaced apart embedment plates fixedly anchored to the base slab of the fuel pool, each embedment plate hermetically seal welded to the pool to form an impervious barrier to outward leakage of pool water through the base slab of the fuel pool; wherein the embedment plates are configured such that lateral movement of the pedestals along the base slab during a seismic event is restricted by engagement between the pedestals and the receptacles of the embedment plates such that laterally acting seismic forces are not transmitted to the pool liner. In some embodiments, the method may further include lowering the second fuel rack into the water-filled fuel pool; insertably engaging each of the pedestals of the second fuel rack with corresponding upwardly open receptacles formed in the plurality of spaced apart embedment plates fixedly coupled to the base slab of the fuel pool; and abuttingly engaging a peripheral edge of the baseplate of the first fuel rack with an adjoining peripheral edge of the baseplate of the second fuel rack. Further areas of applicability of the present invention will become apparent from the detailed description provided hereinafter. All drawings are schematic and not necessarily to scale. Parts shown and/or given a reference numerical designation in one figure may be considered to be the same parts where they appear in other figures without a numerical designation for brevity unless specifically labeled with a different part number and described herein. References herein to a figure number (e.g. FIG. 1) shall be construed to be a reference to all alphabetical subpart figures in the group (e.g. FIGS. 1A, 1B, etc.) unless otherwise indicated. The features and benefits of the invention are illustrated and described herein by reference to exemplary embodiments. This description of exemplary embodiments is intended to be read in connection with the accompanying drawings, which are to be considered part of the entire written description. Accordingly, the disclosure expressly should not be limited to such exemplary embodiments illustrating some possible non-limiting combination of features that may exist alone or in other combinations of features. Furthermore, all features and designs disclosed herein may be used in combination even if not explicitly described as such. In the description of embodiments disclosed herein, any reference to direction or orientation is merely intended for convenience of description and is not intended in any way to limit the scope of the present invention. Relative terms such as “lower,” “upper,” “horizontal,” “vertical,”, “above,” “below,” “up,” “down,” “top” and “bottom” as well as derivative thereof (e.g., “horizontally,” “downwardly,” “upwardly,” etc.) should be construed to refer to the orientation as then described or as shown in the drawing under discussion. These relative terms are for convenience of description only and do not require that the apparatus be constructed or operated in a particular orientation. Terms such as “attached,” “affixed,” “connected,” “coupled,” “interconnected,” and similar refer to a relationship wherein structures are secured or attached to one another either directly or indirectly through intervening structures, as well as both movable or rigid attachments or relationships, unless expressly described otherwise. It will be appreciated that any numerical ranges that may be described herein shall be understood to include the lower and upper numerical terminus values or limits of the cited range, and any numerical values included in the cited range may serve as the terminus values. Referring to FIGS. 3 and 6, a nuclear facility 30 which may be a nuclear generating plant includes a fuel pool 40 according to the present disclosure configured for storing a plurality of nuclear fuel racks 100. The fuel pool 40 may comprise a plurality of vertical sidewalls 41 rising upwards from an adjoining substantially horizontal bottom base wall or slab 42 (recognizing that some slope may intentionally be provided in the upper surface of the base slab for drainage toward a low point if the pool is to be emptied and rinsed/decontaminated at some time and due to installation tolerances). The base slab 42 and sidewalls 41 may be formed of reinforced concrete in one non-limiting embodiment. The fuel pool base slab 42 may be formed in and rest on the soil sub-grade 26, the top surface of which defines grade G. In this embodiment illustrated in the present application, the sidewalls are elevated above grade. The base slab 42 may be located at grade G as illustrated, below grade, or elevated above grade. In other possible embodiments contemplated, the base slab 42 and sidewalls 41 may alternatively be buried in sub-grade 26 which surrounds the outer surfaces of the sidewalls. Any of the foregoing arrangements or others may be used depending on the layout of the nuclear facility and does not limit of the invention. In one embodiment, the fuel pool 40 may have a rectilinear shape in top plan view. Four sidewalls 41 may be provided in which the pool has an elongated rectangular shape (in top plan view) with two longer opposing sidewalls and two shorter opposing sidewalls (e.g. end walls). Other configurations of the fuel pool 40 are possible such as square shapes, other polygonal shapes, and non-polygonal shapes. The sidewalls 41 and base slab 42 of the fuel pool 40 define an upwardly open well or cavity 43 configured to hold cooling pool water W and the plurality of submerged nuclear fuel racks 100 each holding multiple nuclear fuel bundles or assemblies 28 (a typical one shown in FIG. 13). Each fuel assembly 28 contains multiple individual new or spent uranium fuel rods 28a. Fuel assemblies are further described in commonly assigned U.S. patent application Ser. No. 14/413,807 filed Jul. 9, 2013, which is incorporated herein by reference in its entirety. Typical fuel assemblies 28 for a pressurized water reactor (PWR) may each hold over 150 fuel rods in 10×10 to 17×17 fuel rod grid arrays per assembly. The assemblies may typically be on the order of approximately 14 feet high weighing about 1400-1500 pounds each. The fuel racks 100 storing the fuel assemblies are emplaced on the base slab 42 in a high-density arrangement in the horizontally-abutting manner as further described herein. The fuel pool 40 extends from an operating deck 22 surrounding the fuel pool 40 downwards to a sufficient vertical depth D1 to submerge the fuel assemblies 28 in the fuel rack (see, e.g. FIG. 6) beneath the surface level S of the pool water W for proper radiation shielding purposes. The substantially horizontal operating deck 22 that circumscribes the sidewalls 41 and pool 40 on all sides in one embodiment may be formed of steel and/or reinforced concrete. In one implementation, the fuel pool may have a depth such that at least 10 feet of water is present above the top of the fuel assembly. Other suitable depths for the pool and water may be used of course. The surface level of pool water W (i.e. liquid coolant) in the pool 40 may be spaced below the operating deck 22 by a sufficient amount to prevent spillage onto the deck during fuel assembly loading or unloading operations and to account to seismic event. In one non-limiting embodiment, for example, the surface of the operating deck 22 may be at least 5 feet above the maximum 100 year flood level for the site in one embodiment. The fuel pool 40 extending below the operating deck level may be approximately 40 feet or more deep (e.g. 42 feet in one embodiment). The fuel pool is long and wide enough to accommodate as many fuel racks 100 and fuel assemblies 28 stored therein as required. There is sufficient operating deck space around the pool to provide space for the work crew and for staging necessary tools and equipment for the facility's maintenance. There may be no penetrations in the fuel pool 40 within the bottom 30 feet of depth to prevent accidental draining of water and uncovering of the fuel. In some embodiments, a nuclear fuel pool liner system may be provided to minimize the risk of pool water leakage to the environment. The liner system may include cooling water leakage collection and detection/monitoring to indicate a leakage condition caused by a breach in the integrity of the liner system. Liner systems are further described in commonly owned U.S. patent application Ser. No. 14/877,217 filed Oct. 7, 2015, which is incorporated herein by reference in its entirety. The liner system in one embodiment may comprise a liner 60 attached to the inner surfaces 63 of the fuel pool sidewalls 41 and the base slab 42. The inside surface 61 of liner is contacted and wetted by the fuel pool water W. The liner 60 may be made of any suitable metal of suitable thickness T2 which is preferably resistant to corrosion, including for example without limitation stainless steel, aluminum, or other. Typical liner thicknesses T2 may range from about and including 3/16 inch to 5/16 inch thick. Typical stainless steel liner plates include ASTM 240-304 or 304L. In some embodiments, the liner 60 may be comprised of multiple substantially flat metal plates or sections which are hermetically seal welded together via seal welds along their contiguous peripheral edges to form a continuous liner system completely encapsulating the sidewalls 41 and base slab 42 of the fuel pool 40 and impervious to the egress of pool water W. The liner 60 extends around and along the vertical sidewalls 41 of the fuel pool 40 and completely across the horizontal base slab 42 to completely cover the wetted surface area of the pool. This forms horizontal sections 60b and vertical sections 60a of the liner to provide an impervious barrier to out-leakage of pool water W from fuel pool 40. The horizontal sections of liners 60b on the base slab 42 may be joined to the vertical sections 60a along perimeter corner seams therebetween by hermetic seal welding. The liner 60 may be fixedly secured to the base slab 42 and sidewalls 41 of the fuel pool 40 by any suitable method such as fasteners. Referring now to FIGS. 2-6, a perspective view of a fuel rack 100 according to one embodiment of the present invention is disclosed. The fuel rack 100 is a cellular, upright, prismatic module. Fuel rack 100 may be a high density, tightly packed non-flux type rack as illustrated which is designed to be used with fuel assemblies that do not require the presence of a neutron flux trap between adjacent cells 110. Thus, the inclusion of neutron flux traps (e.g. gaps) in fuel racks when not needed is undesirable because valuable fuel pool floor area is unnecessarily wasted. Of course, both non-flux and flux fuel rack types may be stored side by side in the same pool using the seismic-resistant fuel storage system according to the present disclosure. The invention is therefore not limited to use of any particular type of rack. Fuel rack 100 defines a vertical longitudinal axis LA and comprises a grid array of closely packed open cells 110 formed by a plurality of adjacent elongated tubes 120 arranged in parallel axial relationship to each other. The rack comprises peripherally arranged outboard tubes 120A which define a perimeter of the fuel rack and inboard tubes 120B located between the outboard tubes. Tubes 120 are coupled at their bottom ends 114 to a planar top surface of a baseplate 102 and extend upwards in a substantially vertical orientation therefrom. In this embodiment, the vertical or central axis of each tube 120 is not only substantially vertical, but also substantially perpendicular to the top surface of the baseplate 102. In one embodiment, tubes 120 may be fastened to baseplate 102 by welding and/or mechanical coupling such as bolting, clamping, threading, etc. Tubes 120 include a top end 112, bottom end 114, and a plurality of longitudinally extending vertical sidewalls 116 between the ends defining a height H1. Each tube 120 defines an internal cavity 118 extending longitudinally between the top and bottom ends 112, 114. In the embodiment shown in FIG. 2A-B, four tube sidewalls 116 arranged in rectilinear polygonal relationship are provided forming either a square or rectangular tube 120 in lateral or transverse cross section (i.e. transverse or orthogonal to longitudinal axis LA) in plan or horizontal view (see also FIG. 3). Cells 110 and internal cavities 118 accordingly have a corresponding rectangular configuration in lateral cross section. The top ends of the tubes 120 are open so that a fuel assembly can be slid down into the internal cavity 118 formed by the inner surfaces of the tube sidewalls 116. Each cell 110 and its cavity 118 are configured for holding only a single nuclear fuel assembly 28. It will be appreciated that each tube 120 can be formed as a single unitary structural component that extends the entire desired height H1 or can be constructed of multiple partial height tubes that are vertically stacked and connected together such as by welding or mechanical means which collectively add up to the desired height H1. It is preferred that the height H1 of the tubes 120 be sufficient so that the entire height of a fuel assembly may be contained within the tube when the fuel assembly is inserted into the tube. The top ends 112 of tubes 120 may preferably but not necessarily terminate in substantially the same horizontal plane (defined perpendicular to longitudinal axis LA) so that the tops of the tube are level with each other. The baseplate 102 at the bottom ends 114 of the tubes defines a second horizontal reference plane HR. As best shown in FIGS. 2A-B, tubes 120 are geometrically arranged atop the baseplate 102 in rows and columns along the Z-axis and X-axis respectively. Any suitable array size including equal or unequal numbers of tubes in each row and column may be provided depending on the horizontal length and width of the pool base slab 42 and number of fuel racks 100 to be provided. In some arrangements, some or all of the fuel racks 100 may have unequal lateral width and lateral length as to best make use of a maximum amount of available slab surface area as possible for each installation. For convenience of reference, the outward facing sidewalls 116 of the outboard tubes 120A may be considered to collectively define a plurality of lateral sides 130 of the fuel rack 100 extending around the rack's perimeter as shown in FIGS. 2A-B. Tubes 120 may be constructed of any suitable material usable in a nuclear fuel storage rack. In one embodiment, without limitation, the tubes may be formed of a metal-matrix composite material, and preferably a discontinuously reinforced aluminum/boron carbide metal matrix composite material, and more preferably a boron impregnated aluminum. One such suitable material is sold under the tradename Metamic™. The tubes 120 perform the dual function of reactivity control as well as structural support. Advantageously, tube material incorporating the neutron absorber material allows a smaller cross sectional (i.e. lateral or transverse to longitudinal axis LA) thickness of tube sidewalls 116 thereby permitting tighter packing of cells allowing for a greater number of cells per fuel rack to be provided. The baselate 102 is preferably constructed of a metal that is metallurgically compatible with the material of which the tubes 120 are constructed to facilitate welding. Referring to FIGS. 2-6 (inclusive of all alphabetic subparts), each fuel rack 100 comprises a plurality of legs or pedestals 200 which support rack from the base slab 42 of the fuel pool 40. Pedestals 200 each have a preferably flat bottom end 204 to engage the pool base slab 42 and a top end 202 fixedly attached to the bottom of the baseplate 102. The pedestals 200 protrude downwards from baseplate 102. This elevates the baseplates 102 of the rack off the base slab 42, thereby forming a gap therebetween which defines a bottom flow plenum P beneath rack 100. The plenum P allows cooling water W in the pool to create a natural convective circulation flow path through each of the fuel storage tubes 120 (see e.g. flow directional arrows in FIG. 5). A plurality of flow holes 115 are formed in the rack through baseplate 102 in a conventional manner to allow cooling water to flow upwards through the cavity 118 of each tube 120 and outward through the open top ends 112 of the tubes. Commonly owned U.S. patent application Ser. No. 14/367,705 filed Jun. 20, 2014 shows fuel rack baseplates with flow holes, and is incorporated herein by reference in its entirety. The pool water W flowing through the tubes is heated by the nuclear fuel in fuel assemblies, thereby creating the motive force driving the natural thermal convective flow scheme. Referring now then to FIGS. 3 and 5, flow holes 115 create passageways from below the base plate 102 into the cells 110 formed by the tubes 120. Preferably, a single flow hole 115 is provided for each cell 110, however, more may be used as needed to create sufficient flow through the tubes. The flow holes 115 are provided as inlets to facilitate natural thermosiphon flow of pool water through the cells 110 when fuel assemblies having a heat load are positioned therein. More specifically, when heated fuel assemblies are positioned in the cells 110 in a submerged environment, the water within the cells 110 surrounding the fuel assemblies becomes heated, thereby rising due to decrease in density and increased buoyancy creating a natural upflow pattern. As this heated water rises and exits the cells 110 via the tube open top ends 112 (see FIG. 1), cooler water is drawn into the bottom of the cells through the flow holes 115. This heat induced water flow and circulation pattern along the fuel assemblies then continues naturally to dissipate heat generated by the fuel assemblies Pedestals 200 may therefore have a height H2 selected to form a bottom flow plenum P of generally commensurate height to ensure that sufficient thermally-induced circulation is created to adequately cool the fuel assembly. In one non-limiting example, height H2 of the plenum P may be about 2 to 2.5 inches (including the listed values and those therebetween of this range). Pedestals 200 may have any suitable configuration or shape and be of any suitable type. Some non-limiting examples of shapes that may be used include rectangular or square with a rectilinear lateral/transverse cross sectional shape, cylindrical with a circular cross sectional shape, polygonal with a polygonal cross sectional shape, non-polygonal with a non-polygonal cross sectional shape, or combinations thereof. One combination shown in FIG. 1 is a fixed height pedestal including a rectangular upper portion attached to the fuel rack baseplate 102 and enlarged cylindrical disk-shaped lower portion forming a circular cylindrical foot pad for engaging the fuel pool 40 in a stable manner. FIGS. 2A and 2B show an adjustable pedestal 200, as further described herein. FIGS. 4 and 5 show a fixed height pedestal 200 which may have any of the foregoing mentioned shapes or others. It should be noted that the pedestals 200 described herein for a seismic-resistant fuel rack storage system according to the present disclosure are configured for a “free standing” fuel rack 100 as described in the Background (i.e. no provisions such as holes for use in in providing fasteners to affix the pedestals and fuel racks to the bottom of the fuel pool). Pedestals 200 preferably may be made of a corrosion resistant metal of suitable dimension and thickness to provide the strength necessary to adequately support the weight of the fuel assemblies 28 and storage tubes 120 supported by the baseplate 102. Each fuel rack 100 may include a plurality of peripheral pedestals 200 spaced apart and arranged along the peripheral edges and perimeter of the baseplate 102, and optionally one or more interior pedestals if required to provide supplemental support for the inboard fuel assemblies and tubes 120B. In one non-limiting embodiment, four peripheral pedestals 200 may be provided each of which is located proximate to one of the four corners 206 of the baseplate. Additional peripheral pedestals may of course be provided as necessary between the corner pedestals on the perimeter of the baseplate. The pedestals are preferably located as outboard as possible proximate to the peripheral edges 208 of the baseplates 102 of each fuel rack or module to give maximum rotational stability to the modules. With continuing reference to FIGS. 2-6, a seismic resistant nuclear fuel storage rack system further comprises a plurality of specially-configured embedment plates 300 fixedly coupled to the base slab 42 of the fuel pool 40 for engaging the fuel rack pedestals 200. Accordingly, the embedment plates 300 are not movable in relation to the base slab 42 or pool adjoining pool liner 60. Embedment plates 300 are arranged in a laterally spaced apart pattern around the pool base slab 42 and each is positioned to coincide with the location of at least one of the fuel rack pedestals 200. This forms a discontinuous pedestal support system in which no two embedment plates 300 are in contact with each other in certain embodiments. The pool liner 60 is interspersed and extends between the embedment plates in one embodiment. The embedment plates 300 each have a smaller lateral dimension in all directions that the than the fuel racks 100 or sections of the liner. The laterally spaced apart embedment plates 300 are each hermetically seal welded together via seal welds 140 along all of their peripheral lateral sides as shown in FIG. 5 to form a continuous hermetically-sealed liner system completely encapsulating the base slab 42 of the fuel pool 40. In one configuration, the embedment plates 300 may protrude upwards beyond a top surface (floor F) of adjacent portions of the pool liner 60 as shown in FIG. 5 to facilitate forming fillet welds around the entire perimeter of the embedment plate to the liner. Other arrangements and types of welds are possible. The hermetically seal-welded embedment plates 300 and bottom sections of liner 60 thus collectively form a pool bottom which is an impervious barrier to out-leakage of pool water W through the base slab from fuel pool 40. The embedment plates 300 include a preferably flat top wall 212 defining a top surface and have suitable thickness to support a pedestal 200 and a portion of the total dead weight of the fuel rack seated thereon. In the illustrated and preferred embedment, the embedment plates 300 are preferably fixedly attached and anchored directly to the base slab 42 of the fuel pool 40 independently of the liner 60 and without any intervening structures therebetween (best shown in FIGS. 4 and 5). There is no relative movement between the embedment plates 300 and the pool liner 60 or base slab 42. This ensures optimal anchoring and stabilization of the embedment plates 300 to the pool's base slab 42 during of a seismic event (e.g. earthquake) such that the embedment plates cannot slide or move with respect to the base slab or liner 60. This also ensures that horizontally-directed lateral forces F1 produced by a seismic event and the vertical dead weight of a completely filled fuel rack 100 are transmitted directly to the steel-reinforced (e.g. rebar) base slab 42 of the fuel pool 40 without transferring lateral or vertical forces to and adversely affecting the integrity of the liner 60. This permits the liner 60 to be thinner than the embedment plate 300 and designed for only the non-load bearing function of the pool water containment. Due to the structural nature and load bearing function of the embedment plates 300, the plates preferably have a substantially greater thickness T1 than the pool liner 60 thickness T2 (see, e.g. FIG. 5) such as for example at least twice the thickness T2. Embedment plate 300 may have a minimum thickness of 1 inch or more. Each embedment plate 300 may be received in a complementary-configured upwardly open anchorage recess 350 including a bottom 351 and vertically extending sidewalls 352. A conformal fit is preferably provided if possible between the embedment plates 300 and anchorage recess 350 such that the material of fuel pool's concrete base slab 42 on the bottom and sidewalls of the recess 350 is in intimate conformal contact with bottom and sides of the embedment plate (see, e.g. FIG. 5). This can be readily achieved if the embedment plates 300 are installed before the concrete for the base slab is poured, or if concrete grout is added in gaps around the perimeter of the plates 300 between the sides of the plate and sidewalls 352 of a slightly enlarged recess. In any of these construction scenarios, laterally and horizontally acting seismic loads or forces F1 acting on the embedment plate created by engagement between shifting pedestals 200 and sidewalls 204 of the embedment cavity 302 (described below) are laterally transferred directly to the base slab 42 via the vertical sidewalls 352 of the slab contacting the sides of embedment plate 300 without transferring these loads or forces to the thinner less structurally robust pool liner 60 which could otherwise result in damage to and compromise the leak-proof integrity of the pool liner system. Alternatively in some embodiments, if embedment plates 300 are added after the base slab 42 is poured and the perimeter concrete grouting is not added as described above, a minimal appreciable gap preferably should be provided between the sides of the embedment plates 300 and the sidewalls 352 of the recess 350 to allow formation of the perimeter seal welds 140 describe elsewhere herein between the plates and pool liner 60. In addition, one or more through anchors 400 similar to those shown in FIG. 14 and further described herein (represented in dashed lines in present FIG. 5) preferably should be added for anchoring the embedment plate 300 through the bottom 351 of the anchorage recesses 350 into the concrete base slab 42 beneath the embedment plates. Laterally acting seismic loads or forces F1 in this configuration will then be transferred from the embedment plates 300 through anchors 400 into the base slab 42 so that none of these seismic forces are transmitted to the thinner pool liner 60 to protect the integrity of the liner system. To minimize sliding engagement and impact loads between adjacent fuel racks 100 during a seismic event which may damage the racks and fuel storage tubes 120, each embedment plate 300 includes at least one engineered recessed receptacle or cavity 302 configured to capture and engage a pedestal 200 of the fuel rack 100. Each embedment cavity 302 is configured (i.e. shaped and dimensioned) to receive and abuttingly engage the terminal bottom end 204 of a pedestal 300 to restrain lateral/horizontal movement of the pedestal during a seismic event. This is best shown in FIGS. 4 and 5. Each cavity 302 is collectively defined by a flat bottom wall 306 defining a vertically upward facing horizontal bearing surface for engaging the bottom end of a pedestal 200 and plurality of preferably flat sidewalls 304 extending upwards therefrom at a right angles and defining inward facing vertical bearing surfaces for engaging the sides of a pedestal. The cavities 302 have an open top for receiving the pedestals 200 of the fuel racks 100 when they are emplaced in the fuel pool 40. Preferably each embedment plate cavity 302 may be located in the central region of the embedment plate 300 such that a portion of the top wall 212 and surface completely circumscribes and surrounds the cavity on all sides (see, e.g. FIGS. 4-5 and 7-12). This arrangement ensures that portions of the embedment plate 300 surrounding cavity 302 have adequate structural strength to withstand lateral impacts forces acting against the sidewalls 304 of the cavity in a horizontal direction due to impact by a sliding pedestal 200 during a seismic event. Each embedment plate cavity 302 has a depth D2 selected to receive and entrap or restrain a sufficient lower portion of the pedestal 200 within the confines of the cavity. A proper cavity depth D2 may be selected by weighing the competing interests of keeping the pedestal as short as possible to resist cantilevered bending moments imparted to the pedestals during a seismic event on one hand (recognizing that a lower portion of the pedestal will extend in the embedment plate cavity below the floor F of the pool bottom defined by the top surface of the pool liner 60), and maintaining a bottom flow plenum P of adequate height to induce the needed amount of natural thermal pool water circulation through the fuel rack 100 on the other hand to cool the fuel. The depth D2 must also be sufficiently deep enough so that the pedestals 200 do not “jump” out of the cavities during shaking caused the seismic event. In one non-limiting example, depth D2 of the embedment cavity 302 may preferably be about 1-3 inches, more preferably about 1-2 inches, and most preferably about 1-1.5 inches (including the listed values and those therebetween of these ranges). The embedment plate cavities 302 each further have a complementary configuration to the transverse or lateral cross sectional shape of the pedestal 200. Each cavity 302 is preferably sized minimally larger in the lateral or horizontal dimensions than the comparable width or diameter of the pedestal to minimize the amount of lateral movement permitted for the pedestals, and hence the entire fuel rack 100. The maximum transverse cross sectional dimension of lower portion of the pedestal 200 that fits within the cavity 302 may be considered to define a transverse width W2 or diameter D3 as applicable depending on the shape of the lower portion (e.g. rectilinear, polygonal, circular, etc.). The nomenclature used is not important and merely descriptive of this maximum transverse dimension. In a similar vane, depending on transverse cross sectional shape of the embedment cavity 302, the cavity may be defined as having a transverse width W3 or diameter D4. In one embodiment, cavity 302 preferably without limitation may have a maximum transverse cross sectional dimension (e.g. width W3 or diameter D4) which is no more than 5-50% larger (including or therebetween these percentages) than the maximum transverse cross sectional dimension (e.g. width W2 or diameter D3) of the pedestal 200, and more preferably no more than 10-30%. Considered another way, the physical annular clearance or gap G1 formed between the pedestal 200 and the sidewalls 304 of embedment cavity 302 (measured from the maximum transverse cross sectional dimension of the lower portion of the pedestal to the sidewall) preferably may be no more than 0.5-4 inches (including or therebetween these distances), and more preferably no more than 0.5-2 inches. Preferably, the gap G1 is less than ½ the maximum transverse cross sectional dimension (W2/D3) of pedestal 200, more preferably less than ⅓, and most preferably less than ¼ that maximum transverse cross sectional dimension of the pedestal. The maximum transverse dimension of the embedment cavity 302 preferably is as small as possible from a practical standpoint allowing enough clearance for insertion of the lower portion of each pedestal 200 into the cavity when maneuvering the fuel rack 100 via an overhead crane which is typically the method used to emplace or remove fuel racks from the fuel pool 40. In operation, pedestals 200 of the fuel rack 100 are each non-fixedly seated in a cavity 302 of an embedment plate 300 as shown in FIGS. 2A-B, 4, and 5. During a seismic event, the fuel rack 100 will tend to be moved laterally and horizontal by the event. When the frictional interactive force between the bottom end 204 of the pedestal 200 and the bottom wall 306 of the cavity surfaces is exceeded, the fuel rack and pedestals will start to slide laterally/horizontally across the cavities of the embedment plates. If movement is sufficient, the lateral sides 210 of the pedestals 200 which define a first bearing surface will abuttingly engage the sidewalls 304 of the cavity 302 which define a second bearing surface. The pedestals are thus entrapped by the mutual engagement to prevent any further lateral/horizontal movement of the fuel rack to prevent or minimize impact forces between adjacent fuel racks 100. In some embodiments where possible, the sides 210 of at least the lower portion of the pedestals 200 inserted within the embedment plate cavities 302 are configured to be parallel in orientation to the sidewalls 304 of the cavities to maximize the contact area between the colliding bearing surfaces. In some embodiments, at least all the outboard/exterior perimeter or peripheral pedestals 200 at the edges of the fuel rack baseplates 102 are preferably received in a corresponding embedment plate cavity 302 which is sufficient to restraint lateral/horizontal movement of the fuel rack 100 during a seismic event. Any inboard/interior pedestals that may be provided to support the central regions of the fuel racks may optionally be constrained from movement via engagement with embedment plate cavities, but need not necessarily be so constrained. Such inboard/interior pedestals may therefore be engaged by a conventional flat embedment plate without a cavity. Various configurations of embedment plates 300 may be provided depending on the layout of fuel racks 100 in the fuel pool 40. Each seismic resistant embedment plate 300 includes at least one embedment cavity 302 as shown in FIGS. 5, 7, and 12 for example. In a typical fuel pool, the fuel racks 100 are tightly spaced so that at least the corner regions of two or more fuel racks are located proximate to each other as shown in FIG. 3 which is a top plan view of an example fuel pool 40. The vertically lateral sides 130 and upwardly exposed baseplate protruding ledges 220 described herein between adjacent fuel racks are marked (numbered) and appear as double parallel lines to discern the outlines of each fuel rack. The lateral sides 130 of perimeter or peripheral fuel racks in the pool appear as a single line where they lie adjacent to the fuel pool sidewalls 41. At some locations in the fuel pool 40, economies and stability of installation may be achieved by providing a single larger embedment plate 300 having multiple pedestal-restraint cavities 302 for capturing two or more pedestals 200 from two or more fuel racks 100. Non-limiting examples of such embedment plates with multiple cavities are shown in FIGS. 8 and 9 to illustrate the concept. Each cavity 302 is spatially separated from another cavity on the same embedment plate 300 so that a portion of the plate top wall 212 lies between the cavities as shown. The cavities 302 are spaced apart by a suitable distance to account for the dimensions of the adjacent fuel racks 100 and their pedestal 200 locations in accordance with principles of the present disclosure. In FIG. 3, embedment plates 300 are marked by an “X” which would generally coincide of course with the location of one or more pedestals engaging the embedment plates 300 and their cavities 302. As an illustrative example, a cluster of six adjacent fuel racks 100A, 100B, 100C, 100D, 100E, and 100F have been labelled for explanation. A single-cavity embedment plate 300A (e.g. FIG. 7 or 12) is shown in each sidewall 41 corner region 44 of the fuel pool 40 which has a single embedment or restraint cavity 302 configured to receive a single corner pedestal 200 of racks 100A and 100C, for example. A double-cavity embedment plate 300B (e.g. FIGS. 4 and 8) is located along the sidewall 41 of the fuel pool at the perimeter interface or intersection between two adjacent fuel racks 100A and 100B, 100B and 100C, 100C and 100D, and 100A and 100F. A quadruple-cavity embedment plate 300C (e.g. FIG. 9) is located in the interior region of the fuel pool where the corners of four fuel racks meet, such as at the corner interface or intersection between fuel racks 100A, 100B, 100E, and 100F, and racks 100B, 100C, 100D, and 100E. It will be appreciated that the cavities in each multi-cavity embedment plate 300B or 300C do not all have to be of the same shape and will depend on the shape of the fuel rack pedestal 200 to be received in each cavity. FIG. 9, as an example without limitation, shows an embedment plate 300C having three circular cavities 302 and one rectilinear (e.g. square) cavity 302. Accordingly, numerous variations of embedment plates and embedment cavities are possible depending on the design of the fuel racks and their pedestal cross sectional shapes to be accommodated. In FIG. 3, examples of interior or inboard pedestals 200A of each fuel rack 100 are illustrated. These interior pedestals preferably engage a mating embedment plate 300D, which may include a pedestal-restraint cavity 302 or alternatively may be a completely flat conventional embedment plate without any top recesses for inserting the pedestal 200A. As described elsewhere herein, providing the exterior or outboard pedestals of the fuel racks with mating embedment plates 300 having cavities is sufficient to restrain movement of the fuel racks in all horizontal/lateral directions in the event of a seismic occurrence. In the fuel racks shown in FIG. 3, each rack has four exterior corner pedestals for example (other possible embodiments of larger fuel racks may have intermediate exterior pedestals between the corner pedestals). The embedment plates 300 are preferably formed of a suitable corrosion resistant metal of suitable strength such as without limitation stainless steel, aluminum, or another metal. The metal selected may optionally be selected to be compatible for welding to the type of metal used to construct the pool liner 60 without requiring dissimilar metal welding which facilitates installation. According to another aspect of a seismic resistant fuel pool with reference to FIGS. 4-6, the fuel racks 100 may each be configured so that their baseplates 102 protrude horizontally and laterally outwards for a distance D6 beyond the vertical lateral sides 130 of the racks, thereby creating a protruding peripheral ledge 220. Ledge 220 may circumscribe and extend completely around the entire perimeter of the fuel rack 100 to protect the lateral sides of each rack (e.g. tube sidewalls 116) from damage during a seismic event. Each pedestal may be of a predetermined fixed height, shimmed at the bottom if necessary, so that the baseplates 102 of all fuel racks or modules are essentially coplanar falling within the same horizontal plane HP (referenced in FIG. 5). During a seismic event, this positioning of baseplates substantially within the same horizontal plane (recognizing installation tolerances) and the set-back or offset distance D6 of the tube sidewalls 116 from baseplate peripheral edges 208 advantageously protects the cells 110 from damage ensuring that any contact between adjacent sliding fuel racks occurs between the peripheral edges of the racks alone. Typical offset distances D6 used may be for example without limitation 1-3 inches. Larger or smaller offset distances may be used in other embodiments. Alternatively, an adjustable pedestal configuration may be used to avoid the need for shims. Such adjustable pedestals typically equipped with a two-piece threaded leg or pedestal for making vertically height adjustments are well known in the art. FIGS. 2A and 2B show an example of an adjustable pedestal design. These adjustable height pedestals 200 are connected to the bottom surface of the baseplate 102. In one embodiment, for example without limitation, the adjustment means may be accomplished via a threaded pedestal assembly. The adjustable height pedestals 200 ensure that a space exists between the base slab 42 of the fuel pool 40 and the bottom surface of the base plate 102, thereby creating an inlet plenum P for water to flow upwards through the flow holes 115 and cells 110 as describe elsewhere herein. The adjustable height pedestals 200 are spaced to provide uniform support of the base plate 102 and thus the fuel rack 100. Each such pedestal 200 is preferably individually adjustable to level and support the fuel rack on a non-uniform spent fuel pool base slab 42 surface, thereby avoiding the need for shims to ensure that the baseplates 102 of all fuel racks 100 are substantially coplanar. In one example of many possible configurations, the pedestals 200 may each comprise a block-shaped rectilinear upper mounting portion 104 affixed rigidly to the bottom surface of the fuel rack baseplate 102 and an adjustable lower base portion 105 threadably coupled to the mounting portion and moveable vertically with respect to thereto. The base portion 105 may be circular cylindrical in one embodiment as shown to provide a stable base pad for engaging bottom wall 306 of embedment plate cavity 302; however, other suitable shapes may be used. The pedestals mounting portion 104 may be bolted to the baseplate 102 in some embodiments. Of course, in other embodiments, the mounting portions 104 can be attached to baseplate 102 by other means, including without limitation welding or threaded attachment as just two examples. In some embodiments, an additional measure may be provide according to another aspect of the invention that further minimizes or prevents the likelihood of damage between laterally sliding or moving fuel racks during a seismic event. The fuel racks 100 may be arranged on the base slab 42 of the fuel pool 40 such that the proximate facing and mating peripheral edges 208 between baseplates 102 (e.g. horizontal protruding ledges 220) of adjacent fuel racks 100 are placed in abutting mutual edge contact or engagement upon emplacement in the fuel pool under normal operating conditions prior to the occurrence of a seismic event. Such an edge contact arrangement with abutment joints 150 formed between mating baseplate peripheral edges 208 is shown for example in FIGS. 3, 4, and 6. FIG. 4 shows best shows first and second fuel racks 100A and 100B with baseplate abutment joint 150 therebetween. The foregoing edge contact arrangement between baseplates 102 of adjoining fuel racks 100 advantageously precludes any substantial degree of movement between adjacent fuel racks into each other. This eliminates initial impact forces between adjoining baseplates caused lateral shifting of the fuel racks due to seismic activity because the baseplates are pre-engaged. Because of the pre-seismic event edge contact arrangement, the fuel racks 100 so coupled would laterally move or slide in unison together under seismic activity a distance to the point where the entrapped fuel rack pedestals 200 engage the embedment plate cavity walls 204. Advantageously, there is no differential movement of one fuel rack 100 with respect to adjoining fuel racks in the entire array of racks in the fuel pool 40 thereby eliminating any substantial damage to the racks. It will be appreciated that because of metal fabrication tolerances, complete conformal contact although desirable may not be possible along the entire horizontal peripheral edge interface length between two abutting baseplates 102. A minimal gap of for example without limitation no more than ¼ inch is reasonably obtainable at those interspersed locations if any between the adjoining fuel rack baseplates 102 where complete abutting conformal contact might not be fully achieved due to metal fabrication limitations. Preferably, however, abutting conformal contact is achieved for a majority of the length of each abutment joint 150 between mutually engaged pairs of fuel rack baseplate peripheral edges 208 (whether the conformal contact is measured contiguously or dis-contiguously at intermediate lengths along the mating baseplates separated by minor non-conformal contact areas). It bears noting that conformal contact between adjoining fuel rack baseplates may be used in some embodiments as an alternative to the embedment plate cavities 302 describe herein, or preferably in other embodiments in conjunction with the cavities to provide dual protection against fuel rack damage during a seismic event. A process or method for seismic-resistant storage of nuclear fuel in a fuel pool will now be briefly described based on seismic resistant fuel storage system described herein. In one embodiment, the method may comprise transporting and staging a plurality of fuels racks 100 proximate to the fuel pool 40 in a nuclear facility for loading the racks into the pool. The first fuel rack 100 is lifted via a crane (not shown) or other suitable piece of lifting equipment and maneuvered over the fuel pool 40. The first fuel rack 100 is oriented so that the perimeter pedestals 200 are each vertically aligned with a corresponding embedment plate 300 on the base slab 42 of the pool. The next step is lowering the first fuel rack into the water-filled fuel pool and insertably engaging each of the pedestals of the first fuel rack with corresponding upwardly open embedment receptacles or cavities 302 formed in a plurality of embedment plates 300 already fixedly coupled to the base slab of the fuel pool. The bottom ends of the pedestals 200 are seated on the recessed bottom wall 306 of the cavities and the sidewalls 304 trap the pedestals therein. Lateral movement of the pedestals 200 and hence fuel racks along the base slab 42 during a seismic event is restricted by engagement between the pedestals and the sidewalls of the embedment cavities of the embedment plates. After the first fuel rack is positioned in the fuel pool 40, the method may continue with lowering a second fuel rack 100 into the fuel pool, insertably engaging each of the pedestals 200 of the second fuel rack with corresponding upwardly open receptacles or cavities 302 formed in the plurality of embedment plates 300 fixedly coupled to the base slab 42 of the fuel pool, and abuttingly engaging a peripheral edge 208 of the baseplate 102 of the first fuel rack with an adjoining peripheral edge of the baseplate of the second fuel rack. The baseplates 102 of the first and second fuel racks are substantially coplanar as already described herein to ensure mutual engagement. In some situations, at least one pedestal 200 of the second fuel rack 100 and at least one pedestal of the first rack may be engaged with separate receptacles formed in a single shared embedment plate, such as without limitation embedment plates 300B or 300C shown in FIGS. 8 and 9, respectively. Numerous variations in the foregoing method are possible. FIG. 14 shows an alternative embodiment of an embedment plate system in which embedment plates 300 are anchored to the base slab 42 of the fuel pool 40 through the pool liner 60 plate interposed therebetween. The bottom surface of the embedment plate 300 is seated directly on the top surface of the pool liner 60. One or more through metal anchors 400 are provided which vertically extend completely through the embedment plates and liner 60 into base slab 42. In one embodiment, the anchors 400 may be threaded masonry fasteners such as lag bolts threadably secured at their bottom ends into the base slab 42 of fuel pool 40 and having an exposed enlarged head at the opposite end configured for engaging a tightening tool such as a wrench. Use of other types of anchors is of course possible. This embodiment similarly prevents any relative movement between the pool liner 60 and the embedment plates 300. Although direct embedment of the embedment plates 300 in the pool base slab 42 shown in FIG. 5 is preferred when possible, this embodiment is useful for retrofit installations where an embedment plate system according to the present disclosure is added to an existing fuel pool 40 having a liner 60. This eliminates the need to cutout the existing pool liner 60 at the embedment plate locations. The embedment plates 300 may be hermetically sealed welded to the liner 60 completely around their perimeters using fillet welds 140 in a similar manner to that already described herein. FIG. 15 shows a second alternative embodiment of an embedment plate system in which embedment plates 300 are anchored directly to the base slab 42 of the fuel pool 40. In contrast to the embodiment of FIG. 14, in this embodiment no portion of the pool liner 60 plate is interposed between the embedment plate 300 and slab. The bottom surface of the embedment plate 300 is seated directly on the top surface of the base slab 42. One or more through metal anchors 400 are provided which vertically extend completely through the embedment plate into base slab 42. This embodiment similarly prevents any relative movement between the pool liner 60 and the embedment plates 300. The embedment plates 300 may be hermetically sealed welded to the liner 60 completely around their perimeters using fillet welds 140 in a similar manner to that already described herein. While the foregoing description and drawings represent exemplary embodiments of the present disclosure, it will be understood that various additions, modifications and substitutions may be made therein without departing from the spirit and scope and range of equivalents of the accompanying claims. In particular, it will be clear to those skilled in the art that the present invention may be embodied in other forms, structures, arrangements, proportions, sizes, and with other elements, materials, and components, without departing from the spirit or essential characteristics thereof. In addition, numerous variations in the methods/processes described herein may be made within the scope of the present disclosure. One skilled in the art will further appreciate that the embodiments may be used with many modifications of structure, arrangement, proportions, sizes, materials, and components and otherwise, used in the practice of the disclosure, which are particularly adapted to specific environments and operative requirements without departing from the principles described herein. The presently disclosed embodiments are therefore to be considered in all respects as illustrative and not restrictive. The appended claims should be construed broadly, to include other variants and embodiments of the disclosure, which may be made by those skilled in the art without departing from the scope and range of equivalents.
	abstract	A particle beam irradiation chamber in which a passage having a first opening part at a side of an inner wall and a second opening part at a side of an outer wall is provided and which has an isocenter inside the chamber, wherein a first line segment which connects the center of the first opening part and the center of the second opening part passes inside the passage, an angle, which is formed by a second segment, which connects the center of the first opening part and the isocenter and the first line segment, is smaller than 180 degrees, and a width of the passage is narrower than a width of opening of the first opening part.
	abstract	A Potter-Bucky device for a radiation image recording apparatus in which an image recording medium is exposed to radiation which has passed through an object in order to record a radiation image of the object on the recording medium includes a grid which is movably supported between the object and the recording medium and is reciprocated parallel to the recording medium. A counter-weight is connected to the grid and is reciprocated in synchronization with the grid but in an opposite direction. Among other benefits and features, the disclosure reduces or eliminates vibrations produced by the reciprocating grid.
055263852	claims	1. In a nuclear reactor having a core and a pressure vessel with a wall and an interior, a safety device protecting against overpressure failure of the pressure vessel upon insufficient cooling of the core, comprising: a pressure pipe passing pressure-tightly through the wall and extending into the interior of the pressure vessel, said pressure pipe having at least one pressure compensation opening formed therein in the interior of the pressure vessel and having a fusible sealing body sealing said at least one pressure compensation opening; said fusible sealing body being formed of a melting solder melting at a limit temperature and unblocking said at least one pressure compensation opening, but keeping said at least one pressure compensation opening sealed during normal operation. said pressure pipe has a perforated pipe head being sealed by said fusible sealing body and disposed underneath the reactor core; and said pressure pipe is laid downwards in the interior of the bottom hemisphere and next to the bottom hemisphere in the annular space upwards up to a pressure-tight penetration of the wall region between the main coolant branches. 2. The safety device according to claim 1, wherein said pressure pipe is a blow-off pipe and said at least one pressure compensation opening is a pressure relief opening. 3. The safety device according to claim 1, including a blow-off valve disposed outside the vessel, said pressure pipe being a pressure control pipe triggering said blow-off valve for reducing a system pressure. 4. The safety device according to claim 1, wherein the pressure vessel has a cover branch through which said pressure pipe is sealingly guided in a suspended configuration, said pressure pipe having a perforated pipe head being sealed by said fusible sealing body and extended into the interior of the pressure vessel. 5. The safety device according to claim 1, wherein the pressure vessel has a bottom hemisphere with an interior, and main coolant branches defining a region of the pressure vessel wall therebetween, and the pressure vessel contains a core vessel being spaced apart from the pressure vessel wall by an annular space; 6. The safety device according to claim 5, wherein the pressure vessel contains a lower core structure inside which said perforated pipe head is disposed. 7. The safety device according to claim 1, wherein said pressure pipe includes a perforated pipe head having an end surface, having a pipe shell wall with a mutually adjacent plurality of said at least one pressure compensation opening formed therein, and having a pipe plug sealing said end surface, said fusible sealing body being a fusible sleeve soldered to said pipe shell wall for sealing said pressure compensation openings. 8. The safety device according to claim 7, wherein said at least one pressure compensation opening is a plurality of mutually adjacent rings of pressure compensation openings being coaxial with said pressure pipe. 9. The safety device according to claim 7, wherein said pipe plug has a conical profile with a rounded tip. 10. The safety device according to claim 1, wherein said pressure pipe leads outside the pressure vessel as a blow-off line into a blow-off vessel. 11. The safety device according to claim 3, wherein said pressure control pipe has a sealed end extending into the interior of the pressure vessel and has a shell wall having said at least one pressure compensation opening formed therein, and said fusible sealing body has a spherical metal body embedded therein in said at least one pressure compensation opening. 12. The safety device according to claim 11, including a cover sealing said end of said pressure control pipe. 13. The safety device according to claim 11, wherein said at least one pressure compensation opening is an oblique opening having an opening axis being oriented obliquely inward, for dropping said spherical body into the interior of said pressure control pipe in the event of fusion. 14. The safety device according to claim 1, wherein the limit temperature is within a range of from 600.degree. C. to 700.degree. C. 15. The safety device according to claim 1, wherein the melting solder is resistant to radiation. 16. The safety device according to claim 1, wherein the melting solder has a high silver content. 17. The safety device according to claim 1, wherein the melting solder has a silver content of approximately 50%.
	description	This application is a continuation-in-part application of and claims priority to U.S. Non-provisional patent application Ser. No. 12/244,017, filed on Oct. 2, 2008, entitled “Plasma Uniformity Control Using Biased Array.” The entire specification of U.S. Non-provisional patent application Ser. No. 12/244,017, is incorporated herein by reference. The present disclosure is related to semiconductor manufacturing, particularly to semiconductor manufacturing using plasma. Ions are commonly implanted into a substrate in ion implantation processes to produce semiconductor devices. These ion implantations may be achieved in a number of different ways. For example, a beam-line ion implantation system may be used to perform the ion implantation process. In the beam-line ion implantation system, an ion source is used to generate ions, which are manipulated in a beam-like state, and then directed toward the wafer. As the ions strike the wafer, they dope a particular region of the wafer. The configuration of doped regions defines their functionality, and through the use of conductive interconnects, these wafers can be transformed into complex circuits. In another example, a plasma containing ions may be generated near the substrate. A voltage is then applied to the substrate to attract ions toward the substrate. This technique is known as plasma doping (“PLAD”) or plasma immersion ion implantation (“PIII”) process. FIG. 1 shows an exemplary plasma doping system 100. The plasma doping system 100 includes a process chamber 102 defining an enclosed volume 103. Within the volume 103 of the process chamber 102, a platen 134 and a workpiece 138, which is supported by the platen 134, may be positioned. A gas source 104 provides a dopant gas to the interior volume 103 of the process chamber 102 through the mass flow controller 106. A gas baffle 170 is positioned in the process chamber 102 to deflect the flow of gas from the gas source 104. The process chamber 102 may also have a chamber top 118 having a dielectric section extending in a generally horizontal direction and another dielectric section extending in a generally vertical direction. The plasma doping system may further include a plasma source 101 configured to generate a plasma 140 within the process chamber 102. The source 101 may include a RF power source 150 to supply RF power to either one or both of the planar antenna 126 and the helical antenna 146 to generate the plasma 140. The RF source 150 may be coupled to the antennas 126, 146 by an impedance matching network 152 that matches the output impedance of the RF source 150 to the impedance of the RF antennas 126, 146 in order to maximize the power transferred from the RF source 150 to the RF antennas 126, 146. The plasma doping system 100 also may include a bias power supply 148 electrically coupled to the platen 134. The bias power supply 148 may provide a continuous or a pulsed platen signal having pulse ON and OFF time periods to bias the workpiece 138. In the process, the ions may be accelerated toward the workpiece 138. The bias power supply 148 may be a DC or an RF power supply. In operation, the gas source 104 supplies a dopant gas containing a desired dopant species to the chamber 102. To generate the plasma 140, the RF source 150 resonates RF currents in at least one of the RF antennas 126, 146 to produce an oscillating magnetic field. The oscillating magnetic field induces RF currents into the process chamber 102. The RF currents in the process chamber 102 excite and ionize the primary dopant gas to generate the plasma 140. The bias power supply 148 provides a pulsed platen signal to bias the platen 134 and, hence, the workpiece 138 to accelerate ions from the plasma 140 toward the workpiece 138. The frequency of the pulsed platen signal and/or the duty cycle of the pulses may be selected to provide a desired dose rate. The above technique is known to provide high implant throughput. However, the uniformity of the dose is difficult to control. In the beam-line ion implantation system, components such mass analyzer magnets, deceleration electrodes and other beam-line components may be used to manipulate ions into a uniform ion beam, and the workpiece may be uniformly implanted with ions in the uniform ion beam. Such components, however, are not available with a plasma doping system. To uniformly implant the workpiece in the plasma doping system, the plasma generated near the substrate should be uniform, as PLAD implant uniformity is closely related to plasma uniformity. In a typical plasma based system, the generated plasma is typically non-uniform; the plasma density is typically higher in the center of the plasma than near the chamber walls, as shown in FIG. 4. As a result, implant profile on the workpiece shows a similar non-uniform profile—higher implant dose in the middle, and lower dose in the edges of the workpiece. Typically, RF power, gas flow and distribution, magnetic confinements, etc. may be adjusted to improve the plasma uniformity. However, such techniques may mitigate the plasma non-uniformity, but cannot change the generic non-uniform density profile shown in FIG. 4. As such, systems and methods to improve the uniformity of the plasma in a plasma based system are needed. A technique for processing a workpiece is disclosed. In accordance with one exemplary embodiment, the technique may be realized as a method for processing a substrate. The method may comprise: providing the workpiece in the chamber; providing a plurality of electrodes between a wall of the chamber and the workpiece; generating a plasma containing ions between the plurality of electrodes and the workpiece, ion density in an inner portion of the plasma being greater than the ion density in an outer portion of the plasma portion, the outer portion being between the inner portion and the wall of the chamber; and providing a bias voltage to the plurality of electrodes and dispersing at least a portion of the ions in the inner portion until the ion density in the inner portion is substantially equal to the ion density in the periphery plasma portion. In accordance with other aspects of this particular exemplary embodiment, the method may further comprise attracting ions to the workpiece from the inner and outer portions of the plasma. In accordance with further aspects of this particular exemplary embodiment, the providing the bias voltage to the plurality of electrodes may comprise independently biasing the plurality of electrodes. In accordance with additional aspects of this particular exemplary embodiment, the providing the bias voltage to the plurality of electrodes may comprise positively biasing the plurality of electrodes. In accordance with further aspects of this particular exemplary embodiment, the providing the bias voltage to the plurality of electrodes may comprise negatively biasing the plurality of electrodes. In accordance with other aspects of this particular exemplary embodiment, the method may further comprise: coupling the plurality of electrodes to one or more power supplies, where the one or more power supplies may provide a pulsed bias voltage having a duty cycle. In accordance with further aspects of this particular exemplary embodiment, the method may further comprise: adjusting the duty cycle of the pulsed bias voltage provided to the plurality of electrodes. In accordance with additional aspects of this particular exemplary embodiment, the method may further comprise providing a magnetic field in same direction as an electric field created by the biasing the plurality of electrodes. In accordance with other aspects of this particular exemplary embodiment, the plurality of electrodes may comprise a first electrode disposed near the inner portion of the plasma and a second electrode disposed near the outer portion of the plasma, and where the first electrode may be applied with a first bias voltage and the second electrode is applied with a second bias voltage. In accordance with further aspects of this particular exemplary embodiment, the first bias voltages may be more positive than the second bias voltage. In accordance with additional aspects of this particular exemplary embodiment, the first bias voltage may be less positive than the second bias voltage. In accordance with another exemplary embodiment, the technique may be realized as a method for processing a workpiece. The method may comprise: providing the workpiece in a plasma chamber; providing first and second electrodes between a wall of the plasma chamber and the workpiece; generating a plasma containing ions between the plurality of electrodes and the workpiece, the plasma may comprise a first plasma portion disposed near the first electrodes and a second plasma portion disposed near the second electrode, where the first plasma portion may have a greater ion density than the second plasma portion; providing a bias voltage to the first and second electrodes and dispersing at least a portion of the ions in the first plasma portion until the ion density in the first plasma portion is substantially equal to the ion density in the second plasma portion; and biasing the workpiece and attracting ions to the workpiece from the first and second portions of the plasma having substantially equal ion density. In accordance with other aspects of this particular exemplary embodiment, the first plasma portion may be disposed near a middle of the plasma and the second plasma portion may be disposed near a periphery of the plasma next to the first plasma portion. In accordance with further aspects of this particular exemplary embodiment, the second plasma portion may be disposed near a middle of the plasma and the first plasma portion is disposed near a periphery of the plasma next to the second plasma portion. In accordance with additional aspects of this particular exemplary embodiment, the providing the bias voltage to the first and second electrodes comprises independently biasing at least one of the plurality of electrodes. In accordance with further aspects of this particular exemplary embodiment, the providing the bias voltage to the first and second electrodes may comprise positively biasing at least one of the plurality of electrodes. In accordance with additional aspects of this particular exemplary embodiment, the providing the bias voltage to the first and second electrodes may comprise negatively biasing at least one of the plurality of electrodes. In accordance with further aspects of this particular exemplary embodiment, the providing the bias voltage to the first and second electrodes may comprise biasing the first electrodes with a first bias voltage and the second electrodes with a second bias voltage. In accordance with other aspects of this particular exemplary embodiment, the first bias voltage may be more positive than the second bias voltage. In accordance with additional aspects of this particular exemplary embodiment, the first bias voltage is less positive than the second bias voltage. In accordance with another exemplary embodiment, the technique may be realized as a method for processing a workpiece. The method may comprise: providing the workpiece in a plasma chamber; providing first and second electrodes between a top of the plasma chamber and the workpiece; generating a plasma between the plurality of electrodes and the workpiece, where the plasma may comprise first and second portions having different plasma density; independently applying bias voltage first and second electrodes until the difference in the plasma density of the first and second portion of the plasma is reduced. In accordance with additional aspects of this particular exemplary embodiment, more positive bias voltage is applied to one of the first and second electrodes near first and second portions the bias voltage applied to one of the first and second electrodes near one of the first and second portions of the plasma with greater plasma density to improve uniformity of the plasma density between the first and second portions. The present disclosure will now be described in more detail with reference to exemplary embodiments thereof as shown in the accompanying drawings. While the present disclosure is described below with reference to exemplary embodiments, it should be understood that the present disclosure is not limited thereto. Those of ordinary skill in the art having access to the teachings herein will recognize additional implementations, modifications, and embodiments, as well as other fields of use, which are within the scope of the present disclosure as described herein, and with respect to which the present disclosure may be of significant utility. Herein, several embodiments of an apparatus and method for achieving uniform plasma density are disclosed with reference to accompanying drawings. The detailed disclosure contained herein is intended for illustration, for better understanding the disclosure, and not a limitation thereto. For example, the disclosure may be made with reference to a plasma doping or a plasma immersion ion implantation system. However, the present disclosure may be equally applicable to other plasma based systems including plasma based etching and deposition systems. As described above, a plasma doping system is used to create a plasma in close proximity to the workpiece. The workpiece may be then biased to a certain electrical potential. However, the plasma density or the ion concentration within the generated plasma may be non-uniform. Typically, the concentration of ions is the highest near the center and lower near the chamber wall, as shown in FIG. 4. In a plasma based system that is radially symmetric, the ion diffusion pattern may also be radially symmetric along the horizontal direction. As such, the plasma generated in a radially symmetric plasma based system may have approximately concentric density profile. Ion concentration at a point removed from the center of the plasma may be approximately the same as another point equidistanced from the center. Such a characteristic in symmetric plasma based system may result in a dome shaped plasma density profile. Plasma is a quasi-neutral state where positively and negative charged particles show collective behaviors. Charged particles in the plasma are responsive to both electrical and magnetic fields. By using these fields to manipulate the local distribution of the charged particles within the plasma, the implant uniformity can be improved. FIG. 2 represents a first embodiment of the apparatus. In this figure, many of the components in the plasma doping system of FIG. 1 have not been included in FIG. 2 for purpose of clarity and simplicity. However, it should be understood that the components shown in FIG. 1 may also be in the plasma doping system. Referring to both FIGS. 1 and 2, the plasma 140 may be positioned between the workpiece 138 and the baffle 170. The baffle 170 can be a stationary baffle 170 and/or adjustable baffle 170. The adjustable baffle 170 can move in a vertical direction (up and down) relative to the wafer or the chamber ceiling. This movement may occur prior to and/or during wafer processing. Periodic pulses of bias voltage at the workpiece may be applied to accelerate ions toward the workpiece. However, as seen in FIGS. 1 and 2, there are no mechanisms to confine the plasma or control its uniformity. In one embodiment, a set of electrical conductors 200 is preferably located on the underside of the baffle 170 such that the conductors 200 may be positioned above the plasma. These conductors may preferably be electrically insulated from one another and from the baffle. For example, an insulating material (not shown) may separate the conductors 200 from one another and from the baffle 170. In another embodiment, the electrical conductors 200 may be disposed around the plasma (e.g. the side of the plasma). Yet in another embodiment, a set of electrical conductors 200 may be disposed above the plasma and another set of electrical conductors 200 may be disposed around the plasma. In the present embodiment, the electrical conductors 200 may be pins 200. However, those or ordinary skill in the art will recognize that the electrical conductors 200 may be other types of conductor 200. In addition, the electrical conductors may have diameters of other values. In the present embodiment, the pins 200 may preferably be arranged in a two-dimensional array, as shown in FIG. 3. In a plasma doping system, the plasma may have a cylindrical shaped volume, having a diameter of about 50 cm and a height of about 10-20 cm. Thus, if the two-dimensional array is to extend over the plasma region, and the distance between adjacent pins is about 1.0 cm, then the array may contain about 304 pins. However, those of ordinary skill in the art will recognize that the number of the pins in the array may be more or less. For example, if the array of the pins covers the 300 mm wafer region with the distance between adjacent pins of 2.54 cm, then the array would contain only about 110 pins. Additionally, the electrical conductors can be various shapes including rectangular, square, round or other shapes. The most preferred shapes include (1) a flat cylindrical shape (0.1-1.0 cm in diameter) and (2) a pointed-tip cylindrical shape (0.05 cm or less in diameter for pointed tip, 0.1-1.0 cm in diameter for the pin body). For the latter case, the total angle of the pointed-tip may be less than 90 degrees. Each of these pins may be independently controlled. For example, each pin may be biased to a voltage independent of other pins. Furthermore, each pin may be biased either positively or negatively. Finally, these bias voltages may be constant, or pulsed. In addition, the bias voltages may vary between conductors. Furthermore, the magnitude of the bias voltage at a particular conductor may vary over time. Thus, the two-dimensional array may be used to create any desired electrical field, and that field can be static or may vary. By creating an electrical field potential above the plasma, the ion concentration within the plasma can be altered. For example, the use of a positive bias voltage will draw the electrons within the plasma toward those positively biased pins. The magnitude of that bias voltage may determine the size of the affected field. By drawing the electrons toward the upper portion of the plasma, the positive ions may disperse due to space charge effects. Such a dispersion of the positive ions may change the positive ion distribution within the plasma. Therefore, the dispersion may locally lower the concentration of implanted ions when the substrate bias voltage is applied. Negative bias voltages on the pins may have different effect. The negative voltage may repel the electrons and thereby cause the plasma to be locally compressed. This compression increases the local concentration of positive ions near the workpiece. FIG. 4 shows a typical graph of the ion concentration as compared to the distance from the center of the system along one axis. Although this shows ion concentration versus distance in one dimension, similarly shaped graphs exist in all dimensions. Thus, by applying positive bias voltage near the center of the system, the ion concentration can be lowered, thereby improving uniformity. Additionally, applying negative bias voltage near the outer portions of the plasma compresses the plasma, and therefore effectively increases its concentration, further improving uniformity. Furthermore, electrical conductors 230 may be placed vertically around the sides of the plasma, as shown in FIG. 7. Side baffles 235 are positioned about the sides of the plasma. A set of electrical conductors 230 is preferably located on the side of the side baffle 235. These conductors 230 are electrically insulated from one another and from the side baffle 235. Such a configuration may serve to better confine the plasma. As noted above, the bias voltage applied to one or more pins may be constant (DC) or intermittent, such as pulsed. Additionally, the pulsed bias voltage may be positive or negative. Alternatively, the one or more pins may be floated or grounded, as desired. Applying the pulsed bias voltages to the pins has certain advantages over DC bias. Since the plasma electrons are sensitive to the positive bias voltages, DC bias may cause too much perturbation to the plasmas, such as causing plasma instability or redistribution of the plasma in some applications. In such cases, pulsed bias with small duty cycle (50% or less) can minimize the plasma perturbation while providing controllability of the plasma uniformity. The duration of each pulse may preferably be between microseconds and milliseconds in the order of magnitude. As noted above, the bias voltage applied to one or more electrical conductors 200 may be positive or negative. Alternatively, one or more electrical conductors may be grounded or floated. If biased, the bias voltage may be a constant voltage, or varying. In certain embodiments, the bias voltage is a periodic waveform having a duty cycle. This duty cycle can be between microseconds and milliseconds in order of magnitude. Furthermore, the duty cycle can vary, such that the duration of the pulses can change based on the plasma density and the desired density. Thus, bias voltage waveform may change in duration, frequency, magnitude, duty cycle or polarity over time. Although each pin maybe independently controlled, groups of pins can be grouped together in one or more groups, and different groups may be controlled independently of other groups. For example, pins removed from the center by the same distance may be controlled together if the density profile of the non-uniform plasma is radially symmetric. However, if the plasma density is asymmetric, each pin or each group of pins may be controlled independently. While the disclosure describes an array of pins as shown in FIG. 3, other embodiments are possible and within the scope of the disclosure. For example, another embodiment of the electrical biased elements is shown in FIG. 5. In this Figure, it is assumed that the plasma is symmetrical and therefore, the ion concentrations at a same distance from the center are all identical. Each concentric ring represents a set of electrically conductive elements 210, which can be biased independently of the adjacent rings. Thus, the same effect is desired, and therefore the same bias voltage can be applied. Other embodiments are also within the scope of the disclosure. In addition to electrical fields, magnetic fields can be added to further improve the plasma uniformity and therefore implant uniformity. In the above embodiment, there was no magnetic field, thus charged particles are free to move in all directions. By introducing a magnetic field parallel to the electrical field, charged particles will be limited in their freedom of motion. Referring to FIG. 6, a magnetic field is added to the apparatus shown in FIG. 2 and is created parallel to the electrical field. In this embodiment, charged particles are more restricted in their movement, in that the charged particles are confined along the magnetic field lines. Thus the effect of the bias voltages described above is more contained. In other words, each electrically conductive element controls the ion concentration of the plasma in the volume located directly below the element. Thus, the bias voltages applied at one element do not affect the ion concentrations in other areas of the plasma. The magnetic field can be created in a variety of ways, as are known by those skilled in the art. Apparatus and method for improving plasma uniformity in a plasma based system are disclosed. Although the present disclosure has been described herein in the context of particular systems and particular implementations in particular environments for a particular purpose, the present disclosure is not limited thereto. Those of ordinary skill in the art will recognize that its usefulness is not limited thereto and that the present disclosure can be beneficially implemented in any number of environments for any number of purposes. For example, one or more electrical conductors near a portion of the plasma with greater ion or plasma density may be applied with a first bias voltage. The portion of the plasma with greater density may not necessarily be positioned in the inner or middle portion of the plasma. In some embodiments, the portion with greater density may be located at the outer or periphery portion of the plasma, the portion located between the inner portion and the chamber wall. This first bias voltage, which applied to one or more electrical conductors near the portion of the plasma with greater density, may be a positive bias voltage. With this bias voltage, the positively charged ions near the electrical conductors may be dispersed away from the portion of the plasma with greater ion or plasma density. In the process, there may be a local decrease in plasma density in the portion. As a result, plasma with increased uniformity may be achieved. Alternatively, one or more electrical conductors near another portion of the plasma with less ion or plasma density may be applied with a second bias voltage. The second bias voltage may be less positive than what the first bias voltage would have been. In one example, the second bias voltage may be a negative bias voltage. In another example, the second bias voltage may be a positive bias voltage, but less than what the first bias voltage would have been had the first bias voltage been applied to other conductors. With the application of negative bias voltage, the plasma may be compressed. In the process, the density of the less dense portion of the plasma may increase, and the uniformity of the plasma may be enhanced. In another embodiment, one or more electrical conductors near a portion of the plasma with greater ion or plasma density may be applied with a first bias voltage. Meanwhile, one or more electrical conductors near another portion of the plasma with less ion or plasma density may be applied with a second bias voltage, less positive than the first bias voltage. By independently applying one or more bias voltages to one or more electrical conductors, plasma with greater uniform density may be achieved. Referring to FIG. 8A, there is shown another exemplary plasma processing apparatus 800 according to another embodiment of the present disclosure. In FIG. 8B, there is shown a plurality of electrical conductors 810 that may be included in the plasma processing apparatus 800 shown in FIG. 8A. Much like the earlier embodiment shown in FIG. 2, the plasma processing apparatus 800 contains a plasma source 101 for generating the plasma 140 in the chamber 102. In addition, the plasma processing apparatus 800 comprises a platen 134 for supporting a substrate 138. Those skilled in the art will recognize that several components in the plasma processing apparatus shown in FIGS. 1 and 2 are also contained in the plasma processing apparatus 800 of FIG. 8A. Such components in FIG. 8A should be understood in relation to the same components in the plasma processing apparatus shown in FIGS. 1 and 2. As illustrated in FIG. 8A the plasma processing apparatus 800 may comprise a plurality of electrical conductors 810 disposed at various positions within the plasma chamber 102. The electrical conductors 810 may be a part of or attached to the baffle 170, as shown in FIG. 2. In other embodiments, the electrical conductors 810 may be spaced apart from the baffle. For example, the electrical conductors 810 may be located below the baffle and closer to the plasma 140. Yet in another embodiment, the plasma processing system 800 may be without the baffle. In this embodiment, only the electrical conductors 810 are illustrated for clarity and simplicity. The plurality of electrical conductors 810 may comprise at least one first or inner electrical conductor 810a disposed near the center of the chamber 102 or the inner portion of the plasma 140. The plurality of electrical conductors 810 may also comprise one or more second or outer electrical conductor 810b disposed near the outer portion of the plasma 140, the portion between the inner portion of the plasma and the chamber wall. In other embodiments, there may be one or more intermediate electrical conductors disposed between the inner and outer electrical conductors 810a and 810b. If included, the intermediate electrical conductors may comprise one or more third and fourth electrical conductors 810c and 810d positioned between the inner and outer conductors 810a and 810b. Much like the prior embodiments, the first and second conductors 810a and 810b may be electrically isolated from each other. Moreover, each of the first conductors 810a, if two or more are provided, may be electrically isolated from each other. If two or more are provided, each of the second conductors 810b may also be electrically isolated from each other. Likewise, each of the third and fourth conductors 810c and 810d may also be electrically isolated from each other and from each of the first and second conductors 810a and 810b. Further, if two or more are included, each of the third conductors 810c and each of the fourth conductors 810d may be electrically isolated from each other. Each of the first and second conductors 810a and 810b may be independently biased. If included, the third and fourth conductors 810c and 810d may also be independently biased. The bias voltage applied to the conductors 810a-810d may be a continuous or pulsed bias. Moreover, the bias voltage applied may be positive or negative bias voltage. In some embodiments, at least one of the first and second 810a and 810b, and the third and fourth conductors 810c and 810d if included, may remain floating or grounded. In operation, a plasma source 101 in the plasma doping system 800 may generate a plasma 140 between the workpiece 138 and the electrical conductors 810. However, the present disclosure does not preclude generating a plasma above the first and second conductors 810a and 810b. The plasma 140 generated in the chamber 102 may have a density profile similar to the profile shown in FIG. 8C. For example, the plasma 140 may have higher ion density near its inner portion and less ion density near its outer portion. To improve the uniformity, a bias voltage may be provided to one of more of the first and second electrical conductors 810a and 810b. In the present embodiment, a first bias voltage may be applied to the first conductor 810a. This first bias voltage may be a positive bias voltage. If applied with a positive bias voltage, the first electrical conductor 810a may locally disperse the positively charged ions away from the inner portion of the plasma 140. As a result, the difference in the plasma density between the inner portion of the plasma 140 and the outer portion of the plasma 140 may decrease, and the uniformity of the plasma 140 may improve. Alternatively, the second conductor 810b may also be applied with the bias voltage. In the present embodiment, the bias voltage applied to the second electrical conductor 810b may be a second, less positive voltage. For example, the second electrical conductor 810b may be biased with negative bias voltage. By applying a negative bias voltage to the second conductor 810b, the outer portion of the plasma 140 with less plasma density may be compressed. As such, further improvement to the uniformity of the plasma 140 may be achieved. In another embodiment, both the first and second conductors 810a and 810b may be independently biased with the first and second bias voltages. In this embodiment, the dispersion of ions from the portion of the plasma with greater density and the compression of the portion of the plasma with less density may occur. When both bias voltages are applied, they may be applied simultaneously or at different times. If included, the third and fourth electrodes 810c and 810d may also be biased. If biased, the third electrode 810c may be biased with less positive bias than the first electrode 810a, but more positive bias than the second electrode 810b. Meanwhile, the fourth electrode 810d, if biased, may be biased with less positive bias than the third electrode 810c, but more positive than the bias voltage applied to the second electrode 810b. By applying the most positive bias voltage to the electrical conductor near the portion of the plasma with greatest ion density, the uniformity of the plasma, as a whole, may be improved. To enhance the improvement in the uniformity, less positive voltage may be applied to one or more electrical conductors near the portion of the plasma with less plasma density. Those of ordinary skill in the art will recognize that the plasma processing apparatus 800 may also be used for improving the uniformity of plasma having different density profiles. In one example, the density profile of the plasma is such that the outer portion has a greater density and the inner portion of the plasma has less density, as shown in FIG. 9A. In such an embodiment, the second electrical conductor 810b near the outer portion of the plasma may be biased with a positive bias voltage. Alternatively or in addition, the first electrical conductor 810a may be biased with less positive bias voltage. In another example, the density profile of the plasma is such that the inner portion and the outer portion of the plasma have less ion density than a portion of the plasma therebetween, as shown in FIG. 9B. In such an example, the electrical conductors near the portion of the plasma with greater ion density may be biased with more positive bias voltage. Meanwhile, the electrical conductors near the portion of the plasma with less ion density may be applied with less positive bias voltage. In the process, the plasma with non-uniform ion density may be made more uniform. By independently applying the bias voltage and controlling the applied bias voltage, the electrical conductors of the present disclosure may locally control the plasma density and improve the plasma uniformity. Referring to FIG. 10A, there is shown a plasma processing system 1000 according to another embodiment of the present disclosure. In FIG. 10B, there is shown a plurality of electrical conductors 1010 that may be included in the plasma processing system shown in FIG. 10A. As shown in FIG. 10B, the electrical conductors 1010 may comprise a first electrical conductor 1010a disposed near the inner portion of the plasma 140, and a second electrical conductor 1010b disposed near the outer portion of the plasma 140. Optionally, there may be one or more intermediate electrical conductors. In the present embodiment, the intermediate electrical conductors may comprise a third electrical conductor 1010c and a fourth electrical conductor 1010d, disposed between the first and second electrical conductor 1010a and 101b. As shown in FIG. 10A, the first conductor 1010a may be disposed near the inner portion of the plasma 140, whereas the second conductor 1010b may be disposed near the outer portion of the plasma 140. In addition, the second electrical conductor 1010b may be configured to surround the first conductor 1010a. Although not necessary, the first and second conductors 1010a and 1010b of the present embodiment may be concentric. In some other embodiments, the first and second conductors 1010a and 1010b may have shapes other than circular shape as shown in FIG. 10A. Much like the prior embodiments, the first and second conductors 1010a and 1010b may be electrically isolated from each other. If included, each of the third and fourth conductors 1010c and 1010d may also be electrically isolated from each other and from each of the first and second conductors 1010a and 1010b. In addition, each of the first and second conductors 1010a and 1010b may be independently biased. If included, the third and fourth conductors 1010c and 1010d may also be independently biased. Each conductor 1010a-1010d may be independently biased with a continuous or pulsed with bias voltage. The bias voltage applied may be positive or negative. Or, at least one of the first and second conductors 1010a and 1010b, and the third and fourth conductors 1010c and 1010d, if included, may remain floating or grounded. In an exemplary operation, a plasma source 101 in a plasma doping system shown in FIGS. 1 and 2 may generate a plasma 140 with non-uniform density shown in FIG. 4. To improve the uniformity, a bias voltage may be provided to one of more of the first and second electrical conductors 1010a and 1010b. For example, if the plasma 140 has higher ion density near its inner portion, as shown in FIG. 4, a first positive bias voltage may be provided to the first conductor 1010a. With the positive bias voltage, the first electrical conductors 1010a near the inner portion of the plasma 140 may disperse the positively charged ions away from the inner portion of the plasma 140, toward the outer portion of the plasma 140. As a result, the uniformity of the plasma 140 may improve. Alternatively, the second conductor 1010b may be applied with a bias voltage that is less positive than what the first bias voltage would have been had the first bias voltage been also applied to the first conductor 1010a. For example, the second conductor 1010b may be applied with negative bias voltage. With the application of negative bias voltage, the plasma may be compressed, and the density of the less dense portion of the plasma may increase. In the process, the uniformity of the plasma may be enhanced. In some embodiments, both the first and second conductors 1010a and 1010b may be independently biased. In this embodiment, the first conductor 1010a may be applied with a first, positive bias voltage, and the second conductor 1010b may be applied with a second, less positive (e.g. negative) bias voltage. If included, the third and fourth electrical conductors 1010c and 1010d may also be applied with bias voltage. To increase the uniformity of the plasma such as the plasma shown in FIG. 4, the third and fourth electrical conductors 1010c and 1010d may be biased with less positive bias voltage than the first electrical conductor 1010a, but more positive bias voltage than the second electrical conductor 1010b. In addition, the third electrical conductor 1010c may be biased with more positive bias voltage than the fourth electrical conductor 1010d. In the process, a plasma with more uniform ion density, as shown in FIG. 8, may be achieved. Although the present disclosure has been described herein in the context of particular systems and particular implementations in particular environments for a particular purpose, the present disclosure is not limited thereto. Those of ordinary skill in the art will recognize that its usefulness is not limited thereto and that the present disclosure can be beneficially implemented in any number of environments for any number of purposes. Accordingly, the claims set forth below should be construed in view of the full breadth and spirit of the present disclosure as described herein.
	abstract	In a fuel assembly, a plurality of fuel rods are arranged in an array of 10 rows and 10 columns in the cross section of the fuel assembly. A flow resistance member is disposed in a central portion in the cross section at upper end portions of partial length fuel rods which are a part of the fuel rods. In the flow resistance member, resistance members are each disposed between ferrules arranged in an array of 6 rows and 6 columns in the diagonal direction of the flow resistance member. Resistance members are each disposed between the ferrules in a peripheral portion of the flow resistance member. By disposing the resistance members, the pressure loss in an inner region in the cross section of the fuel assembly is increased, and the flow rate of a gas-liquid two-phase flow in an outer region surrounding the inner region is increased.
	description	A solid immersion lens (SIL) is a refractive or diffractive, optical element that can be formed on or otherwise affixed to a substrate. Typically, a SIL is part of an objective lens that is brought into adjacent contact with the optical medium through which it is desirable to view an embedded object. A refractive SIL increases the magnification and the resolution of an object buried in the optical medium by modifying refraction as the light passes from the optical medium into air. SILs are becoming commercially available on advanced imaging systems capable of observing buried features with light that penetrates or emits from the medium. A flip-chip application specific integrated circuit (ASIC) fabricated on silicon (Si) is an exemplary candidate for a SIL since light longer than the wavelength of the optical band gap of silicon can easily transmit through the backside silicon of the flip-chip, reflecting off the circuitry beneath to provide an image of the circuitry for diagnostic purposes. FIGS. 1A and 1B are prior art schematic diagrams illustrating the effect of a refractive SIL when viewing circuitry through an optical medium, such as silicon (Si). FIG. 1A shows circuitry 11 formed in silicon 12, which is referred to as “backside silicon” because it extends generally away from the plane on which the circuitry 11 is located. FIG. 1A shows the optical path used to view circuitry 11 on a flip-chip ASIC without a SIL. Light of wavelength longer than the energy band gap of silicon passes through the backside Si 12 and then continues to an objective lens of a microscope, also referred to as a “backing objective” 14. Without a SIL, light refracts as it crosses the boundary between the backside Si 12 and the air 16. This boundary point is shown using reference numeral 21. Light passing through the backside Si 12, shown as light ray 22, forms an angle “θ” with respect to an optical axis 24, which extends normal to the plane 26 on which the circuitry 11 is located. As the light refracts at the boundary point 21, light ray 28 forms an angle “φ” with respect to the optical axis 24, where the angle “θ” is less than the angle “φ” according to Snell's law. FIG. 1B illustrates a sectional prior art illustration of a SIL 30. The SIL 30 is a section of a sphere made from Si and held in intimate contact with the surface 31 of the backside Si 12. In this example, the radius “r” of the SIL 30 is 1.5 mm, and the thickness of the backside Si is approximately 780 μm. The exposed portion 33 of the SIL 30 and the backside Si enclosed in the dashed arc 34 forms a hemisphere. With this geometry, the plane 26 of the circuitry 11 bisects a sphere where the Si hemisphere that forms the SIL 30 is used to direct the light. All light from the center of the sphere crosses the boundary 35 between the backside Si 12 and the SIL 30 without refraction if the SIL 30 and the backside Si 12 are in adjacent contact. Light passing through the backside Si 12, shown as light ray 36, forms an angle “φ” with respect to the optical axis 24. When a SIL 30 is implemented, all light 36 from the center of the sphere crosses the boundary 35 between the backside Si 12 and the SIL 30 without refraction, as shown at points 37, and maintains the constant angle “φ” with respect to the optical axis 24. The light rays 36 then cross the boundary 38 between the SIL 30 and air 16 normal to the boundary 38, exiting the SIL 30 without refraction. The increase in the effective numerical aperture (NA, defined as sin(θ) in FIG. 1A and sin(φ) in FIG. 1B) for the SIL 30 is a key to the improvement in resolution when viewing the circuitry 11. The resolution of the optical system defined by FIG. 1A is the Raleigh condition:R=λSi/(2NAθ), where λSi and NAθ are the wavelength and numerical aperture of the light in the Si, respectively. Relative to their values in air, the wavelength of light in Si is λ/n where n=3.5 is the index of refraction of Si near-IR wavelengths (1.1 μm to 1.7 μm), and NA is governed by Snell's law nsin(θ)=sin(φ) with φ being the angle of the light after refraction. In FIG. 1B the surface of the Si is reshaped to be hemispherical to prevent refraction. Since all light rays in FIG. 1B strike the Si/air surface perpendicularly, refraction vanishes and the resolution becomes:RSIL=λSi/(2*NAφ), where Snell's law no longer affects NA. The net effect of the hemispherical surface is to improve the resolution defined by the Raleigh condition according to the relationship:RSIL=R/n, and to improve the magnification by a factor of n. The configuration of the SIL 30 is called a centric SIL because the object (portions of the circuitry 11 that are at a focal area of the SIL 30) is physically at the center of the hemisphere. In practice, the SIL 30 does not require the exact geometry shown because the backing objective 14 can move in the vertical dimension to compensate although the resolution and magnification will be affected. A SIL is commercially available as a separate structure, or can be commercially formed on a surface of an optical medium using a focused ion beam (FIB) projected through a gray scale rendering of a milling pattern. A focused ion beam (FIB) uses a beam of Ga+ ions to strike and mechanically erode a surface of an optical medium. The length of time the Ga+ beam dwells at a point determines the depth of the mill. A prior technique can be used to form a hemispherical surface in an optical medium by projecting the hemispherical shape into a two-dimensional gray scale image where darker gray scale levels correspond to deeper milling. The gray scale then determines the dwell time, i.e. the length of time the FIB mills at each point. FIG. 2 is a diagram illustrating a two-dimensional gray-scale rendering 50 of a three-dimensional hemisphere. The two-dimensional gray-scale rendering 50 can be used to create a milling pattern to form the SIL 30. Such a SIL is formed as a hemispherical structure directly on the optical medium. Unfortunately, many FIB milling tools cannot use a two-dimensional gray-scale rendering to control the milling performed by the FIB. Therefore, it would be desirable to have an alternative way of forming a high quality SIL on an optical medium. In an embodiment, a method for forming a solid immersion lens (SIL) includes generating a focused ion beam, and projecting the focused ion beam onto an optical medium at locations defined by a binary bitmap milling pattern, wherein the locations at which the focused ion beam impact a surface of the optical medium are randomized over successive raster scans of the surface of the optical medium to form at least a portion of a hemispherical structure in the optical medium. Other embodiments are also provided. Other systems, features, and advantages of the invention will be or become apparent to one with skill in the art upon examination of the following figures and detailed description. It is intended that all such additional systems, methods, features, and advantages be included within this description, be within the scope of the invention, and be protected by the accompanying claims. Embodiments of the apparatus and method for forming a solid immersion lens (SIL) using a binary bitmap milling pattern are implemented by a focused ion beam (FIB) milling apparatus to create hemispherical surfaces on the backside Si of an application specific integrated circuit (ASIC) and other circuit structures. For example, in an integrated circuit architecture that employs a “flip chip” architecture it is desirable to be able to visually inspect at least portions of the active circuitry through the bulk silicon. FIG. 3 is a diagram illustrating a binary bitmap rendering of a gray-scale image. The binary (i.e., black and white) bitmap rendering 100 can be generated by interpreting a gray-scale as a probability that a pixel should be black or white. On a gray scale of 0 to 255, for example, a level of 200 could mean a probability of 200/255=0.784 that the pixel should be white. The binary bitmap rendering 100 illustrates a field 110 of pixels that can be set either to black or white. An exemplary black pixel is illustrated at 112 and an exemplary white pixel is illustrated at 114. A black pixel indicates a location where the Ga+ beam would impact and abrade the surface of the backside Si and a white pixel indicates a location where the Ga+ beam would not impact or abrade the surface of the backside Si. The number of pixels in the bitmap is determined by the geometry and milling parameters of the FIB milling apparatus. In a typical embodiment, the bitmap is approximately 100×100 to 300×300 pixels. The pattern can be used to define a hemispherical structure formed in the backside Si. In this manner, the binary bitmap rendering 100 is created and can be used to control the beam of an FIB milling machine. When converting a high-resolution image (for example, a gray-scale image) to a low-resolution image (for example, a binary bitmap rendering) a loss of information may occur due to a condition known as aliasing. Aliasing can result in a jaggedness appearing on the surface of a SIL created using the binary bitmap rendering 100. There are a number of antialiasing techniques that can be implemented to reduce the jaggedness of the surface of the SIL and result in a smooth surface. In addition to the binary bitmap rendering 100, and to create a hemispherical structure having a smooth surface having a high optical quality, in an embodiment, it is desirable to be able to rotate the rastering axes of the Ga+ beam during the milling process relative to the surface of the backside Si. The rastering axes of the Ga+ beam can be rotated while the Si can be held stationary, the Si can be rotated while the rastering pattern of the Ga+ beam is held stationary, or both the rastering axes of the Ga+ beam and the Si can be rotated relative to each other to create relative rotational movement between the Ga+ beam and the Si. Creating relative rotation between the rastering axes of the Ga+ beam and the Si randomizes the dwell time of the Ga+ beam on the surface of the backside Si according to the binary bitmap rendering 100, thereby antialiasing the binary bitmap and providing a uniformly smooth surface having a high optical quality. An alternative technique for randomizing the milling imparted to the surface of the backside Si is to vary the distribution, sequence, or arrangement, of black and white pixels in the binary bitmap rendering 100 to create one or more successive binary bitmap renderings that functionally represents the original gray-scale image, but that have a different distribution of black and white pixels than does the original binary bitmap rendering, so long as the distribution of black and white pixels still provides the desired hemispherical surface. A number of different binary bitmap renderings 100 can be created and used in successive milling operations to randomize the milling imparted to the surface of the backside Si without using relative rotational movement between the Ga+ beam and the surface of the backside Si. Repeatedly milling the surface of the backside Si using a number of different successive binary bitmap milling patterns randomizes the dwell time of the Ga+ beam on the surface of the backside Si, thereby providing a uniformly smooth surface having a high optical quality. In an embodiment, 0-255 gray levels can be collapsed down to 2 levels. Therefore, 256/2=128 binary bitmaps will regain any information lost due to aliasing. That is, 128 binary bitmap renderings are the maximum needed. If the Ga+ beam is defocused so that it covers 3×3=9 pixels at a time, 128/9=14 binary bitmap renderings can be used to make a reasonably smooth surface. In such an embodiment, fourteen unique binary bitmap renderings can be successively applied during the milling process to create a smooth, hemispherical surface. FIGS. 4A through 4C collectively illustrate an apparatus that can use the binary bitmap rendering 100 of FIG. 3 and method described above to form a hemispherical structure in an optical medium. As mentioned above a focused ion beam (FIB) can be used to perform milling on an optical medium, such as, but not limited to, bulk silicon. In an embodiment, a typical FIB uses a liquid metal ion source (LMIS) to produce a beam of Ga+ ions that is focused by a column of ion optics onto a sample surface. The ion optical column shapes the beam in magnitude (e.g., several picoamperes (pA) to many nanoamperes (nA)), voltage (e.g., tens of kilovolts (kV)), and size (e.g., several nanometers (nm) to several micrometers (μm)), and has steering plates (not shown) at the exit of the milling head that can move the nominal center of the beam to a new spot on the sample, raster the beam in a boustrophedonic pattern over an area (ranging from approximately one square micrometer (μm2) to approximately one square millimeter (mm2) to produce an image of the surface, and rotate the boustrophedon to any pair of orthogonal axes in the X-Y plane. As used herein, the term “boustrophedon” refers to a pattern of producing lines of Ga+ beam that alternate from left to right and from right to left. In a typical application, the thickness of the bulk silicon substrate is between approximately 40 μm and 780 μm. In an embodiment, it may be desirable to mechanically or chemically thin the backside Si to a thickness thinner that 780 μm prior to milling. For example, it may be desirable to thin the backside Si to a thickness of approximately 40 to 120 μm to facilitate the production of an effective SIL quickly since milling large patterns can take several hours with existing equipment. The apparatus 200 includes a circuit device 201 on which it is desired to form a hemispherical structure, which can be used as a SIL. The circuit device 201 includes active circuit portion 204 and bulk silicon 206. The bulk Si 206 is also referred to as the backside Si. In an embodiment, the bulk silicon 206 is approximately 780 μm thick. The active circuit portion 204 includes circuitry, the visual inspection of which is desirable through the bulk Si 206. The circuit device 201 is located on a movable surface 202 within a FIB chamber (not shown for simplicity). In an embodiment, the movable surface 202 can be part of a table, or other type of support structure or platform, the location of which can be precisely controlled in three dimensions, typically in the X, Y and Z dimensions; and in rotation about the X, Y and Z axes. The apparatus 200 also includes a milling head 212 that produces a focused ion beam 214. The milling head 212 is part of a milling apparatus that generates, focuses and controls the power of the focused ion beam, as known in the art. The milling head 212 is positioned above the surface 207 of the bulk silicon 206 so that the focused ion beam 214 can be directed to specific areas of the surface 207 of the bulk silicon 206 to mechanically erode selected portions of the surface 207 to form a hemispherical structure. The milling head 212 includes a support structure 218 which fixes it in space. The movable surface 202 allows the sample to be placed under the milling head 212, and the steering elements (not shown) of the ion column allow rotation of the boustrophedonic raster pattern around the Z axis. The milling head 212 also includes an opening 222, which allows the focused ion beam 214 to exit the milling head 212 and to image and mechanically erode the bulk silicon 206. In an embodiment, the focused ion beam can be a Ga+ beam generated by using a liquid metal ion source (LMIS). However, other materials can be used to generate other types of focused ion beams. In an embodiment, the focused ion beam can have a current of approximately 4.4 nanoamperes (nA) and a physical size of approximately 0.03 to 10 micrometers (μm) at the focus. The focused ion beam can raster over the surface 207 in a boustrophedon to encompass an area of approximately 10,000 μm2 for a 100×100 pixel array where each pixel is 1 μm2. The focused ion beam can dwell at a given point for approximately 0.5 μs, can step approximately 0.3 μm to 1 μm to the next dwell point, and can then repeat. The milling head 212 directs the focused ion beam 214 onto the surface 207 through a binary bitmap milling pattern 120. The movable surface 202 holding the circuit device 201 can be controlled so that the circuit device 201 can be precisely located in three dimensions about the X, Y and Z axes relative to the milling head 212. Rotating the axes of the boustrophedonic rastering pattern of the focused ion beam 214 about the Z axis causes the binary bitmap milling pattern 120 to rotate about its center over the surface 207 of the bulk silicon 206, thereby providing cylindrical symmetry to the erosion rate of the Ga+ beam about the Z axis on the surface 207 of the bulk Si 206, and thereby provide a cylindrically uniform, smooth erosion pattern. The binary bitmap milling pattern 120 follows the rastering axes. For example, at a beginning time, t=0, the X axis of the sample (the circuit device 201) is parallel to the X axis of the motor drive (not shown) of the movable surface 202 and the Y axis of the sample (the circuit device 201) is parallel to the Y axis of the motor drive (not shown) of the movable surface 202. The Z axis is vertical. The rastering axes are parallel to the X and Y axes described. With rotation on, the rastering axes can be rotated relative to the stationary X and Y axes of the movable surface 202. The X axis on the display (642, FIG. 6) is horizontal and the Y axis is vertical. If the rotation of the Ga+ beam is 1 Hz, then 0.25 seconds later the x axis used to define the rastering pattern on the surface 207 of the circuit device 201 is along a diagonal relative to the X and Y axes of the motor drives of the movable surface 202, and the Y axis is still perpendicular to the X axis. The milling system draws the image on the display with the X axis being horizontal and the Y axis being vertical. The net effect is that the image on the display appears to rotate as if the surface 207 of the circuit device 201 had actually rotated while the Ga+ beam remained constant. The sample remains stationary, but the rastering axes of the Ga+ beam have rotated by 45 degrees, making it appear that the binary bitmap milling pattern 120 also rotated accordingly. The rotation is an electronic means of simulating rotation of the movable surface 202. At the dimensions described herein, true mechanical motion would result in a surface having a significant amount of error. Therefore, the rotation feature smoothly simulates rotation of the circuit device 201 about the Z axis. The software that controls the FIB milling apparatus either directs the Ga+ beam to the sample surface, or deflects the Ga+ beam out of the column where it strikes a plate. The Ga+ beam rasters a boustrophedon with the binary bitmap milling pattern 120 superimposed. If a pixel is black, the beam passes through to the bulk Si 206. If the pixel is white, the beam is deflected away from the bulk Si 206 into a plate within the column. The gray-scale image is rendered to a binary bitmap, and the binary bitmap controls whether the beam exits the milling head 212 (black) and impacts the surface 207, or strikes the plate within milling head 212 (white) and does not impact the surface 207. The focused ion beam 214 mechanically erodes the surface 207 at the locations defined by the black pixels (FIG. 3) on the binary bitmap rendering 100, resulting in a section of a hemispherical structure being formed in the bulk Si 206. FIG. 4B is a schematic diagram illustrating an apparatus 230 in which the bulk Si 206 has been eroded to partially expose a portion 234 of a hemisphere 232. The application of the Ga+ focused ion beam 214 through the binary bitmap milling pattern 120 results in a cavity 224 being formed (i.e., milled) into the bulk silicon 206 in accordance with the milling process described above. In accordance with an embodiment of the apparatus and method for forming a solid immersion lens (SIL) using a binary bitmap milling pattern, the focused ion beam 214 is applied to the surface 207 through the binary bitmap milling pattern 120 for a predetermined period of time. In an embodiment, the milling process can be applied for approximately one hour with the focused ion beam current at approximately 4.4 nA. However, this predetermined time and beam energy can vary substantially and is dependent upon the dimensions of the structure sought to be formed and other factors. The cavity 224 formed (i.e., milled) into the bulk silicon 206 exposes a sector 234 of the hemisphere 232. The exposed sector 234 and the backside Si enclosed in the dashed arc 236 forms the hemisphere 232. FIG. 4C is a schematic diagram illustrating an apparatus 250 in which the bulk Si 206 has been eroded to expose a sector 238 of the hemisphere 232. The exposed sector 238 and the portion of the backside Si defined by the dashed arc 236 forms a SIL 235. FIG. 5 is a plan view illustrating an exemplary binary bitmap milling pattern. The binary bitmap milling pattern 120 is shown in plan view superimposed over the surface 207 of the bulk Si 206. The outer diameter of the SIL 235 (FIG. 4C) is illustrated using reference numeral 505. The binary bitmap milling pattern 120 is divided into square regions 505, each region 505 depicting a pixel in a 16×16 array of pixels. A 16×16 array is shown for simplicity of illustration. In practice, a larger array of between 100×100 and 300×300 pixels is typically used. An example of the milling process begins at an exemplary pixel 506 and progresses in a boustrophedon along the direction indicated by the arrow 510 until reaching an exemplary ending pixel 508. To achieve the boustrophedon illustrated by the arrow 510, the Ga+ beam is electronically steered along the path indicated by the arrow 510 so that the beam impacts the surface 207 in locations where a black pixel is present and is prevented from impacting the surface 207 where a white pixel is present. In this manner, a hemispherical shape is formed on the surface 207. In an embodiment, one progression of the Ga+ beam across all of the pixels in the binary bitmap milling pattern 120 is completed in approximately 50 milliseconds (ms). During the progression, the rastering axes of the projected Ga+ beam are rotated at a rate of approximately 0.5 Hz while the process repeats. Rotating the rastering axes of the Ga+ beam for each pixel over the surface 207 of the bulk Si 206 randomizes the dwell time of the Ga+ beam on the surface of the backside Si according to the binary bitmap rendering 100, thereby providing a uniformly smooth surface having a high optical quality. In alternative embodiments, the Ga+ beam can be defocused to encompass more than a single pixel to facilitate antialiasing. In an alternative embodiment, the movable surface 202 can be rotated about the Z axis by a table drive element 602 (FIG. 6) to create relative movement between the Ga+ beam and the surface 207 of the bulk Si 206, while the rastering axes of the Ga+ beam remains stationary relative to the Z axis. In another alternative embodiment, a number of different, but functionally equivalent, binary bitmap milling patterns 120 can be created and used for each progression of the Ga+ beam along the path indicated by the arrow 510 without creating relative rotation between the Ga+ beam and the surface 207 of the bulk Si 206. In this embodiment, each successive and randomly created binary bitmap milling pattern 120 will likely have different white and black pixels, while the overall distribution of white and black pixels still creates the hemispherical surface described above. In this manner over successive milling iterations using different binary bitmap milling patterns, the locations where the Ga+ beam impacts the surface 207 is sufficiently randomized to create the hemispherical shape described above having a smooth surface having high optical quality. FIG. 6 is a block diagram illustrating an embodiment of an apparatus 600 for forming a solid immersion lens (SIL) using a binary bitmap milling pattern. The apparatus 600 includes the apparatus portion 200 adapted to receive a liquid metal ion source (LMIS) 608, from which a focused ion beam 214 is generated and directed by the milling head 212. As shown in FIG. 6, the substrate on which the SIL is made is located on a table drive element 602. The table drive element 602 is controlled by a controller 620 and allows controllable movement in a horizontal plane defined by the X and Y directions and in a vertical plane defined by the Z direction. The table drive element 602 is used to center the desired portion of the circuit device 201 under the focused ion beam 214. The milling head 212 is positionally fixed and incapable of movement. Rotation of the binary bitmap milling pattern 120 is controlled by the steering elements (not shown) of the ion beam column. In an embodiment, the focused ion beam 214 makes one complete pass of the bitmap in approximately <100 ms, and the raster pattern makes one complete revolution relative to the surface in about 2 s. Therefore, the focused ion beam 214 makes 20 complete passes of the binary bitmap milling pattern 120 in one complete revolution of the focused ion beam 214. Alternatively, it is possible for the table drive element 602 to rotate eucentrically about the Z axis to rotate the bulk Si 206 about the axis (for example, the Z axis) of the focused ion beam 214, and thereby realize the equalization of the impact of the focused ion beam 214 on the surface 207 according to the binary bitmap milling pattern 120. As used herein, the term “eucentrically rotate” refers to ability to define a point as the center of rotation. To illustrate, assume a motor rotates the movable surface 202 (FIG. 4B) about a center on the Z axis located at x=0, y=0. Now assume that it is desired to rotate the movable surface 202 about a center at a location x=1, y=1. The system software 624 rotates the movable surface 202 and simultaneously translates the movable surface 202 in the X and Y dimensions to keep the point x=1, y=1 from moving relative to the Z axis. The controller 620 includes a system processor 622, system software 624, a table drive controller 626, an input/output (I/O) element 628, an image processor 632, and a milling head controller 644 coupled together over a system bus 634. The system bus 634 can be any communication bus that allows bi-directional communication between and among the connected elements. The controller 620 also includes a display 642. The system processor 622 can be any general-purpose or special-purpose processor or microprocessor that is used to control the operation of the system 600. The software 624 can contain executable instructions in the form of application software, execution software, embedded software, or any other software that controls the operation of the controller 620 and the elements in the system 600. The table drive controller 626 is operatively coupled to the table drive element 602 over connection 646. The milling head controller 644 is operatively coupled to the support structure 218 over connection 636. In accordance with an embodiment of the system for forming a solid immersion lens (SIL) using a binary bitmap milling pattern, the table drive controller 626 is used to control the relative position of the focused ion beam 214 and the surface 207 of the bulk Si 206, so that the circuit device 201 can be moved in a plane, defined in the X and Y directions. FIG. 7 is a perspective view of an optical medium 700 including a SIL formed thereon. The optical medium 700 can be, for example, a portion of bulk Si 206 as described above. A SIL 235 is formed in the bulk Si 206 using the apparatus and method for forming a solid immersion lens (SIL) using a binary bitmap milling pattern, as described herein. FIG. 8 is a flowchart 800 illustrating the operation of an embodiment of a method for forming a solid immersion lens (SIL) using a binary bitmap milling pattern. The blocks in the flow chart 800 can be performed in or out of the order shown. In addition, at least some of the steps can be performed in parallel. In block 802 a binary bitmap milling pattern 120 is created using the binary bitmap rendering 100 of FIG. 3. In block 804, the bulk Si 206 is milled using a focused ion beam and the binary bitmap milling pattern created in block 802. In block 806, the point at which the focused ion beam impacts the surface of the backside Si for each pixel in the binary bitmap milling pattern is randomized for successive scans. In an embodiment, the rastering axes of the focused ion beam are rotated over the surface of the bulk Si to form a SIL having a smooth surface of high optical quality. In another embodiment, the bulk Si is rotated while the rastering pattern of the focused ion beam remains stationary. In yet another embodiment, a number of different, but functionally equivalent, binary bitmap milling patterns 120 are created and sequentially used in sequential raster scan milling operations to form a SIL having a smooth surface of high optical quality. In block 808, the milling process is terminated and the process ends. This disclosure describes the invention in detail using illustrative embodiments. However, it is to be understood that the invention defined by the appended claims is not limited to the precise embodiments described.
	abstract	The present invention relates to a passive heat removal system which circulates cooling fluid to a steam generator via a main water supply line connected to the lower inlet of the steam generator, and a main steam pipe connected to the top outlet of the steam generator, to remove sensible heat of a nuclear reactor coolant system and residual heat of a core. The heat removal system comprises supplementary equipment for receiving surplus cooling fluid or for supplying supplementary cooling fluid in order to maintain the flow rate of the cooling fluid within a predetermined range. The supplementary equipment comprises: a supplementary tank, installed at a height between the lower inlet and the top outlet of the steam generator; a first connection pipe, connected to the main steam pipe and the supplementary tank; and a second connection pipe, connected to the supplementary tank and the main water supply pipe.
	claims	1. A radiation-attenuation shirt, comprising:a plurality of radiation-attenuating material panels adapted to conform to the contours of a body, having a plurality of attaching mechanisms on one side;a front portion, made of a compression material and having a plurality of attaching mechanisms disposed on one side of the front portion, the front portion including a first pocket for retaining a first radiation attenuating material panel; anda back portion, made of a compression material and having a plurality of attaching mechanisms disposed on one side of the back portion, the back portion including a second pocket for retaining a second radiation attenuating material panel,wherein the front portion and the back portion are secured together to form a shirt, such that the attaching mechanisms and the first and second pockets are disposed within the shirt,wherein the first radiation-attenuating material panel is removably disposed within the first pocket and the plurality of attaching mechanisms of the first radiation-attenuating material panel are removably coupled to the plurality of attaching mechanisms of the front portion,wherein the second radiation-attenuating material panel is removably disposed within the second pocket and the plurality of attaching mechanisms of the second radiation-attenuating material panel are removably coupled to the plurality of attaching mechanisms of the rear portion. 2. The radiation-attenuation shirt of claim 1, wherein the radiation attenuating panels are comprised of lead. 3. The radiation-attenuation shirt of claim 1, wherein the radiation attenuating panels are comprised of lead alloy. 4. The radiation-attenuation shirt of claim 1, wherein the radiation attenuating panels are covered in a removable, machine washable material. 5. The radiation-attenuation shirt of claim 1, wherein the front portion and the back portion include a breathable mesh proximate to the radiation attenuating panels. 6. A radiation-attenuation garment system, comprising:a plurality of radiation-attenuating material panels adapted to conform to the contours of a body;a radiation attenuation shirt, comprising:a front shirt portion, made of a compression material; anda back shirt portion, made of a compression material,wherein the front portion and the back portion are secured together to form a shirt, such that a first radiation-attenuating material panel is removably disposed within the shirt;radiation-attenuation underwear shorts, comprising:a front underwear portion, made of a compression material; anda back underwear portion, made of a compression material,wherein the front underwear portion and the back underwear portion are secured together to form underwear shorts, such that a first radiation-attenuating material panel is removably disposed within the underwear shorts. 7. The radiation-attenuation garment system of claim 6, wherein the radiation attenuating panels are comprised of lead. 8. The radiation-attenuation garment system of claim 6, wherein the radiation attenuating panels are comprised of lead alloy. 9. The radiation-attenuation garment system of claim 6, wherein the radiation attenuating panels are covered in a removable, machine washable material.
	description	This current application is a continuation of a application Ser. No. 13/042,444, filed Mar. 7, 2011 now U.S. Pat. No. 8,134,216, with the same title, inventors, assignee, and specification, which was recently allowed. Thus, this current application incorporates by reference all of the teachings and specification of its parent case, as also included here. Ser. No. 13/042,444 in turn is a continuation-in-part of (and related to) U.S. application Ser. No. 12/888,521 filed Sep. 23, 2010 now U.S. Pat. No. 8,017,412, and Ser. No. 12/851,555, filed Aug. 6, 2010 now U.S. Pat. No. 8,487,392, which are based on the provisional applications 61/250,504, filed Oct. 10, 2009, 61/231,863, filed Aug. 6, 2009, and 61/306,541, filed Feb. 21, 2010, with common inventor(s), and same assignee (Widetronix Corporation). All of the above teachings are incorporated by reference here. We introduce a new technology for Manufactureable, High Power Density, High Volume Utilization Nuclear Batteries. Betavoltaic batteries are an excellent choice for battery applications which require long life, high power density, or the ability to operate in harsh environments. In order to optimize the performance of betavoltaic batteries for these applications or any other application, it is desirable to maximize the efficiency of beta particle energy conversion into power, while at the same time increasing the power density of an overall device. Increasing power density is a difficult problem because, while both the active area of the semiconductor used for the beta energy conversion and the layer of radioisotope that provides the betas for this conversion are very thin (100's of nanometers), the thickness of the substrate supporting the radioisotope layer and the overall thickness of the semiconductor device wafers are on the order of 100's of microns. In another embodiment for this technology, there are several technical constraints that must be considered when designing a low cost, manufacturable, high volume, high power density silicon carbide (SiC) betavoltaic device. First, consideration must be given to the energy profile of radioisotopes to be used, and the volume at which such material can be produced. For example, tritium is one of the several viable radioisotope candidates, since it can be produced in sufficient quantities to support high volume device manufacture, and its energy profile fits well with a range of power generation design parameters. Secondly, in order to produce high power density in betavoltaics, a large device surface area is required. There are issued and pending betavoltaic patents that mention patterning methods for pillars, pores or other structures which yield such high surface area—patent application Ser. No. 11/509,323 is an example, and can be used as a reference for pillared betavoltaic device construction. These methods must be optimized appropriately in order to meet fabrication objectives, while controlling costs. Thirdly, SiC has been shown to be the ideal material for betavoltaic devices, e.g. see reference patent application Ser. No. 11/509,323. However, SiC has unique processing, fabrication and design requirements which must be met in order to produce a workable device. For example, fabrication of SiC devices requires high temperature epitaxial processes. Because of such high temperature requirements, these epitaxial processes add an element of complexity and cost, not seen with processes relating to other semiconductors, such as Si, and must be taken into account accordingly, or fabrication techniques must be developed to remove such complex and costly processes entirely. Fourthly, it is desirable to integrate betavoltaic devices directly with Silicon (Si)-based electronics, including, but not limited to, microprocessor and memory devices. Thus, there is a need for designs and fabrication processes which anticipate such integration. Devices which address or anticipate the aforementioned design considerations are disclosed in this current or co-pending applications, as mentioned above. Methods for fabricating same are also disclosed. The small (submicron) thickness of the active volume of both the isotope layer and the semiconductor device is due to the short absorption length of beta electrons. The absorption length determines the self absorption of the beta particles in the radioisotope layer as well as the range, or travel distance, of the betas in the semiconductor converter which is typically a semiconductor device comprising at least one PN junction. We define a volume utilization factor, Volutilization, to quantitatively track how well a betavoltaic device is using the volume of the radioisotope source and the volume of the semiconductor converter (equation 1). To illustrate this, consider the simple betavoltaic structure shown in FIG. 1. There are three important length scales for optimization of such a device: 1) the self absorption length of the beta electrons in the radioisotope 2) the range of the beta electrons in the semiconductor converter material 3) the diffusion length of minority carriers in the semiconductor, Ldiff. Ldiff determines the maximum thickness of any doped region (p-type or n-type) forming the PN junction. Note that although these design principles apply to any semiconductor material, including, but not limited to Si, GaAs, GaN, and diamond, herein, we focus on SiC because SiC has been shown to be the ideal material for a beta converter. Also, this invention can be implemented using any beta emitting radioisotopes. Herein, we will consider the three isotopes Nickel-63 (N63), tritium (H3) and the tritides (Scandium Tritide, Titanium Tritide, etc.), and promethium-147 (Pm147). These isotopes have properties as listed in table 1. In this illustration for a simple structure shown in FIG. 1, the radioisotope is supplied by means of a foil. This foil could be carrying either N63, a tritiated metal such as scandium Tritide, or Pm147. We denote the range of the betas in SiC as LSiC and the self absorption length in the radioisotope as Lisotope. The volume utilization in this geometry, neglecting the contacts and isotope volume, is calculated as: Vol utilization = ( t cell ) ⁢ Area ( t substrate + t cell ) ⁢ Area = ( t cell ) ( t substrate + t cell ) ( 1 ) Where Area=the total device area, and tsubstrate=the thickness of the SiC substrate tcell=the thickness of the active SiC region. Note that the value of Volutilization is between zero and one. In order to maximize the power output, this planar style betavoltaic device has to be designed to capture as close to all of the beta electrons leaving the surface of the foil as possible. This means that tcell must be at least greater than the diffusion length of the minority carriers (tcell>Ldiff). However, any material thicker than this limit will not actively participate in energy conversion, so while tcell>Ldiff must be true, tcell must be as close as possible to Ldiff so as to maximize volume utilization. Further, the location of the PN junction depth from the surface of the device must be <Ldiff in order to collect the maximum number of electron hole-pairs. In addition, one embodiment of this invention is a novel SiC betavoltaic device which comprises one or more “ultra shallow” P+N− SiC junctions and a pillared or planar device surface. Junctions are deemed “ultra shallow”, since the thin junction layer (which is proximal to the device's radioactive source) is only 300 nm to 5 nm thick. In one embodiment of this invention, tritium is used as a fuel source. In other embodiments, radioisotopes (such as Nickel-63, promethium or phosphorus-33) may be used. This is also addressed in our co-pending applications, mentioned above. Here are some embodiments of this invention: In order to maximize the power output, this planar style betavoltaic device has to be designed to capture as close to all of the beta electrons leaving the surface of the foil as possible. This means that tell must be at least greater than the diffusion length of the minority carriers (tcell>Ldiff). However, any material thicker than this limit will not actively participate in energy conversion, so while tcell>Ldiff must be true, tcell must be as close as possible to Ldiff so as to maximize volume utilization. Further, the location of the PN junction depth from the surface of the device must be <Ldiff in order to collect the maximum number of electron hole-pairs. TABLE Iβ-emitting radioisotope and theirranges in SiC and self absorption lengthsβ-EmittingMeanSelf absorption lengthSiC absorption lengthIsotopesenergy(at mean beta energy)(at mean beta energy)N6317.4 keV0.67 μm1.84 μmScandium 5.6 keV0.27 μm0.25 μmTrititidePromethium 67 keV8.59 μm19.56 μm Once the output power has been maximized, the only way to increase the power density is to reduce the thickness of the substrate by wafer polishing. A typical SiC wafer is about 350 microns, so if the thickness of the substrate was reduced to 50 microns, this would result in a seven times increase in power density. The total power out of this planar betavoltaic device is given by:PTotal=CtisotopeArea(SSSA) (2) If we take into account the substrate thickness tsubstrate, the power density produced by this geometry is given as: P Density = P Total Total ⁢ ⁢ Device ⁢ ⁢ Volume = Ct isotope ⁢ Area ⁡ ( S SSA ) ( t substrate + t cell ) ⁢ Area = Ct isotope ⁢ S SSA ( t substrate + t cell ) ( 3 ) The conversion constant C takes into account the energy per beta electron the semiconductor loses (phonon, recombination etc.), the reflection of beta electrons at the semiconductor interface, the emission spectrum of the foil, and is directly related to the device efficiency. ‘Area’ is the area of the device as viewed from the top, and the thickness of the radioisotope is denoted by tisotope. SSSA is the specific surface activity, and is defined as the number of electrons per unit area which leaves the surface of the foil in the direction of the converter. This quantity is a measured value for a particular foil. For a particular thickness, tisotope, of the radioisotope, only the betas that are not self absorbed leave the surface and are made available for harvesting by the SiC converter. This thickness of the radioisotope within which all the beta particles generated can leave the surface is called the self absorption length. The self absorption length of the beta particles with average energy is denoted by Lisotope. For the semiconductor, the range of penetration into the SiC of the beta particles with average energy is denoted by LSiC. Both LSiC and Lisotope are calculated from the following relationship. Range ⁡ ( in ⁢ ⁢ microns ) = 4 100 ⁢ ⁢ ρ ⁢ E ⁡ ( in ⁢ ⁢ keV ) 1.75 ( 4 ) where ρ is the density of either SiC or the radioisotope foil, and an expression for the ratio of the density of the two SiC to radioisotope can be written as: L SiC L isotope = ρ isotope ρ SiC ( 5 ) One embodiment of the invention is shown in FIG. 2. While the invention can be implemented with multiple junctions, this first embodiment will be described using a single junction. The top part of FIG. 2 shows the starting geometry which can be viewed as a combination of two slabs—a radioisotope slab and a SiC converter slab. The top slab (shown in red) is the radioisotope slab, and the bottom slab (shown in blue and yellow) is the PN junction slab. The top surface cross sectional dimensions (not shown) of the semiconductor slab are cellx and celly in the x and y directions respectively, and the z dimension (the thickness of the junction, also not shown) is denoted by tcell. In one example, we introduce additional isotope slabs to completely surround up to all four sides of the PN junction slab plus one isotope slab covering the junction slab's bottom or top surface or two additional slabs covering both the top and bottom junction surface. Multiple, and typically thousands, of these isotope enclosed semiconductor slabs will be fabricated across the wafer, resulting in a total top surface area of semiconductor slabs and isotope slabs equal to the final footprint of the new betavoltaic device. For comparison purposes, in this document, the total surface area of the high volume utilization betavoltaic design will approximate the original planar betavoltaic geometry area denoted as “Area” in the description of that planar device in the section above. Note that there can be embodiments of this high volume utilization betavoltaic invention that use two isotope slabs, or three, or up to six isotope slabs, or e.g. the maximum number that can be physically added. For a given thickness of the junction, tcell, an increase in the number of isotope slabs will lead to an increase in the amount of beta electrons per unit volume available for harvesting by the betavoltaic, and therefore, an increase in the amount of power out for the overall total area of a device. The relationship between the total area of the betavoltaic device and the cross sectional area, Acell, of the individual semiconductor slabs can be found by taking advantage of the square cross section of the slab design and creating a unit cell that includes both the semiconductor slab cross section and the isotope slabs surrounding it as shown in FIG. 2b. Then the area of the unit cell, Auc, is given by:Auc=(cellx+2tisotope)(cellY+2tisotope) (6) For illustrative purposes, the semiconductor slab dimensions cell and cells shall be equal, however, in some embodiments of the invention this may not be the case. If cell and cells are equal, then:cellx=cellY And Auc becomes:Auc=(cellx+2tisotope)(cellx+2tisotope)Auc=(cellx+2tisotope)2 (6b) The total area, denoted as “Area”, covered by all the N unit cells on the device is equal to:Area=N(cellx+2tisotope)2 (7) And N, the number of cells in the active area of the device, can be found from: N = Area ( cell x + 2 ⁢ ⁢ t isotope ) 2 ( 7 ⁢ a ) The values of each of the parameters defined above are determined by the material characteristics of both the isotope and the semiconductor. The following is a listing of the parameters and their determining material characteristics: tcell: This parameter is determined by the minority carrier diffusion length, Ldiff, of the semiconductor material. It is important that all the electron hole pairs that are formed in the device active area can make it back to the junction. Keeping tcell close to Ldiff will ensure the maximum collection of electron-hole pairs. In some embodiments of the invention, the range for tcell can be 1 μm to 150 μm. cellx: This parameter is determined by the range of the betas in the semiconductor, which means that it is also isotope dependent. Because there are isotope slabs on all four sides of the semiconductor slab in one or more embodiments of the invention, then for these embodiments, the cross section of the semiconductor slab can be substantially square to give equal range to the betas in all directions. In some of these embodiments of the invention, the range for cell can be 0.5 μm to 250 μm. tisotope: This parameter is determined by the self absorption length, Lisotope, of the betas in their respective isotope sources. In one embodiment, tisotope is at least equal to Lisotope to ensure the most efficient volumetric use of the isotope slab. In some embodiments of the invention, the range for tisotope can be 0.1 μm to 20 μm. One of the major differences between the planar betavoltaic design as well as designs which use textured active device areas with PN junctions that are conformal to a textured surface geometry, and this new high volume utilization betavoltaic invention is that certain surfaces/faces of as many as four isotope slabs are substantially perpendicular to one or more semiconductor slab PN junctions, thus, a significant amount of the betas whose energy are being harvested and used for power conversion enter the device in both the n-type and p-type regions within a diffusion length, Ldiff, of the junction(s). Using this configuration, we can significantly increase the number of betas per unit volume which can be harvested which will directly impact the total power output of the cell, as well as the power density. To further illustrate the improvements of the invention over a planar device, we can calculate the relative power, PRel, of the new high volume utilization betavoltaic design relative to the standard planar betavoltaic design. The relative power is the ratio of the power of the high volume utilization geometry to the power of the planar single isotope slab geometry, or: P Rel = P multi - slab P planar ( 8 ) The following are examples of Prel calculations for 6, 5 and 3 isotope slabs. As mentioned herein, other slab configurations in terms of slab quantity and position are possible. The power for the high volume utilization betavoltaic invention with six isotope slabs, P6 slabs, is given byP6 slabs={Ctisotope{[4cellxtcell]+[2(cellx)2]}SSSA}Nαedge2 (9) Where αedge is an edge effect factor that adjusts for the intrinsic attenuation of the beta current from the isotope slabs around each individual SiC cell. To calculate Prel we need the output power for the planar betavoltaic which was given in equation (2a) as:PPlanar=CtisotopeArea(SSSA) (2a) Therefore, P Rel - 6 ⁢ ⁢ sides = P 6 ⁢ ⁢ slabs P planar = [ [ 4 ⁢ ⁢ cell x ⁢ t cell ] + 2 ⁢ ( cell x ) 2 ] ⁢ N ⁢ ⁢ α edge 2 Area ( 10 ) But from equation (7a) we know that: N = Area ( cell x + 2 ⁢ ⁢ t isotope ) 2 ( 7 ⁢ a ) So substituting (7a) in (10), we get, P Rel - 6 ⁢ ⁢ sides = ( 4 ⁢ ⁢ t cell ⁢ cell x + 2 ⁢ ( cell x ) 2 ) ⁢ ( Area ) ⁢ α edge 2 Area ⁡ ( cell x + 2 ⁢ ⁢ t isotope ) 2 ( 10 ⁢ a ) Which gives, P Rel - 6 ⁢ ⁢ sides = ( cell x ) 2 ⁢ ( 4 ⁢ t cell cell x + 2 ) ⁢ α edge 2 ( cell x + 2 ⁢ ⁢ t isotope ) 2 And finally, P Rel - 6 ⁢ ⁢ sides = ( 4 ⁢ t cell cell x + 2 ) ⁢ α edge 2 ( 1 + 2 ⁢ ⁢ t isotope cell x ) 2 ( 10 ⁢ ⁢ aa ) If we only consider 5 radioisotope slabs, around the SiC cell (remove the bottom isotope), then the ratio for 5 is given by P Rel - 6 ⁢ ⁢ sides = ( 4 ⁢ t cell cell x + 1 ) ⁢ α edge 2 ( 1 + 2 ⁢ ⁢ t isotope cell x ) 2 ( 11 ) Similarly, for 3 isotope slabs (one on top, two on the sides) the ratio becomes P Rel - 3 ⁢ ⁢ sides = ( 2 ⁢ t cell cell x + 1 ) ⁢ α edge ( 1 + 2 ⁢ ⁢ t isotope cell x ) 2 ( 12 ) The power density of the high volume utilization betavoltaic device is also an importance metric. The equation for the power density of a device with six isotope slabs, for example, is given by: P Density = { C ⁢ { [ 4 ⁢ ⁢ t cell ⁢ cell x ] + [ 2 ⁢ ⁢ ( cell x ) 2 ] } ⁢ S SSA ( t substrate + t cell ) ⁢ Area } ⁢ α edge 2 ⁢ ⁢ Area ( cell x + 2 ⁢ ⁢ t isotope ) 2 Single Junction Ni63 Embodiment of Invention The present invention may have embodiments as a single or multi junction device with either Ni63, tritium, or promethium-147, or other beta emitting isotopes. The following describes an embodiment of the invention which comprises a single junction with Ni63 used as the isotope source. This embodiment is shown in FIG. 3. In this case we have a single P/N junction surrounded by 3 slabs of radioisotopes shown in blue. The isotopes are electrically isolated from the P/N junction by a thin oxide layer (not shown). The N+ region is the SiC substrate. FIG. 4 shows a 3D representation of this embodiment. For clarity, space is inserted between the adjacent radioisotope vertical slabs, where such space would normally be occupied by PN layers. Ohmic contacts are formed in the rear of the device and on the back of the substrate, and these contacts are shown in black. Edge Effects and Design Equations Typically, in designing a betavoltaic device, assumptions can be made regarding beta radiation traveling in a straight line with a density proportional to the specific activity. This is a good approximation for the planar case where the length of the foil is large compared to the absorption length in the SiC. However for the present invention, as one example, for each individual cell, one must take into account the edge effects for each mini cell. For a given beta energy and beta emitter position, the beta emitter will emit betas in all directions (all 360 degrees around). There will be an angle α which defines the edge effects. For angles less than 180 degrees there will be a loss of potential carriers given by α/180. We use the expression αedge in the above equations to represent the edge effects as a dimensionless quantity that takes into account carrier loss. Fabrication of the High Volume Utilization Structure One exemplary method for the fabrication of the high volume utilization betavoltaic invention is as follows: 1—Deep Silicon Carbide Etch: The channels for the vertical radioisotope slabs have to be etched first. This etch depth exposes the entire thickness of the active SiC cell to the radioisotope. 2—Oxide Passivation Thermal oxide will be grown on the SiC to serve as insulation from the shorting of the device junction on the sidewalls of the individual cells. 3—Amorphous Silicon Deposition A layer of amorphous Silicon (a-Si) will be blanket deposited over the deeply etched SiC wafer to allow for the re-planarization of the top surface. 4—CMP Planarization To ensure that lithography can be performed on the patterned surface of the SiC sample after etching, the a-Si deposited on the sample in the previous step has to be planarized. This planarization step provides a flat template for the subsequent photoresist and lithographic processes. 5—Wet Oxide Etch A wet oxide etch is done to remove any residual oxide that might be on the surface of the SiC before the metals for the ohmic contact are deposited. The presence of oxide would compromise the quality of the ohmic contact. 6—Ohmic Contact Metallization The metallization for the formation of ohmic contacts to p-type SiC is selectively deposited on the top surface of the SiC cells. 7—Reactive Ion Etch Removal of a-Si in Trenches The a-Si is removed from the surface of the device by Reactive Ion Etching (RIE) 8—Rapid Thermal Anneal The ohmic contact metallization deposited in step 6 is now annealed using a Rapid Thermal Annealer (RTA). This step forms low resistance contacts to the SiC devices. 9—Frontside Ni Blanket Metallization After the ohmic contacts are formed and annealed, a final blanket Nickel metallization will be done to connect all the individual SiC betavoltaic cells together and to serve as a seed layer for the eventual electroplated Nickel-63 radioisotope layer. 10—Backside Metallization The SiC betavoltaic device is a vertical device and as such may have an ohmic contact on the front and back of the device. This step forms the ohmic contact on the backside of the device. Summary of Some of the Advantages of this Embodiment for Ni63 We can summarize some of the advantages of this invention, as one embodiment: 1. The VUtilization factor for this structure ˜1 because all of the material is either emitting or collecting betas 2. Because of the high volume utilization, the power density will increase 3. This structure can efficiently allow for series combining of junctions to allow for a higher voltage output 4. This structure allows for the deposition of Ni63 by electro chemistry because the “seed” layer for the deposition is at the bottom of the isotope channel and does not “shield” the beta emission. 5. Unwanted beta emissions are easily shielded by the ohmic contacts that may be formed at the bottom of the structure along with, in some embodiments, an additional metal layer deposited on top of the structure. Passivation of the Endfire Surface The advantage of the Endfire betavoltaic concept is the increased area for beta particle input. Therefore, a larger source of energy is available for harvesting, relative to a planar betavoltaic device design. The disadvantage of this approach is that the increase in surface area comes with a potential introduction of surface charges and/or surface traps. Surface charges and/or surface traps can reduce the “effective minority lifetimes” of carriers in the device. The result of these charges is that carrier collection is reduced, which results in lower power output by the device. Surfaces are literal terminations of crystal lattices and the dangling bonds that are formed as a consequence of this termination create localized energy states that can act as generation-recombination centers. These surface states have the potential to reduce the effective minority carrier lifetimes in devices. When the surface-to-volume ratio of a device increases, as is the case with going from a planar to the Endfire betavoltaic design, the total number of surface states increases, which can reduce the power output. To mitigate this surface effect in the Endfire design, a novel metal-oxide-semiconductor (MOS) capacitor will be integrated with the betavoltaic device. The MOS device will be formed on the surface between the SiC device sidewalls, the insulating oxide, and the metal radioisotope source. This MOS capacitor will be biased in accumulation mode. (see FIGS. 11 and 12) The MOS capacitor band diagram shown in FIG. 12(a) illustrates the flat band mode where there is no voltage bias on the metal terminal. This condition is characterized by the absence of band bending in the SiC and by the absence of charge build up at the surface. As a negative charge is introduced to the metal-semiconductor contact (FIG. 12(b)), an electric field is set up across the MOS capacitor. This field attracts the positively charged majority carriers in the p-type SiC to the surface where they quickly accumulate. This particular condition is called the accumulation mode. In the accumulation mode, the majority carrier density is increased at the surface and electric fields are produced which act to repel minority carriers from the surface. The action of the electric field on the minority carriers have the effect of isolating them from the traps. This electric field isolation allows for the Endfire design to be less susceptible to the effects of surface traps. Biasing the MOS Capacitor: The integrated MOS capacitor can be biased into accumulation by several sources including, but not limited to, the Endfire betavoltaic's generated voltage and the voltage from fixed oxide charges introduced during the fabrication of the devices. Since the SiC Endfire betavoltaic will produce an open circuit voltage of 2 Volts, a portion of this voltage can be used to bias the MOS capacitor on the sidewalls. Fixed negative charge can also be implanted into the oxide to permanently bias the MOS capacitor into accumulation. The fixed negative charge will allow the device to remain in accumulation, regardless of the external resistive loads that the device may be connected to and will also simplify the fabrication process of the device, by eliminating the need to connect the negative output of the betavoltaic to the MOS terminal. Alternate Embodiment of the Endfire Design The Endfire betavoltaic concept can be implemented in different p-n junction configurations. An alternate configuration is shown in FIG. 9. Rather than just being a simple mini p-n junction slab (as the embodiment shown in FIG. 10), there are two back to back p-n junctions in parallel, built into the device, and both harvest beta energy to contribute to the total power output. The structure can be n+-p−-n+ (as shown in FIG. 10), or the mirror structure of p+-n−-p+. The advantages of this embodiment of the device are as follows: For the n+-p−-n+ structure, the minority carrier lifetimes are larger in p-type material The maximum depth of the device can be increased The total power output is higher Surface passivation is easier to achieve In summary, we have the following figures: FIG. 1 shows schematic of beta voltaic converter, corresponding to FIG. 5. FIGS. 2a-c show: Schematic illustration of one embodiment of the invention, corresponding to FIGS. 6a-c. The drawing shows a slap converter geometry being replaced by a number of cube-based converters. FIG. 3 shows: Schematic of a beta voltaic device embodiment, corresponding to FIG. 7. FIG. 4 shows a 3D representation, corresponding to FIG. 8. For clarity, space is inserted between the isotope vertical slabs. Ohmic contacts are formed in the rear of the device and on the devices bottom side. FIG. 5 shows schematic of beta voltaic converter: green region is the SiC power converter, the blue region is the radio isotope, while the black regions are the ohmic contacts. FIGS. 6a-c show: Schematic illustration of one embodiment of the invention. The drawing shows a slap converter geometry being replaced by a number of cube-based converters. FIG. 7 shows: Schematic of a beta voltaic device embodiment: Green region is the SiC power converter, the blue region is the radio isotope, while the black regions are the ohmic contacts. FIG. 8 shows a 3D representation. For clarity, space is inserted between the isotope vertical slabs. Ohmic contacts are formed in the rear of the device and on the devices bottom side and these contacts are shown in black. FIG. 9 shows the diagram of n+-p−-n+ embodiment of the Endfire structure. FIG. 10 shows drawing for n-p-n Comb Endfire device. FIG. 11 shows: MOS capacitor formed on sidewall of the Endfire Betavoltaic device. FIG. 12 shows: P-type MOS capacitor (a) with Vg=0, biased in the flatband mode (b) with Vg<0, biased in the accumulation mode. Maximizing Charge Collection in SiC Betavoltaics—Influence of Junction Depth This is also addressed in our co-pending applications, mentioned above: To quantify the extent of the surface, it is necessary to know the penetration depth, or range, RB in μm, of the beta electron in the semiconductor, which is given as:RB (μm)=[4×E01.75 (keV)/100]/ρ(g/cm3) (1e),where E0 is the incident beta energy in keV, and ρ is the density of the semiconductor in g/cm3. The penetration depth is simply a function of the energy spectrum of the β-radiation, which is known. The spectrum, to first order, is given byf(E0)=K√{square root over (E02+2mc2E0)}(E0(max)−E0)2 (2e)where f(E) is the energy distribution function, m the electronic mass, c the speed of light, and K a normalization constant, such that we have: ∫ 0 E 0 ⁡ ( max ) ⁢ f ⁡ ( E 0 ) ⁢ ⅆ E 0 = 1 ( 3 ⁢ e ) The energy extends to a maximum, E0(max), that typically lies at ˜3 times the mean energy. For a given beta emitting isotope, a single E0(max) completely specifies the spectrum, as eq. 2e indicates. There is a Coulombic penetration factor that modifies equation (2e) above. This factor accounts for electrons being retarded by the Coulombic attraction from the nucleus, which skews the spectrum towards lower energies. Considering this factor, equation (2e) becomes:f(E0)=KF(ZD,E0)√{square root over (E02+2mc2E)}(E0(max)−E0)2 (4e)where F(ZD,E0), called the Fermi function, takes into account the Coulombic penetration effects. This function is tabulated in relevant semiconductor literature, and is related to the daughter nucleus atomic number, ZD, and the energy of the emitted β particle, E0. It can be approximated by: F ⁡ ( Z D , E 0 ) = 2 ⁢ ⁢ π ⁢ ⁢ v 1 - exp ⁡ ( - 2 ⁢ ⁢ π ⁢ ⁢ v ) ⁢ ⁢ where ⁢ ⁢ v = 1.16 × 10 - 3 ⁢ Z D / E 0 2 + 2 ⁢ ⁢ mc 2 ⁢ E 0 m 2 ⁢ c 4 + E 0 2 + 2 ⁢ ⁢ mc 2 ⁢ E 0 ( 5 ⁢ e ) The penetration depth is then estimated as described in equation (1e). From (4e), ˜65% of the spectrum energy lies at or below the mean, 5.5 keV for Tritium, while >80% of the energy lies below E(max)/2, which is ˜9 keV for Tritium. Assuming that all the beta-generated electron-holes beyond the surface junction p-type layer are collected, while none of those generated in the surface junction layer are collected, we can estimate the charge collection as a function of energy, or as simply the fraction of the total path length (RB) that lies beyond the junction region (Xj). This fraction at each energy in the beta spectrum is (RB−Xj)/RB. Integrating the total charge collection function, we obtain the total charge collection efficiency. More details and results are given in our co-pending applications, mentioned above, which are incorporated by reference here. Any variations of the teachings above are also meant to be covered and protected by this current application.
	claims	1. A fuel assembly comprising:a fuel rod array of a plurality of fuel rods including a plurality of first fuel rods including enriched uranium and not including a burnable poison, and a plurality of second fuel rods including said enriched uranium and said burnable poison; andat least one water rod;wherein said fuel rod array has a plurality of corner sections, each corner section being a region having three fuel rods in an outermost layer of said fuel rod array including one fuel rod at a corner of said outermost layer of said corner section and two fuel rods in said outermost layer directly adjacent said one fuel rod at said corner;wherein one of said second fuel rods is said one fuel rod at said corner of a first corner section of said plurality of corner sections;wherein another of said second fuel rods is at least one of said two fuel rods directly adjacent to said one of said second fuel rods at said corner of said first corner section;wherein others of said second fuel rods are placed adjacent to said at least one water rod;wherein a second corner section of said plurality of corner sections includes a fuel rod other than one of said second fuel rods at a corner in said outermost layer that faces a control rod when said fuel assembly is loaded in a core; andwherein the number of said second fuel rods placed in said second corner section is less than the number of said second fuel rods placed in said first corner section. 2. The fuel assembly according to claim 1, wherein said at least one water rod is a large-diameter water rod having a lateral cross section that occupies an area where at least two fuel rods are placeable. 3. The fuel assembly according to claim 1, wherein, in said second corner section, said second fuel rods are placed in a layer adjacent to said outermost layer of said fuel rod array. 4. The fuel assembly according to claim 1, wherein said plurality of corner sections includes two first corner sections. 5. The fuel assembly according to claim 1, wherein said fuel rod array has a substantially rectangular configuration in cross section and having four corner sections. 6. The fuel assembly according to claim 1, wherein when an active fuel length of said fuel assembly is denoted by L and a distance from a lower end of said active fuel length L in an axial direction of said fuel assembly is denoted by h, burnable poison regions including said burnable poison in said second fuel rods are located in a region defined by L/8≦h≦L/2. 7. The fuel assembly according to claim 1, wherein when an active fuel length of said fuel assembly is denoted by L and a distance from a lower end of said active fuel length L in an axial direction of said fuel assembly is denoted by h, and further when said active fuel length is divided into an upper region and a lower region at a position defined by h=L/2, said adjacent second fuel rods in said outermost layer include said burnable poison in most of said lower region and do not include said burnable poison in most of said upper region. 8. The fuel assembly according to claim 1, wherein an average enrichment in a lateral cross section of said fuel assembly is within a range from 3.7 wt % to 10.0 wt %.
	summary
	summary
048045144	description	DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS Reference will now be made in detail to the presently preferred embodiments of the invention, examples of which are illustrated in the accompanying drawings. FIG. 1 shows the basic structure of a first embodiment of a neutron dosimeter according to the present invention. Numeral 10 designates means exposed to neutron flux for generating fission fragments at a rate proportional to the intensity. Adjacent fission fragment generating means 10 is light generating means 20. Light generating means 20 is a material which produces light, or scintillates, when traversed by a fission fragment. Optically coupled to light signal generating means 20 is electrical signal generating means 30. In the specific embodiment illustrated in FIG. 1, these means take the form of a photomultiplier tube attached to light signal generating means 20. It will be understood by one of ordinary skill in the art, however, that it is not necessary that the light signal generating means and electrical signal generating means be physically proximate, and is sufficient if they are merely optically coupled by light pipes or similar apparatus for conveying light. The electrical signal generating means are electrically coupled to means for processing electrical signals. These means may be any means for taking the raw data represented by the electrical signal from electrical signal generating means 30 and processing it into a form which reveals the measured intensity of neutron flux impinging on the fission fragment generating means. Thus, signal processing means 40 may include front-end electronics such as filters, triggers, and logical operation circuits. Signal processing means may also include a computer for manipulating, storing, and retrieving the data from the front-end electronics. Fission fragment generating means 10 in the embodiment of FIG. 1 takes the form of a fissile layer, light signal generating means 20 the form of a scintillator layer, electrical signal generating means 30 the form of a photomultiplier tube, and signal processing means 40 the form of signal processing electronics and a computer. As mentioned, a light pipe may be inserted between the scintillator and photomultiplier. The operation of such a device is straightforward as follows. The capture of neutrons by the fissile layer produces fission fragments. The scintillator layer is arranged in conjunction with the fissile layer so that the scintillator layer is traversed by a known fraction of the produced fission fragments. (In this context and in the claims, the known fraction may be any value up to and including unity). This known fraction of fragments produces light pulses in the scintillator which are detected by the photomultiplier. The electronic pulse response of the photomultiplier to the light pulses is then processed by the signal processing means. If the signal processing means includes a computer, the computer stores the data and retrieves it as quickly as necessary. Considerations in the selection of material for the fission fragment generating means 10 are familiar to one of ordinary skill in the art. Many types of neutron dosimeters commonly employ a layer of fissile material. The neutron capture results in fission fragment ions with energies approaching 100 MeV. These energetic ions produce tracks in some dosimeters and in others they are used to ionize a gas. In the present invention, the ionizing power of these energetic ions is employed. Neutron spectral information can be obtained by using various fissile isotopes, having differing neutron energy thresholds for fission. For example, .sup.235 U and .sup.239 Pu can be made to fission by neutrons of energies from thermal (approximately 0.1 eV) through fast (average energy approximately 1 MeV). The fission cross sections of .sup.238 U and .sup.237 Np, on the other hand, show distinct fission threshold versus neutron energy characteristics (approximately 0.8 MeV for .sup.237 Np and approximately 1.5 MeV for .sup.238 U). In fact, these "fast-fission" isotopes are currently in use in neutron dosimetry measurements. Further flexibility in determining threshold energies can be achieved through the use of cadmium coatings. These coatings show a near complete neutron absorption below 0.41 eV, and so can supply a basis for inference of thermal neutron flux by comparison of data from coated fissile layers and uncoated fissile layers. As an example, consider .sup.238 U, which has the lowest spectrum-weighted fission cross section (0.3 barns) relative to a 1.5 MeV threshold flux, and thus provides the most conservative (smallest) number of fission fragments. If the dosimeter were to be used in a reactor cavity to obtain information on neutron flux at various locations outside of the thermal shield, the fast neutron flux at these locations is conceivably as low as 10.sup.7 n/cm.sup.2 /sec. The reaction rate at this location is thus 3.times.10.sup.-18 reactions/atom/sec for .sup.238 U. The thicker that a 1 cm.sup.2 layer of this material is, the more reactions can be produced and, hence, the more detector operation can be enhanced up to some maximum useful thickness. The maximum useful thickness can be determined by the range of fission fragments in the fissile material. For example, continuing with .sup.238 U as an example, the ranges of 100 MeV fission fragments for such a material lie between 5 and 9 microns. Thus, a maximum useful thickness would be on the order of 5 microns. A 1 cm.sup.2 of .sup.238 U, 5 microns thick, contains 2.4.times.10.sup.19 atoms. After a day at the reaction rate assumed above, there would be produced approximately 6.times.10.sup.6 reactions. After a month and a year, between 2.times.10.sup.8 and 2.3.times.10.sup.9 reactions/cm.sup.2 would have accrued. As for the scintillator layer, every form of radiation will produce light pulses in known scintillators. The basic problem, therefore, is finding ways of discriminating pulses produced by fission fragments from pulses produced by other sources. This task is very familiar to one of ordinary skill in the art, and many such means and ways have been developed for coping with this problem. One method of discriminating pulses is to take advantage of the fact that every fission fragment entering a scintillator will produce a response and that these fragments will have a well defined and short range in the materials. The maximum range of 100 MeV fragments in ZnS, CsI and NaI will be about 20 microns. If the scintillator were made of one of these materials, restricting its thickness to, for example, 25 microns means that it will be possible to produce all possible fragment produced light. Neutrons and gamma rays would have a very low probability of interaction in such a thin crystal. Further, electrons (beta rays or Compton electrons) above approximately 55 keV have a range greater than 25 microns. Thus, for gamma rays and beta rays it would be anticipated that maximum light output would correspond to 55 keV. It is also possible to take advantage of the fact that light output is a strong function of the energy deposition rate, dE/dx, of the incident radiation. For electrons with MeV energies (typical of most beta rays and Compton electrons) the dE/dx is approximately 10.sup.-1 to 10.sup.-2 MeV/mg/cm.sup.2. The neutrons will produce light pulses via the recoiling scintillator atoms with maximum energies less than 0.5 MeV. The maximum dE/dx for these will be less than 1 MeV/mg/cm.sup.2. By contrast, the 100 Mev fission fragments will have dE/dx values on the order of 30 MeV/mg/cm.sup.2. Thus, the intensity of the light pulses they produce will be significantly greater than those produced by other forms of radiation. As a result, the small height background radiation electron pulses from the photomultiplier can be easily discriminated by setting a suitable threshold on a single channel analyzer. The above approaches should suffice to eliminate background radiation as a source of spurious signals. It is possible, however, to employ additional or alternative techniques. For example, it is possible to discriminate between light pulses produced by heavy, energetic fission fragments and those produced by other forms of radiation by detecting how long it takes the pulses to decay. The light produced by heavy, energetic fission fragments should take longer to decay. Considerations for selecting a specific scintillator material are known to one of ordinary skill in the art. Virtually any scintillator can be used, but the scintillator materials mentioned above, zinc sulphide (ZnS), sodium iodide (NaI) and cesium iodide (CsI) have the advantage of being commercially available and also being fairly well understood, particularly with respect to their utilization as fission fragment detectors. CsI is non-hydroscopic, and, therefore, presents less of a problem in packaging and handling than does NaI. NaI, on the other hand, has a greater scintillation efficiency and a longer scintillator decay time than CsI. ZnS has an excellent efficiency and high light output for particle detection, but because it is polycrystalline it tends toward opacity for thicknesses greater than or equal to approximately 25 mg/cm.sup.2. The particular design parameters of a given application will in most instances suggest which material is the best choice for the scintillator. The light or scintillator pulse is converted to electron pulses by electrical signal generating means 30, in the presently preferred embodiment, a photomultiplier tube. The selection of the photomultiplier is not critical because the scintillator decay times of ZnS, NaI and CsI crystals, approximately 0.1 microseconds, 0.23 microseconds and 0.7 microseconds, respectively, are so long that speed is not crucial. Any commercially available tube commonly used with either of these materials should suffice. The long pulse decay times, however, make it important that the reaction rates produced by the fissile layer be selected to be sufficiently low that the pulse rate can be handled comfortably by the photomultiplier. Thus, reaction rates of interest are those for which the probability of pulse pileup is low. For practical applications the upper limit would be approximately 5.times.10.sup.5 pulses/sec. It can be anticipated that at certain locations neutron recoils and gamma rays could produce scintillator pulses in excess of this limit. For example, near the primary concrete shield of a nuclear reactor, the neutron flux could be approximately 5.4.times.10.sup.13 neutrons/cm.sup.2 /hr, and the gamma ray flux could be as high as 9.times.10.sup.12 photons/cm.sup.2 /hr. The neutron cross sections and gamma ray attenuation coefficients enable calculation of the pulse rates produced by recoils and background gamma rays. Both quantities are in excess of 10.sup.6 pulses/sec. This will, however, not pose a problem because the energy deposited per pulse by these forms of background radiation is insignificant compared to the fission-fragment produced pulses. Thus, even if there is pulse pileup the effect will be minimal as long as the fission fragment pulse rate remains below the maximum level. The electronics for the remainder of the system could typically include standard pulse processing systems commonly employed in nuclear spectroscopy. The signal pulses can be conditioned by amplification and discrimination, and then stored. In high radiation areas, it may be necessary to shield the electronics and wiring against induced currents due to background. An alternate approach could be to use bunched optical fibers as a light conduit between the scintillator and phototube. For long term use such a conduit would have to replaced periodically when the color centers produced by radiation attenuate the light signal too severely. The computer can, of course, be programmed in a fashion clear to one of ordinary skill in the art to display graphics showing hourly, daily, or yearly variations in fluence. The shifts in neutron energies with time obtained from a set of detectors having different fissile layers can also be displayed in a similar fashion. FIG. 2 is an example of a system in which two detector assembles, each including a fissile layer, scintillator layer, and photomultiplier tube, are used in order to obtain spectral data on the neutron flux. The material comprising the second fissile layer 50 are selected to have different minimum threshold energies for fission. This is accomplished by judicious choice of materials or coatings as suggested above. For example, assume that the energy for fission of the material of first fissile layer 10 is less than the energy for fission of the material comprising the second fissile layer 50. The difference in flux measured by the first scintillator layer 20 and first photomultiplier tube 30 as opposed to that measured by the second scintillator layer 55 and second photomultiplier tube 57 would then give a measure of or a basis for inferring the rate of flux of neutrons having energies between the first energy of fission and the second energy of fission. Obviously, an array of several detector assemblies can be devised for obtaining a calculation and measurement for various energy ranges. An actual neutron detector for dosimetry applications has been described above which is based on the use of a thin film of fissionable material placed next to a thin scintillator (ZnS, NaI or CsI) which emits flashes of light which can be identified as originating from the passage of neutron-induced fission fragments through the scintillator. For some applications, a complication may arise due to the emission of alpha particles by the fissionable materials. For high count rates, fission events may be difficult to distinguish from randomly summed alpha pulses. For example, if a .sup.238 U foil is placed in the reactor cavity of an operating power reactor, there will be a fission rate of about 3.times.10.sup.-18 reactions/atom/sec. The effective maximum thickness (asymptotic sensitivity) of a uranium foil is 1.098.times.10.sup.19 atoms/cm.sup.2 for fission fragments, so that the fission rate for a thick foil would be EQU (3.times.10.sup.-18 fissions/sec/atom)(1.098.times.10.sup.19 atoms/cm.sup.2)=0.3 pulse/cm.sup.2 -sec The alpha decay rate for the same foil (assuming an effective thickness for alpha emission of 5 mg/cm.sup.2) is EQU (1.24.times.10.sup.4 dps/g)(5.times.10.sup.-3 g/cm.sup.2)=62 dps/cm.sup.2 Since the alpha emission rate is only about 200 times the fission fragment rate for this foil, adequate discrimination between fission fragments pulses and alpha particle pulses should be possible. In the cases of .sup.235 U and .sup.239 Pu fissionable layers, the alpha specific activity increases dramatically, but the fission rates also increase so alpha pulse pile-up should still not be a problem. In the case of .sup.237 Np, however, the fission rate is only about four times the fission rate for .sup.238 U or 1.2 fissions/cm.sup.2 -sec. The alpha decay rate (again assuming an effective thickness of 5 mg/cm.sup.2) is EQU (2.61.times.10.sup.7 dps/g)(5.times.10.sup.-3 g/cm.sup.2)=1.3 10.sup.5 dps/cm.sup.2 Since the alpha pulse rate will be about 10.sup.5 times higher than the fission fragment pulse rate for .sup.237 Np, pulse pile-up may well be a problem. This is significant because .sup.237 Np is a key isotope for neutron dosimetry applications due to its favorable neutron energy response function. The embodiment of FIG. 3 is intended to alleviate this problem. It comprises an arrangement of a fissionable layer sandwiched between two optically isolated thin scintillators. As seen in FIG. 3, the thin fissionable layer 60 is placed between a first scintillator layer 70 and a second scintillator layer 80. Each scintillator layer has its own photomultiplier, designated respectively by 90 and 100. These photomultipliers will be referred to as the left-hand photomultiplier 90 and the right-hand photomultiplier 100, respectively. Pulses output by left-hand photomultiplier 90 are amplified by a suitable amplifier 110 while similarly the pulses output by right-hand photomultiplier are amplified by a right-hand amplifier 120. The pulses amplified by left-hand amplifier 110 are counted by a first scalar 130 while the pulses amplified by right-hand amplifier 120 are counted by a second scalar 140. In addition, the pulses amplified by right-hand amplifier 110 and left-hand amplifier 120 are checked for coincidence and summed if a coincidence exists by fast coincidence and summing circuit 150. A third scalar 160 accumulates the events summed by fast coincidence and summing circuit 150. The apparatus of FIG. 3 would operate as follows. Neutron-induced fission events in the .sup.237 Np foil result in simultaneous emission of fission fragments in opposite directions, so that scintillations would be produced in both scintillators (which may be made of ZnS, NaI, or CsI) simultaneously. Third scalar 160 counts only coincident pulses resulting from either fission events (true coincidences) or random coincidences (accidental coincidences). If the coincidence fission pulses are summed, the resulting pulse would be much larger than any possible random coincidence (except for random fission coincidences which would be unlikely for low fission rates), since approximately 200 MeV is liberated in the fission process and only about 10 MeV can result from random coincidences of alpha particles. The arrangement of FIG. 3 has the advantage that fission events are identified uniquely. In addition, all alpha background is suppressed, as well as beta, gamma, and neutron-induced recoil backgrounds. The arrangement has the additional advantage that the source of random noise, such as photomultiplier dark current, preamplifier noise, amplifier noise, and cable noise, are also suppressed. These advantages can be extremely valuable if the detector is to be used in harsh reactor environments where background radiation and electronic noise almost always present problems. Long-term stability of the system can be verified by requiring that the count rates of the two scintillators (the totals in the first scalar and second scalar) be constant and equal. Any malfunctions due to loss of fissionable material, electronic noise, or other sources would be immediately apparent as this test would not be passed in those events. The primary design requirement for the system is that the fissionable isotope layer be thin enough so that both fission fragments have a high probability of escape and detection. Foils thinner than about 100 .mu.g/cm.sup.2 fulfill this requirement. Foils of this thickness are also thick enough to ensure that the two scintillators are optically isolated. FIG. 4 shows a nuclear reactor incorporating one embodiment of a dosimeter according to the present invention. Numeral 10 again designates means for producing fission fragments, which are again in the form of a fissile layer. Numeral 20 designates a scintillator layer. Numeral 170 designates means for conveying light generated in scintillator 20 to photomultiplier 30. As mentioned above, light conveying means 170 may be any suitable means such as bunched optical fibers or a light pipe. Numeral 180 designates a reactor core, comprising a matrix of fuel rods 190. The reactor core is surrounded sequentially by a baffle 200, a core barrel 210, and a thermal shield 220. The fissile layer 10 is positioned within the reactor vessel 230 but outside the thermal shield 220. The foregoing description has been in terms of a preferred embodiment merely for the purposes of illustrating the underlying principles of the invention. Nothing in the foregoing should be construed as limiting the invention to the specific embodiments discussed. Quite the contrary, it will be readily apparent to one of ordinary skill in the art that the concepts underlying the particular embodiments discussed herein have extremely broad application. The invention should therefore not be regarded as being limited to any of these specific embodiments, but instead should be regarded as being fully commensurate in scope with the underlying concept, as reflected in the following claims.
062815083	description	DETAILED DESCRIPTION FIGS. 3a-3b show a microlens component 51 according to the present invention. Microlens component 51 is, for example, a 500 .mu.m thick, 7 mm.times.7 mm silicon chip. At the center of microlens component 51 is a 1 to 1.5 .mu.m thick, 1 mm.times.1 mm membrane window 53, at the center of which is a 2.5 .mu.m diameter aperture 55. (These dimensions are merely illustrative.) FIG. 3b is a side view of the FIG. 3a structure. Aperture 55 must be precisely aligned with the apertures of the other microlens components (not shown) when the microlens is assembled. Rather than utilizing the painstaking prior art process of manually aligning the apertures using a microscope to observe the alignment, the present process utilizes a structure such as a standard optical fiber 59 to align the multiple layers. First, alignment openings 57a-57b are formed in the microlens component 51. Because microlens component 51 is of silicon, conventional silicon processing techniques may be utilized to form the openings 57a-57b through the microlens component 51. Such techniques are well known in the art and enable the etching of holes in silicon to very precise tolerances. Optical lithography can be used to first pattern the holes in the silicon, followed by a silicon etching process to etch the holes through the microlens component 51. Techniques such as electron-cyclotron-resonance (ECR) etching, active silicon ion etching, reactive ion etching (RIE), inductively-coupled-plasma (ICP) etching, or any of the known methods for etching silicon may be used to quickly, reproducibly, and precisely form alignment openings 57a-57b. These techniques can be used to etch silicon with tolerances in the nanometer range, thereby allowing the apertures 55 to be positioned accurately with respect to the alignment openings 57a-57b. For improved efficiency, this etching step may be carried out in conjunction with and using the same processes as the etching steps required for forming window 53 and aperture 55. After the alignment openings 57a-57b are formed, aligners 59a-59b are inserted through the openings 57a-57b. In one embodiment, aligners 59a-59b are short lengths of standard optical fiber, which are circular dielectric waveguides typically used to transport optical energy and information. These commercially available fibers are made of doped silica, possibly coated with several layers of cushioning material, such as acrylate. One suitable fiber material is commonly known as Pyrex. Pyrex optical fibers are commercially available from the Newport Corporation. This is just one example of the materials that may be used as the aligner. It is only necessary that the aligners be sufficiently strong and stiff to prevent shearing or bending of the assembled microlens, as described below. In addition, if the aligners are not removed from the microlens after assembly, typically they must be electrically nonconductive as well. The optical characteristics of the fibers are of no importance; optical fibers are utilized in this embodiment because they are relatively inexpensive, readily available, nonconductive, and are formed with very tight dimensional tolerances. Also, over the short lengths needed, they are sufficiently rigid. FIGS. 4a-4d illustrate in side views the assembly of a microlens according to one embodiment of the present invention using the FIG. 3a, 3b structures. FIG. 4a shows a side view of microlens component 51 after formation of window 53, aperture 55a, and alignment openings 57a-57b. Aligners 59a-59b are inserted into alignment openings 57a-57b, respectively, of microlens component 51, as shown in FIG. 4b. Next, an insulating spacer 61 is attached to the assembly by threading aligners 59a-59b through alignment openings 57c-57d in the spacer 61, and positioning spacer 61 atop microlens component 51. Spacer 61 is provided with large aperture 63, which must be aligned so as not to block aperture 55 in microlens component 51. Because the purpose of spacer 61 is to provide separation and insulation between the electrode layers of the microlens, aperture 63 can be made quite large and is not particularly difficult to properly align. As can be seen in FIG. 4d, successive microlens components 65 and 69 each define small apertures 55b-55c which are to be precisely aligned with aperture 55a of the base microlens component 51. Microlens components 65 and 69 include alignment openings 57e-57f through which aligners 59a-59b are threaded. Because aligners 59a-59b are sufficiently rigid, when the elements of the microlens are assembled as shown in FIG. 4d, the layers are securely held relative to each other so that the alignment openings 57a-57f of each layer are precisely aligned. By accurately etching the alignment openings 57a-57f in relation to the apertures 55a-55c in each layer, the apertures 55a-55c will all also align correctly. In the embodiments shown in FIGS. 3a-3b and 4, two aligners 59a-59b are used in the assembly of the microlens. The invention is not limited to only two aligners; it is also possible to assemble a microlens using a greater number of aligners or just one. When using only one aligner fiber, another structure may be used to better stabilize the microlens and prevent rotation of the microlenses relative to each other. For example, one edge of each microlens component may be aligned with another structure to prevent misalignment of the apertures caused by relative rotation of the layers. FIG. 5 shows a side view of a completed microcolumn according to the present invention. The microcolumn is formed of layers of microlens elements and other components 51a-51i alternating with layers of insulating spacers 61a-61h, with aligners 59a-59b serving to keep the components properly aligned. At the top is conventional tip assembly 77, which may include, for example, a scanning tunneling microscope mounted in microlens 51i. Beneath the tip assembly 77 is the source lens 79 of the microcolumn, followed by the dual silicon deflector 81 and the Einzel lens 83. After the microcolumn structure is complete, the layers may be bonded together to provide increased strength and stability. This is accomplished anodically as described above, e.g., by connecting the assembled structure to a voltage source and applying a potential across the layers of the microcolumn under elevated temperatures. By applying both a positive followed by a negative potential across the alternating glass-silicon layers, the individual layers of the microcolumn are anodically bonded. This may also result in the bonding of the optical fiber aligners 59 to the microlenses 51. Alternatively, the layers are laser bonded. Aligners 59 are used to maintain the precise alignment of the assembly while the laser spot welding is carried out, and may be removed, if desired, after the layers are bonded together. However, because the aligners 59 are sufficiently small relative to the overall surface area of the microlens components 51 and usually are not located too close to the aperture 55 of the microlens components 51, they should not interfere with the operation of the microcolumn and do not have to be removed after assembly. Excess length of aligner 59 is clipped off the microcolumn, or left in place to allow for additional structures to be added later. The accuracy of the alignment of the microlens components is dependent on the accuracy of the aligners 59 and the precision of the alignment openings 57. Because the aligners 59 are used to precisely align the apertures 55 of the microlenses, it is advantageous that the aligners 59 be precisely fitted into the alignment openings 57. This is balanced with the desire to increase the efficiency of the assembly process, however. Micromachining the alignment openings 57 to exactly the diameter of the fiber aligner 59 may provide accurate alignment of the apertures 55, but the threading of the aligner 59 through the equally-sized alignment opening 57 may be problematic. The alignment openings 57 are larger than the diameter of the aligners 59 to ease threading. Two methods of addressing this problem are illustrated in FIGS. 6a-6b and 7. FIGS. 6a-6b each illustrate in perspective and plan views an embodiment in which the alignment opening 100 of microlens component 104 includes two eccentrically formed holes, threading portion 101 and locking portion 103. Threading portion 101 has a diameter slightly larger than the diameter of optical fiber aligner 102. Locking portion 103 has a diameter closely approximating that of aligner 102. In one embodiment, aligner 102 has a cross-sectional diameter of 250 .mu.m, threading portion 101 has a diameter of 300-350 .mu.m, and locking portion 103 has a diameter of 250-260 .mu.m, depending on the tolerance requirements for the apertures 55 of the microlenses. FIG. 6a shows the first step in which aligner 102 is threaded through the threading portion 101 of the alignment opening 100. Because of the increased diameter of threading portion 101, threading the aligner 102 through the alignment opening 100 is easily accomplished. After the aligner 102 is inserted, it is then moved into the locking portion 103 of the alignment opening 100 in the direction of arrow 105. FIG. 6b shows aligner 102 fixed in locking portion 103. Precise etching of the location of locking portion 103 with reference to the location of aperture 55 (not shown in FIGS. 6a-6b) in each microlens component ensures that all layers will align properly. FIG. 7 shows in a plan view a different method of providing for accurate alignment of the components. Here, alignment openings 107a-107b are formed in the shape of a square, each side of the square being slightly longer than the diameter of the associated aligner 111a-111b. After aligners 111a-111b are threaded through the alignment openings 107a-107b, they are pushed towards each other in the direction of arrows 105a-105b. A manipulator located beneath the bottom layer of the microlens is used to force the protruding ends of the aligners 111a-111b together. When the aligner 111a is pushed towards an edge of the alignment opening 107a, the aligner 111a contacts the alignment opening at two contact points 109a-109b, which are located along two adjacent sides of the square alignment opening 107a. These two contact points 109a-109b provide the reference location for properly aligning the apertures 55 of the microlenses (not shown in FIG. 7). Because the optical fiber aligners 111a-111b are relatively short, they are sufficiently stiff to provide proper alignment throughout the length of the microcolumn. Because the aligner 111a is pressed against the two contact points 109a-109b, the size of the alignment opening 107a does not affect the precision of the alignment, provided that the two contact points 109a-109b are properly placed. In this case, the aligners 111a-111b are locked against the two edges of the square alignment openings 107a-107b and the alignment accuracy is defined by the variation in the diameters of the fiber aligners 111a-111b, the reproducibility and positioning of the square alignment openings 107a-107b, and the perpendicularity of the two edges which define the placement accuracy. It is not necessary that the alignment openings 107a-107b have perfectly square shapes, or that they be located at opposite corners of the microlens component 108. It is additionally not required that the aligners 111a-111b be forced inwards toward the center of the microlens 108. It is only necessary that the aligners 111a-111b be able to precisely align the multiple components by contacting an edge of the alignment openings 107a-107b to provide an identifiable reference location. For example, the two alignment openings 107a-107b can both be formed in the upper half of the microlens component 108 shown in FIG. 7, and can be formed as rectangles, triangles, or any other size polygon, so long as the contact points 109 can be accurately identified and positioned. Furthermore, the aligners 111a-111b may, for example, be arranged so that they are forced in a direction away from each other, so long as the forces they apply against the microlens component 108 complement each other in such as way as to maintain a stable alignment. Although the invention has been described with reference to particular embodiments, the description is only an example of the invention's application and should not be taken as a limitation. Various other adaptations and combinations of features of the embodiments disclosed are within the scope of the invention as defined by the following claims.
	claims	1. A drilling instrument for machining tubes in tube sheets of heat exchangers in a radioactive environment, comprising:a support device;a transport device having retaining elements; anda drilling device having retaining fingers and a resting plate;wherein the transport device, and the drilling device are located proximate to each other and are each connected to the support device, such that the retaining elements and the retaining fingers are arranged on a common first side of the transport device and of the drilling device,wherein the support device has a support plate and wherein the resting plate is located proximate to and supported by the support plate,wherein the support plate is connected to the resting plate by at least one connecting element that is movable between a first position and a second position,wherein the resting plate is fixed in position relative to the support plate when the at least one connecting element is in the first position, andwherein the resting plate is movable relative to the support plate when the at least one connecting element is in the second position. 2. The drilling instrument as claimed in claim 1, wherein the transport device has at least four retaining elements on the first side which are subdivided into two groups, wherein each group is controllable separately, wherein a first group of the two groups is capable of pivoting with respect to the second group, and wherein at least one of the two groups is displaceable with a linear movement. 3. The drilling instrument as claimed in claim 2, wherein the drilling device has at least two retaining fingers, and wherein the retaining elements of the transport device and the at least two retaining fingers are movable in a direction perpendicular to the first side. 4. The drilling instrument as claimed in claim 1, wherein the drilling device has at least two retaining fingers, and wherein the retaining elements of the transport device and the at least two retaining fingers are movable in a direction perpendicular to the first side. 5. The drilling instrument as claimed in claim 4, wherein the support plate or the resting plate has at least one limiting element which is arranged in a recess, and in that the shape of the recess allows play in the second position of the connecting element. 6. The drilling instrument as claimed in claim 5, wherein the support plate or the resting plate has at least one approximately conical or frustoconical centering element that is configured to cause the support plate and the resting plate to be arranged in a predefined position with respect to one another when the connecting element is positioned in the first position. 7. The drilling instrument as claimed in claim 4, wherein the support plate or the resting plate has at least one approximately conical or frustoconical centering element that is configured to cause the support plate and the resting plate arranged in a predefined position with respect to one another when the connecting element is positioned in the first position. 8. The drilling instrument as claimed in claim 7, wherein the connecting element has a pneumatic or hydraulic drive device that is configured to cause the connecting element to be movable selectively into the first position or the second position. 9. The drilling instrument as claimed in claim 1, wherein the connecting element has a pneumatic or hydraulic drive device that is configured to cause the connecting element to be movable selectively into the first position or the second position. 10. The drilling instrument as claimed in claim 9, wherein the drive device contains a biasing spring that is configured to cause the resting plate to be held with the support plate in the second position when the drive device is switched off. 11. The drilling instrument as claimed in claim 9, wherein the drive device has a hydraulic or pneumatic drive mechanism configured to apply an opposing force to that of the biasing spring such that the connecting element is movable into the first position by a predefined distance perpendicular to the first side. 12. The drilling instrument as claimed in claim 11, wherein a peg or pin is arranged on the support plate and is operable to apply a lifting force to the resting plate such that the resting plate is capable of being moved a predefined distance in the direction of the second position. 13. The drilling instrument as claimed in claim 9, wherein a peg or pin is arranged on the support plate and is operable to apply a lifting force to the resting plate such that the resting plate is capable of being moved a predefined distance in the direction of the second position. 14. The drilling instrument as claimed in claim 13, wherein the peg or pin applies the lifting force to the resting plate in a location on the surface of the resting plate that is colinear with the center of gravity of the drilling device. 15. The drilling instrument as claimed in claim 14, wherein the lifting force is no more than 10% higher than is necessary to offset the opposing weight force of the drilling device. 16. The drilling instrument as claimed in claim 15, wherein the functions of one or more of the transport device, the drilling device, and the retaining element are controllable by a control device. 17. The drilling instrument as claimed in claim 1, wherein the functions of one or more of the transport device, the drilling device, and the retaining element are controllable by a control device. 18. The drilling instrument as claimed in claim 1, wherein the support plate or the resting plate has at least one limiting element which is arranged in a recess, and in that the shape of the recess allows play in the second position of the connecting element. 19. The drilling instrument as claimed in claim 1, wherein the support plate or the resting plate has at least one approximately conical or frustoconical centering element that is configured to cause the support plate and the resting plate to be arranged in a predefined position with respect to one another when the connecting element is positioned in the first position.
049903055	summary	TECHNICAL FIELD The present invention relates generally to the production of tubing having a desired final diameter by combination of mechanical and thermal processing steps and particularly to a method for producing tubing of a desired final diameter made of zirconium alloys which creates an enhanced radial texture in the tubing. BACKGROUND ART Tubing made of zirconium alloys is widely employed in the nuclear industry, primarily as cladding for nuclear fuel rods. This particular application requires a relatively thin-walled tubing that is resistant both to chemical attack and mechanical attack. Such tubing is typically formed by a combination of mechanical and thermal treatments. Pilgering, one commonly employed tubing formation method, changes the texture of the tubing by gradually reducing the cross-sectional diameter of the tubing while simultaneously increasing the axial elongation of the tubing until the desired optimum final diameter and tube wall thickness are achieved. This process causes the hydrides in the tubing material to be oriented in a circumferential direction. Nuclear fuel cladding tubes subjected to nuclear radiation for the prolonged periods of time characteristic of nuclear reactor operating cycles tend to expand axially. This irradiation-induced axial cladding deformation is also accompanied by a reduction in the radial thickness of the tubing wall and a concomitant decrease in end of cycle life ductility. Texturing the cladding tubes has been found to reduce or avoid some of the problems associated with lengthy use in a reactor environment. One method of texturing, which reduces axial irradiation growth and increases end of cycle life ductility, has been proposed in U.S. Pat. No. 3,804,708. This result is achieved by expanding the tube diameter while constraining the tube ends to prevent an increase in length relatively greater than the increase in diameter. Although tubing having the texture produced according to this method represents an improvement over that of previously available tubing, it is not as resistant to pellet-cladding interaction as could be desired. The zirconium alloys preferred for use in tubing for nuclear applications have a hexagonal close packed crystal structure. The orientation of the basal poles of the metal crystals has been determined to have a significant effect on the texture and, hence, the ultimate properties of tubing formed from zirconium alloys. Increasing the zirconium alloy radial texture produces tubing that is less likely to be susceptible to chemical and/or mechanical attack than nontextured or only slightly textured tubing metal. The method of the aforementioned U.S. Pat. No. 3,804,708 orients the tubing zirconium alloy crystals so that the basal pole principal components are predominantly in both the radial and axial directions. However, orientation of the basal poles to the radial direction to increase the radial texture produces tubing with enhanced properties. A method of increasing the radial texture of basal poles in the crystal structure of zirconium is disclosed in U.S. Pat. No. 4,765,174. The method described in this patent, however, only achieves an appreciable increase in final tube radial texture when the tubing is processed during the intermediate stages prior to expansion to the final tubing diameter. The zirconium alloy tubing processing method described in this patent exhibits an improved texture over that previously achieved. However, the split radial texture characteristic of zirconium alloy tubing produced according to this method may not provide optimal pellet-cladding-interaction resistance or resistance to chemical attack and/or mechanical deformation. Moreover, because the thermal and mechanical processing described in U.S. Pat. No. 4,765,174 takes place before the final tubing expansion, the degree of precise control over the final tubing texture desired may not always be possible. A need exists, therefore, for a method of texturizing zirconium alloy tubing for use as cladding in nuclear fuel rods that is performed during the final tubing processing stages to produce tubing with an increased radial texture. The prior art fails to disclose a method for producing a zirconium alloy tubing characterized by a high degree to radial texture wherein the increased texture is produced during the final stage of tubing fabrication. Further, the prior art fails to disclose a method for producing a highly textured zirconium alloy tubing having a single peak radial basal pole texture. SUMMARY OF THE INVENTION It is a primary object of the present invention, therefore, to overcome the disadvantages of the prior art discussed above and to provide a method for producing a relatively thin-walled, textured zirconium alloy tubing wherein a high degree of radial texture is imparted to the tubing walls during the final stage of tubing fabrication. It is another object of the present invention to provide a method for producing a highly textured zirconium alloy tubing having a single peak radial basal pole texture. It is a further object of the present invention to provide a method for producing a textured zirconium alloy tubing wherein enhanced radial texture is achieved by expansion of the tubing to final size dimensions. It is yet a further object of the present invention to provide a method for producing textured zirconium alloy tubing wherein an enhanced split pole radial texture is achieved by a final mechanical expansion followed by a final heat treatment. In accordance with aforesaid objects, a method for producing a highly textured zirconium alloy tubing having a single pole radial texture suitable for use in forming cladding for nuclear fuel rods is provided. The method includes the steps of processing the tubing to a diameter near the desired final diameter of the finished tubing, preferably to a diameter that is less than about 10 to 20% of the final diameter, optionally subjecting the expanded diameter tubing to a stress relief anneal or to a recrystallization anneal, and then expanding the diameter of the tubing to the desired final diameter, thereby producing a unique single peak radial texture in the finished tubing. An alternate embodiment of this method includes the additional step of performing a final recrystallization anneal on the finally sized tubing to produce an enhanced split pole radial texture in the tubing. Other objects and advantages of the present invention will be apparent from the following description, claims and drawings.
	description	Referring to FIG. 1, a typical projection electron beam lithography system is generally shown at 10. Lithography system 10 comprises an exposure column unit 12 and a control unit 14. The exposure column unit 12 includes an electron beam generator 16 having a cathode 18, a grid 20 and an anode 22. A slit 24 is provided in the exposure column unit 12 for rectangular shaping of the electron beam. A lens 26 is provided for converging this shaped beam. A slit deflector 28 is provided for deflecting a position of the shaped beam to a mask 30 based on a deflection signal. The mask 30 is mounted movably in a horizontal direction between two opposing lenses 32 and 34. Deflectors 36-39 are provided to deflect the beam between lenses 32 and 34 based on position information to select an aperture in the mask 30. The exposure column unit 12 further includes a blanking aperture electrode 40 for cutting off or passing the beam in response to a blanking signal, a lens 42 for converging the beam, an aperture 44, a refocus coil 46 and a lens 48. Also, a dynamic focus coil 50, a set of dynamic stigmator coils 52, an objective lens 54 projecting the beam onto a wafer 55 are provided. A main deflector 56 and a sub deflector 58 position the beam on the wafer in response to exposure position signals. A stage 60 for carrying the wafer and moving it in X-Y directions and alignment coils 62-65 are provided. Such is merely exemplary of a typical exposure control unit. The control unit 14 includes a processor 72 and associated memory having stored design data. Control management for each of the aforementioned components is provided by control unit 14, as is well known. Details of the control management are well known and are not the subject of the present invention. Such lithography systems are well known and the above is provided for illustration purposes only, it is not in any way intended to limit the present invention which is applicable to lithography systems/tools in general. In the prior art, pattern resolution in the system was maximized using a test pattern on a mask. This test pattern would have geometries (e.g., lines or elements) which are smaller (or finer) than the lines or elements of a production pattern on a production mask. In the present invention, the system maximizes pattern resolution using a test mask having test pattern geometries that are the same size as (and even larger than that of) the geometries of a production pattern of a production mask. However, unlike in the prior art, the test wafer is exposed to the test mask image for a period of time shorter than a product pattern exposure. It has been found that image quality differences are more easily detected at these lower exposure dose levels, just as they are also more easily detected with smaller (finer) geometries in the prior art. Adjustments to the system during setup are made in the same manner as in the prior art, such being well known, e.g., current changes to lenses and correction coils, deflection positioning and stage positioning. In this way, smaller geometries for the pattern of the test mask are not required. The ease in detection of image quality differences is evidenced with reference to FIGS. 2-4, where a five-by-five test patten of a x+ symbol is shown with the in-axis stigmator varied by row and the off-axis stigmator varied by column. In FIG. 2 the test pattern is exposed at 1.0xc3x97nominal dose and the x+ symbols all appear to be of similarly good image quality. However, in FIG. 3 the test pattern is exposed at 0.9xc3x97nominal dose and the x+ symbols at the perimeter are clearly of a lesser quality than the x+ symbols at the center. One can see that the rectangles that make up the x+ symbols are no longer of uniform width and quality. This is even more prominent in FIG. 4 where the test pattern is exposed at 0.8xc3x97nominal dose. At this level information is lost at the perimeter and the apparent image quality degrades the further away from the center these features are located. A review of these FIGS. 2-4 clearly shows that evaluating the image quality at a below nominal dose level would enhance the sensitivity to stigmator adjustment errors. This enables an operator to make more accurate stigmator correction adjustments and thereby improves the resolution for product patterns exposed at the nominal dose levels. In other words, adjusting the system at the 0.8xc3x97nominal dose (FIG. 4) for optimal image quality would provide better ultimate pattern resolution than adjusting the system at 1.0xc3x97nominal dose (FIG. 2), since small errors are more readily apparent. The exposure dose of these symbols (features) at 10%-20% below that required for normal resolution, is such that the bottoms of the processed features are just on the verge of clearing to the substrate interface after normal development. In this state small variations in the system image quality have a dramatic effect on the perceived quality of the developed feature (symbol). Testing of this process on an electron beam projection lithography tool with a particular resist process has shown that by dosing a stigmator varied pattern with 200 nm features at xcx9c85% of the nominal dose, the clearing of the features is so marginal that one can easily discern the best image quality parameters using an optical microscope for evaluation. Scanning electron microscope evaluation of 80 nm lines and spaces printed after image quality adjustment confirmed these results. For other resist processes the optimum test pattern dose would have to be determined by experimentation, but should be in the range of 75-90% of the nominal dose. Referring to FIG. 5, a simulation of developed trenches is shown at a nominal dose and at 0.75xc3x97nominal dose at focus settings from xe2x88x9210 microns to +10 microns. At the nominal dose the developed trenches for all focus settings in the range xe2x88x926 microns through +6 microns from the nominal image plane would look similar based on top down optical microscope evaluation. At 0.75xc3x97nominal dose for the resist conditions simulated here a noticeable difference in the trenches occurs at focus settings of xe2x88x924 microns and +4 microns from the nominal image plane, further evidencing the advantages of the present invention. While the above exemplary embodiment is direct towards lithography using a mask with an electron beam such is equally applicable to lithography using a mask with an ultraviolet light (beam) and direct write lithography with an electron beam. In each of these lithography systems, the system can be adjusted to maximize pattern resolution using the method (process) of the present invention. In direct write lithography a test pattern is written at a reduced exposure dose, with such information being used for setup of the system. Also in developing a pattern on a mask (a photolithographic mask) using an electron beam, a test pattern can be projected at a reduced exposure dose, with such information being used for setup of the system. While preferred embodiments have been shown and described, various modifications and substitutions may be made thereto without departing from the spirit and scope of the invention. Accordingly, it is to be understood that the present invention has been described by way of illustrations and not limitation.
	summary
059294582	description	DESCRIPTION OF PREFERRED EMBODIMENTS A radiation shield 1, which is shown in FIG. 1, includes a flexible bag made from synthetic resin cloth, a rubber plate or their composite material. The radiation shield 1 has a hollow interior, as shown in FIG. 2, and water is injected into the hollow interior as a shielding liquid. For the purpose of injecting water, lower and upper side portions of the radiation shield 1 are respectively provided with a water injecting port 3 and an exhaust port 4, as shown in FIG. 1. Either of the water injecting port 3 and the exhaust port 4 can be openably closed with a stopper or the like. A plurality of longitudinal ribs 12, which are made of the same material as the bag of the radiation shield 1, are integrally formed at spaced intervals on outside surfaces of the bag of the radiation shield 1. A reinforcement pipe 2 is inserted in each of the longitudinal ribs integrally with the bag. The material of the reinforcement pipe 2, whether metallic or non-metallic, is selected to have a higher bending strength than the bag. When the radiation shield 1 is to be used, the respective stoppers are removed from the water injecting port 3 and the exhaust port 4. Then, water is injected through the water injecting port 3 and the internal air is exhausted from the bag through the exhaust port 4, whereby the radiation shield 1 is fitted with water so that the radiation shield 1 has a thickness which can shield radiation. After that, the water injecting port 3 and the exhaust port 4 are closed with the respective stoppers. Owing to an increase in the weight of the radiation shield 1 due to the water contained therein, the radiation shield 1 tends to deform so that its lower portion swells and its upper portion becomes too thin to shield radiation. However, since such deformation is prevented by the longitudinal ribs and the strength of the reinforcement pipes 2 of the respective longitudinal ribs, a sufficient thickness for radiation shielding can be maintained over the whole of the radiation shield 1. After the use of the radiation shield 1, the water injecting port 3 and the exhaust port 4 are opened to discharge the water from the radiation shield 1, and the radiation shield 1 is folded into a compact form by folding the portion between each of the reinforcement pipes 2, or it is rolled for storage without any of the reinforcement pipes 2 being folded or bent. Accordingly, the radiation shield 1 is easy to handling because of its compactness and can be stored in a small space. It is more preferable to set the strength of the reinforcement pipes 2 so that no large deformation occurs in the radiation shield 1 even if the water inside the radiation shield 1 is shaken by an external force such as an earthquake. Since the reinforcement pipes 2 have lengths extending in their longitudinal directions and are not connected to one another, the radiation shield 1 might fall horizontally. To cope with this problem, as shown in FIG. 3, a plurality of reinforcement pipes 2 may be connected to one another by connectors 5a which are bent at their opposite ends, for the purpose of horizontal reinforcement. Such connection is made by first fitting one bent end of any of the connectors 5a into one end of any of the reinforcement pipes 2 and then fitting the other bent end of the connector 5a into one end of the reinforcement pipe 2 located in the desired reinforcement direction. When the radiation shield 1 is to be put away, the radiation shield 1 is rolled or folded with the connectors 5a removed from the reinforcement pipes 2. The connectors 5a may be replaced with connectors 5b each having an arrangement in which fitting metals to be removably fitted into the reinforcement pipes 2 are connected to each other by a metal chain 5c. If a longitudinally expanded surface is to be constructed as a radiation protection surface, a plurality of radiation shields 1 into which water is injected may be stacked in the vertical direction, as shown in FIG. 5. In the stacking of the radiation shields 1, the stacking positions of the radiation shields 1 are adjusted so that the reinforcement pipes 2 are arranged in a line in the vertical direction. In the stacking of the radiation shields 1, as shown in FIG. 6, connectors 6 each having a flange which is larger in diameter than the reinforcement pipes 2 are fitted at vertical intermediate positions in such a manner that each of the connectors 6 is inserted between adjacent ones of reinforcement pipes 2 stacked in the vertical direction, whereby the reinforcement pipes 2 are linked together in the vertical direction so that the radiation shields 1 located in an upper position do not easily fall or come off. If a horizontally expanded surface is to be constructed as a radiation protection surface, a plurality of radiation shields 1 into which water is injected are arranged adjacent to one another in the horizontal direction, as shown in FIG. 7. Each of the radiation shields 1 is connected to the adjacent one at the reinforcement pipes 2 located at respective adjacent sides, by connectors 7a. Each of the connectors 7a is made from a U-shaped bar member. As shown in FIG. 8, one connector 7a is fitted at one end into the reinforcement pipe 2 of one of two adjacent radiation shields 1 and at the other end into the reinforcement pipe 2 of the other radiation shield 1, whereby the adjacent radiation shields 1 are connected to each other and the deviation of the relative position between them is restrained so that a gap through which radiation leaks is prevented from easily occurring. In addition, if a connector 7b is employed, the deviation of the relative position is restrained to a further extent, so that the occurrence of a gap through which radiation leaks is more securely prevented. As shown in FIG. 9, the connector 7b is a member having a U-shaped cross section and clamps the longitudinal-rib of one of two adjacent radiation shields 1 and the longitudinal rib of the other radiation shield 1. The longitudinal ribs are clamped at two or three positions dispersed in the vertical direction. The connectors 7b may be used alone or together with the connectors 7a. In either case, the connectors 7b restrain gaps from occurring between adjacent ones of the radiation shields 1. To make the radiation shield 1 more portable and easier to handle, the structures shown in FIGS. 10 to 13 are adopted. In the structure shown in FIGS. 10 and 11, running means each having a wheel 8 are fitted to the bottom ends of the respective reinforcement pipes 2 of the radiation shield 1 so that the radiation shield 1 can readily be moved by the rolling of the wheels 8. If this structure is adopted, the radiation shield 1 filled with water can readily be moved to and installed at a radiation shielding position, and can readily be moved away therefrom. The structure shown in FIGS. 12 and 13 is provided with the wheels 8 similarly to the structure shown in FIGS. 10 and 11, but the following structure is added. Specifically, two upper and lower portions of each of the longitudinal ribs are cut away and the pipe reinforcement 2 is partly exposed. The exposed portions of each of the pipe reinforcements 2 are respectively provided with sliders 9a which are movable upward and downward, and links 9 which cross each other in an X-like form are vertically swingably fitted to adjacent ones of the sliders. The crossing of the links 9 assembled in the X-like form is swingably fitted. When such expandable link mechanism is expanded rightward and leftward, the radiation shield 1 can be rapidly unfolded to be set to a usable state. When the radiation shield 1 is to be put away, water is discharged from the radiation shield 1 and the link mechanism is shrank, whereby the radiation shield 1 can be rapidly folded into a compact shape suited to storage. Since this example is also provided with the wheels 8, the handling and movement of the radiation shield 1 are easy. Although each of the above-described embodiments adopts the reinforcement pipes 2 as reinforcement members, bars which are not pipe-shaped but solid may replace the reinforcement pipes 2 as reinforcement members. In this case, each kind of connector is made from a hollow shaped member, and the relation between the fitting side and the fitted side is reversed. The manner in which the wheels 8 are fitted is similarly reversed.
047175320	description	DETAILED DESCRIPTION The present invention provides a pressure control system for a pressurized water nuclear reactor plant containing an improved two stage sparger for use in the pressurizer relief tank that provides efficient steam distribution and condensation in the pressurizer relief tank while limiting the pressure drop for both normal and anticipated transient without trip events. In FIG. 1, there is schematically illustrated a pressure control system for a pressurized water nuclear reactor which can incorporate the two stage sparger apparatus of the present invention. The pressure control system 1 contains a pressurizer 3, normally formed as a vertical, cylindrical vessel, composed of carbon steel with austenitic stainless steel cladding on all surfaces exposed to primary reactor coolant. Electrical heaters 5 are provided in the bottom portion of the pressurizer 3 and spray nozzles 7 are provided in the upper portion thereof. The pressurizer is designed to accommodate positive and negative surges caused by load transients on the system. A surge line 9, attached to the bottom of the pressurizer 3 connects the pressurizer with the hot leg of a reactor coolant loop. During an insurge, the spray nozzles 7 which are fed from the cold leg of the reactor coolant loop through line 11, spray water into the upper portion of the pressurizer 3 to condense steam in the pressurizer 3 to prevent the pressure in the pressurizer from reaching the setpoint of power operated relief valves 13 in lines 15. During an outsurge, flashing of water to steam and generating of steam by actuation of heaters 5 keep the pressure above the low pressure reactor trip setpoint. The pressurizer 3 is also provided with safety relief valves 17, three being shown in FIG. 1. The safety relief valves 17, in lines 19, are spring loaded or self-activated with back pressure compensation, and loop seals 21 are provided in lines 19 for valve protection. Such loop seals (not shown) are also normally provided in lines 15 for the protection of power operated relief valves 13. The combined capacity of the safety relief valves 17 is equal to, or greater than, the maximum surge rate resulting from the complete loss of load without reactor trip or any other control. Power operated relief valves 13 discharge to line 23 to a pressurizer relief tank 25 which contains a sparger, while safety relief valves 17 discharge into branch lines 27 which communicate with line 23 and to the pressurizer relief tank 25. In such conventional systems, the line 23 to the pressurizer relief tank was normally 12 inches in diameter and the sparger, attached to the end of line 23, within the pressurizer relief tank, was also a 12" diameter pipe with 1/2 inch orifices therein, and having an end cap. In the improved pressure control system of the present invention, the sparger within the pressurizer relief tank is a two stage sparger which is comprised of a primary conduit having orifices, a secondary conduit having orifices, and an intermediate valve means actuated by a pressure differential between the two conduits. Referring now to FIGS. 2 and 3, the two stage sparger 27 in a closed pressurizer relief tank 25 has a primary conduit 31 which has orifices 33 formed in walls 35 thereof, and a secondary conduit 37 attached to the terminus 39 of the primary conduit, the secondary conduit 37 having orifices 41 formed in the walls 43 thereof. The orifices 33 and 41 are formed in long rows, along the horizontal axis of the conduit and are located, in the side walls of the conduit, normally within an area no more than about 30.degree. above and below the horizontal axis. A coupling 45 is attached to the inlet end of the primary conduit 31 for connection thereof to the line 23 of the pressure control system. The two stage sparger is located in the pressurizer relief tank within a supply of liquid coolant 47 such as water, and rupture disks (not shown) are provided on the closed pressurizer relief tank. A valve means 49 (FIGS. 4 and 5) connects the primary conduit 31 and the secondaary conduit 37 together, such that communication between the two conduits is effected, when desired. The valve means 49 preferably comprises a spring loaded check valve having a valve seat 51, closure element 53 and a biasing spring 55 that biases the closure element against the valve seat to close the valve. As illustrated, the terminus 39 of the primary conduit may have a step-down section 57 thereon which enables the use of a secondary conduit 37 of a diameter d' less than the diameter d of the primary conduit 31. As an example of such dimensions, the line 23 from the power operated relief valves and safety relief valves can be about 16 inches in diameter, the primary conduit 31 would be about 16 inches in diameter, and the secondary conduit 37 of about 12 inches in diameter. At least a portion of the orifices 41 in the secondary conduit 37 should be below the level, or horizontal plane. of the orifices 33 in the primary conduit 31 so as to insure that water will pass through the primary conduit 31, and valve 49 to the secondary conduit 37 for discharge therefrom. Also, the total area of flow through the orifices 41 of the secondary conduit 37 should be greater than the total area of flow through the orifices 33 of the primary conduit 31, and preferably about 125 percent of the flow area of the orifices of the primary conduit. This increase in flow area through the orifices in the secondary conduit can be effected by the use of a larger number of such orifices or by providing orifices through the walls of the secondary conduit of a diameter greater than the diameter of the orifices through the walls of the primary conduit. As illustrated, a preferred construction of the secondary conduit 37 has a connecting section 59 communicating with the check valve 49, and is bifurcated (FIG. 2) at cross section 61 to form two leg portions 63 which extend back towards the primary conduit 31. The leg sections have end caps 65 at the ends thereof. The orifice 39 can be formed in the connecting section 59, cross section 61, the leg sections 63, and the end caps 65, the number of orifices and placement thereof along the secondary conduit dependent upon the flow are desired. In the operation of the pressure control system of the present invention, in the event of an anticipated transient without trip, the following would occur. With a loss of load on the reactor, without a reactor trip, or shutdown, the reactor coolant system pressure in the pressurizer 3 increases rapidly. The power operated relief valves 13 and safety valves 17 would open to discharge cold water loop water seals and steam into the line 23 to the pressurizer relief tank 25. The loop seal "plugs", or water seals, compress nitrogen in the line 23 which forces water in the primary conduit 31 out through the orifices 33. As the pressure in the primary conduit 31 exceeds a predetermined pressure, preferably of about 50 psig (such pressure could rise to a value of about 400 psig in a conventional sparger), the check valve 49 opens and the fluid passes to the secondary conduit 37, which, with orifices 41 more than doubles the total flow area for fluid from the sparger. Water in the primary conduit 31 is discharged therefrom to the secondary conduit 37, and out of the orifices 41, reducing the pressure in the primary conduit to below the predetermined value and the check valve 49 closes as steam reaches the pressurizer relief tank through line 23. Steam discharged through the orifices 33 of the primary conduit 31 is condensed in the water 47 in the pressure relief tank 25 since the design of the primary conduit is effective to maintain a pressure differential below the predetermined value that would actuate the check valve. The steam content of the pressurizer 3 is discharged with the pressurizer relief tank at about 50 psig pressure and 200.degree. F. The reactor coolant system pressure increases (2500 psig to 2750 psig) when the pressurizer becomes water filled and water discharge from the pressurizer 3 through power operated relief valves 13 and safety valves 17 to line 23 begins. With the pressurizer filled with, and discharging water, the volumetric discharge through the power operated relief valves and safety valves is unchanged but mass flowrate is increased by roughly a factor of 5. As the water flow increases through the line 23 to the primary conduit 31, valve backpressure increases and the pressure differential between the primary conduit 31 and secondary conduit 37 increases to above the predetermined value and the check valve 49 again opens to provide for flow through the orifices of both sparger conduits, until the reactor is tripped either automatically or manually. The flow through the sparger is illustrated in FIGS. 4 and 5 which illustrate fluid flow through the orifices 33 of the primary conduit 31 when the check valve 49 is closed, and fluid flow through both the orifices 33 of primary conduit 31 and orifices 41 of secondary conduit 37 when the check valve 49 is in open position.
	claims	1. An apparatus for securely storing one or more radioactive sources in source containers within a mobile structure comprising:an enclosure having a plurality of sides formed with horizontal and vertical sides and a back, the plurality of sides forming an open side;removable or moveable shelves coupled or fixed within the enclosure;a hinge coupled to an edge section of one of said horizontal sides;a door section coupled to said hinge, said door section comprising a frame and a transparent section coupled to the frame which is sized to view each of the removable or moveable shelves;a locking mechanism coupled to the door section and adapted to selectively engage with the enclosure, wherein the locking mechanism is formed to automatically lock and engage with the enclosure when the door section is rotated to abut door jamb section of the enclosure, wherein the locking mechanism is formed with a key reader and a key etched with a pattern encoded or etched by a laser or another etching machine that is read by the key reader;a door handle coupled to the door section;a closing assist magnet coupled to a section of the enclosure disposed so it magnetically engages the door section to pull the door section against the enclosure when the door section is within a magnetic field of the closing assist magnet;an alarm mechanism comprising a first and second section, wherein the first section comprises a magnet that is coupled to the door section, wherein the second section is coupled to the enclosure, wherein the alarm mechanism has an activation section that activates an audible alarm when the first section is not adjacent to the second section;a tether section comprising a first and second attachment section and a tether coupler attached to the first and section attachment sections, wherein the first attachment section is coupled to an external section of the enclosure and the second attachments section is adapted to be coupled to an adjacent structure within the mobile structure; anda removable and selectively lockable protective bracket that is coupled with the alarm mechanism to lock the alarm mechanism in place and to obstruct or cover an on/off switch of the alarm mechanism. 2. An apparatus as in claim 1, wherein the mobile structure is a submersible vessel. 3. An apparatus as in claim 1 further comprising an object tracking system disposed within or in proximity with the enclosure comprising a bar code reader or radio frequency identification (RFID) tag reader adapted to read the bar code or RFID tag that is placed on a radiological source object that is removed or inserted into the enclosure. 4. An apparatus as in claim 3, further comprising a tracking computer system that is communicatively coupled with the object tracking system. 5. An apparatus as in claim 1, further comprising said radioactive source containers disposed within the enclosure and placed on said shelves. 6. An apparatus as in claim 5, further comprising locking or retention structures that are coupled with one or more said shelves that includes latching or retention structures that selectively receive and fix in place said radioactive source containers with respect to the shelves. 7. An apparatus as in claim 6, wherein each of the locking or retention structures further comprise a base section that has a different shape, wherein each of the radioactive source containers have a different base section that is each shaped to fit into a corresponding one of the base sections that each have a different shape. 8. An apparatus as in claim 6, wherein each of the radioactive source containers have a first keyway section and each of the locking or retention structures further comprise a second keyway section, wherein the first and second keyway sections are adapted to selectively engage with each other, each of the first and second keyway sections have a corresponding female and male engaging section that fit into each other. 9. An apparatus as in claim 1, wherein said transparent section comprises transparent Plexiglas, polycarbonate material, or leaded glass material. 10. An apparatus as in claim 1, further comprises transparent plastic film adhered to the transparent section. 11. An apparatus for securely storing one or more radioactive sources in source containers within a mobile structure comprising:an enclosure having a plurality of sides formed with horizontal and vertical sides and a back, the plurality of sides forming an open side;removable or moveable shelves coupled or fixed within the enclosure;a hinge coupled to an edge section of one of said horizontal sides;a door section coupled to said hinge, said door section comprising a frame and a transparent section coupled to the frame which is sized to view each of the removable or moveable shelves;a locking mechanism coupled to the door section and adapted to selectively engage with the enclosure, wherein the locking mechanism is formed to automatically lock and engage with the enclosure when the door section is rotated to abut door jamb section of the enclosure, wherein the locking mechanism is formed with a key reader and a key etched with a pattern encoded or etched by a laser or another etching machine;a door handle coupled to the door section;a first magnet coupled to a section of the enclosure disposed to magnetically engage a side of the door section to pull the door section against the enclosure when the door section is within a magnetic field of magnet;an alarm mechanism comprising a first and second section, wherein the first section comprises a second magnet that is coupled to the door section, wherein the second section is coupled to the enclosure, wherein the alarm mechanism has an activation section that activates an audible alarm when the first section is not adjacent to the second section;a tether section comprising a first and second attachment section and a tether coupler attached to the first and section attachment sections, wherein the first attachment section is coupled to an external section of the enclosure and the second attachments section is adapted to be coupled to an adjacent structure within the mobile structure;an object tracking system disposed within or in proximity with the enclosure comprising a bar code reader or radio frequency identification (RFID) tag reader adapted to read the bar code or RFID tag that is placed on a radiological source object that is removed or inserted into the enclosure;a tracking computer system that is communicatively coupled with the object tracking system;a plurality of said radioactive source containers disposed within the enclosure and placed on said shelves; anda plurality of locking or retention structures coupled with said removable or moveable shelves adapted to selectively receive and fix in place said radioactive source containers;a removable and selectively lockable protective bracket that is coupled with the alarm mechanism to lock the alarm mechanism in place and to obstruct or cover an on/off switch of the alarm mechanism;wherein each of the locking or retention structures further comprise a base section that has a different shape, wherein each of the radioactive source containers have a different base section that is each shaped to fit into a corresponding one of the base sections that each have a different shape;wherein the mobile structure is a ship or submersible vessel. 12. A method of using a storage apparatus for securely storing radioactive source comprising:providing a mobile structure that is either a shop or submersible vessel;providing an apparatus for storage of radioactive sources, said apparatus comprising:an enclosure having a plurality of sides formed with horizontal and vertical sides and a back, the plurality of sides forming an open side;a plurality of removable or moveable shelves coupled or fixed within the enclosure;a hinge coupled to an edge section of one of said horizontal sides;a door section coupled to said hinge, said door section comprising a frame and a transparent section coupled to the frame which is sized to view each of the removable or moveable shelves;a door handle coupled to the door section;a first magnet coupled to a section of the enclosure disposed so it magnetically engages the door section to pull the door section against the enclosure when the door section is within a magnetic field of the first magnet;a locking mechanism coupled to the door section and adapted to selectively engage with the enclosure, wherein the locking mechanism is formed to automatically lock and engage with the enclosure when the door section is rotated to abut a door jamb section of the enclosure, wherein the locking mechanism is formed with a key reader and a key etched with a pattern encoded or etched by a laser or another etching machine that is read by the key reader;an alarm mechanism comprising a first and second section, wherein the first section comprises a magnet that is coupled to the door section, wherein the second section is coupled to the enclosure, wherein the alarm has an activation section that activates an audible alarm when the first section is not adjacent to the second section; anda tether section comprising a first attachment section and a tether coupler attached to the first attachment sections;providing a submersible vessel with an interior compartment formed with a plurality of wall sections, wherein a second attachment section is coupled with one of the wall sections, wherein the first attachment section is coupled to the second attachment section by the tether coupler;operating the submersible vessel to transit a first section of water and thereby altering an orientation of the interior compartment with respect to a gravitational vector;inserting said key and disengaging the locking mechanism;pulling the door handle while maintaining a first force on the key to keep the key engaged/rotated to open the door section, wherein once the said door section is opened, the second magnet moves out of detection range of the alarm mechanism and thereby causing the alarm mechanism to generate a continuous audible alarm sound that will sound and continue until the door is securely closed and locked;placing the radioactive source inside the enclosure on one of the shelves and into a source receiving section that fixes the radioactive source into a fixed position;closing the door section and releasing pressure on the key so the rotates into a spring loaded locked position then removing the key from the locking mechanism;conducting an inventory of the one or more radiological sources by viewing the radiological sources through the transparent section and determining if the radiological sources are present within the enclosure while the door section is in a closed position;when a user is ready to remove the radioactive source, obtaining the key from a locked storage container;inserting the key into the locking mechanism and disengaging the locking mechanism from engagement with the enclosure;pulling the door handle while maintaining light pressure to keep key engaged/rotated to rotate and open the door, wherein once door is opened, the magnet will disengage and the audible alarm will sound and continue until the door section is closed and locked by the locking mechanism;removing the radioactive source from one of the shelves inside the enclosure; andclosing the door section, thereby deactivating the alarm; andreleasing pressure on the key and removing the key from the locking section.
048615200	description	DETAILED DESCRIPTION OF THE INVENTION The known prior art source capsules can be best understood from FIG. 1, which shows a typical example thereof. The source capsule is composed of a tubular body 1 having a closed back end 2 and a cavity 3 for receiving and containing one or a plurality of radioactive sources 4, seven of which are shown in FIG. 1. The sources 4 are sealed in cavity 3 by plug 5 which, as shown in FIG. 5, is usually attached to tubular body 1 by a weld 6. Tubular body 1, in turn, is attached to a flexible cable 8 by means of a further weld 7. In such prior art devices, as shown in FIG. 1, the tubular body 1 has a typical length of 5.8 millimeters and the plug 5 has a typical length of 1.35 millimeters, the length of the plug being necessary to secure the plug in a holder during welding and performing the pull test, as explained above. Thus, the overall length of the capsule is 7.2 millimeters, apart from weld 7. Turning now to FIG. 2, which shows an embodiment of the present invention, the drivable radioactive source capsule of the present invention comprises a similar tubular body 1 having a similar cavity 3, but having a first end 10, which is preferably rounded as shown, formed thereon. By having such a rounded first end 10, as opposed to the plug 5 of the prior art (see FIG. 1), the first end of the capsule is much shorter than the front end of the prior art device containing the plug. This not only shortens the capsule, as noted above, but most desirably, locates the radioactive sources more near the front end of the capsule. This allows more accurate placement of the radioactive sources in the tissues being treated. Further, the entire metal walls and end of the capsule act as radiation shields. As shown in FIG. 1, the relatively massive plug 5 effects much greater shielding than the walls of the capsule. This results in a "dimple" in the isodose of radiation (the radiation dosage with respect to distance from the surfaces of the capsule). By eliminating the plug 5, the present rounded end does not effect a significant "dimple". Elimination of such "dimples" is especially important when using low energy-long wave length isotopes, such as Ir.sup.137, as opposed to higher energy sources such as cobalt 60. Tubular body 1 also has a second end 11 which is the terminus of the tubular body, as shown in Figure 2. As opposed to the prior art, the present invention provides a plug 12 having an elongated closure portion 13 with the diameter of the closure portion being substantially equal to the inside diameter (shown by arrows 14) of the tubular body 1. It will be appreciated in this regard that the term "substantially equal" means that the diameter of the elongated closure portion 13 is close to but slightly less than the inside diameter of tubular body 1, so that the closure portion 13 may be snugly fitted into tubular body 1. As can also be seen from FIG. 2, the closure portion 13 is of a sufficient length that it can be accurately placed in tubular body and will snugly contain the precise numbers of sources in the capsule such that the sources are not free to move within the capsule during use of the capsule. Of course, the radioactive sources will normally have a diameter essentially equal to (but slightly less than) the inside diameter of tubular body 1. Plug 12 also has a connection portion 15 disposed adjacent to the closure portion and forming a unitary plug. The diameter of the connection portion 15 is substantially equal to the outside diameter of the tubular body 1 and, also, preferably is substantially equal to the diameter of the drive cable 8. Thus, preferably the diameters of the drive cable 8, the plug 12 and the tubular body 1 are all substantially equal, so that the combination of the drive cable and capsule may be passed through a tubular guide for correctly disposing the capsule, with the radioactive sources therein, in the patient being treated. This is also preferred since if the diameter of the cable is substantially less than the diameter of the capsule, e.g. one-third less, (and, hence, also substantially less than the internal diameter of the tubular guide), the cable can bend or "snake" within the tubular guide during movement therethrough. This can result in not only binding of the cable, but also result in the cable actually moving the capsule a shorter distance through the tubular guide than would be indicated by the actual length of the cable having been driven from the head of the after loading apparatus. This could give a false indication as to the final position of the capsule in the patient being treated. Elongated flexible drive cable 8 is connected to the connection portion 15 of the plug 12 prior to assembly of the tubular body 1 to plug 12. By this arrangement, the radioactive sources 4 may be placed in the tubular body 1 through the opened second end 11, and the tubular body 1 is closed by disposing the closure portion 13 of plug 12 into the second end 11 of the tubular body 1 and attaching the closure portion 13 to the second end 11 of the tubular body 1. As noted above, the capsule is normally attached to the drive cable by welding. In the present invention, the drive cable 8 is welded to the connection portion 15 of plug 12 prior to closing tubular body 1 by plug 12. Thus, tongs or other similar holding devices can easily grip the combination of plug 12 and drive cable 8 for accurately placing plug 12 into tubular body 1, welding plug 12 thereto and performing the required pull test. By this arrangement, as opposed to the prior art, there is no need for an elongated plug, which elongated plug substantially increases the overall length of the capsule and adversely effected the isodose line, as explained above. Further, the radioactive sources are contained within the tubular body 1 between the first end 10 and the closure portion 13 of plug 12 so that the radioactive sources are very snugly held within the capsule. After such assembly, the closure portion 13 of plug 12 is attached to the second end 11 of tubular body 1 in any manner desired, but most often this attachment will be by a weld, as in the prior art. However, as noted above, that weld can be easily achieved without the necessity of the extended plug, which was required in the prior art. It has been found in this latter regard, that the weld of the closure portion 13 to the second end 11 is most advantageously carried out when the weld is an electron-beam weld. While electron-beam welding is well known in the art and the details thereof need not be set forth herein for sake of conciseness, with electron-beam welding, weld 6 (see FIG. 4) can very accurately and precisely attach closure portion 13 to second end 11 of tubular body 1. As can be appreciated, the welding of plug 12 to tubular body 1 must be a very accurate weld in order to ensure that plug 12 is fully seated in tubular body 1 and that the attachment of plug 12 to tubular body 1 is quite secure in order to avoid radiation source leakage or dislodgment of the plug, for the reasons explained above. Also, drive cable 8 can be attached to connecting portion 15 of plug 12 in any desired manner, but it is far preferable that that attachment be by means of a weld. However, in this case, it is preferred that the weld is a laser weld. Again, laser welding is well known in the art and need not be described herein for conciseness purposes, but laser welding has been found to be most effective in attaching the flexible drive cable (usually made of steel) to the connecting portion 15 of plug 12. It has been found that laser welding ensures a good connection of cable 8 to plug 12 so that no separation thereof occurs even when the cable and capsule are passed through tortuous turns in the tubular guide. In regard to the method of producing the capsule of the present invention, it is only necessary to provide the tubular body 1 having the rounded end 10. Means of producing such tubular bodies are well known in the art and need not be described herein. The tubular bodies are normally made of steel or like rigid metal material and can be formed by conventional machining techniques. The plug 12 preferably is machined, but it can be formed by a die casting technique, both of which processes are well known in the art. The flexible drive cable is then attached to the connection portion 15 of plug 12, preferably by welding as described above. A plurality of radioactive sources are then placed in tubular body 1 through second end 11. The closure portion 13 of plug 12 is then disposed within the tubular body through the second end 11. Thereafter, the closure portion 13 is attached to the second end 11, preferably by welding as described above. Preferably, the wires of cable 8 are first welded together to form a solid end and then simply butt welded to connection portion 15. However, if desired, a thread may be formed on the solid welded end of cable 8. The connection portion 15 may also be internally threaded (not shown in the drawings) and the threaded solid end of cable 8 is threaded into the internal threads of connection portion 15 and then welded, as described above. As shown in FIG. 2, with the present arrangement, the same number of sources, i.e. 7 sources, can be contained in a capsule with an overall length of 5 millimeters, as opposed to the overall length of 5.85 millimeters in the prior art capsules. Further, since the present plug 12 being previously attached to cable 8 can be much shorter than the prior art plug 5, the overall length of the capsule and plug of the present invention, as shown in FIG. 2, can be 5.5 millimeters, as opposed to the overall length of 7.2 millimeters with plugs of the prior art. This results in a shortening of the overall length of the capsule by 1.7 millimeters. While this shortening may seem quite small, that amount of shortening provides considerable advantages in passing the rigid capsule through a tortuous turn of, for example, a tubular guide and is a decided advantage in the art. In addition, the present rounded end provides a much more desirable isodose line, as explained above. FIG. 3 shows an embodiment similar to FIG. 2, but where the capsule contains eight radioactive sources. Again, it will be seen that the overall length is 6 millimeters, as opposed to the prior art capsule containing only seven sources and having an overall length of 7.2 millimeters. Thus, with the present invention, an additional source can be added to the capsule, while at the same time, substantially shortening the capsule, as opposed to prior art capsules. It will be apparent to those skilled in the art that modifications of the above-described invention can be easily appreciated, and it is intended that those modifications being included- .within the spirit and scope of the annexed claims.
	abstract	Systems and methods for providing and using molten salt reactors are described. While the systems can include any suitable component, in some cases, they include a graphite reactor core defining an internal space that houses one or more fuel wedges, where each wedge defines one or more fuel channels that extend from a first end to a second end of the wedge. In some cases, one or more of the fuel wedges comprise multiple wedge sections that are coupled together end to end and/or in any other suitable manner. In some cases, one or more alignment pins also extend between two sections of a fuel wedge to align the sections. In some cases, one or more seals are also disposed between two sections of a fuel wedge. Thus, in some cases, the reactor core can be relatively long (e.g., to be a pipeline reactor). Other implementations are also described.
	abstract	For maskless irradiating a target with a beam of energetic electrically charged particles using a pattern definition means with a plurality of apertures and imaging the apertures in the pattern definition means onto a target which moves (v) relative to the pattern definition means laterally to the axis, the location of the image is moved along with the target, for a pixel exposure period within which a distance of relative movement of the target is covered which is at least a multiple of the width (w) of the aperture images as measured on the target, and after said pixel exposure period the location of the beam image is changed, which change of location generally compensates the overall movement of the location of the beam image.
041464290	claims	1. A dispersement apparatus for a mass of fissionable material comprising: a collecting chamber; a first passage means connected to said collecting chamber, said first passage means comprises at least one in number of first passages, the upper end of said first passage means capable of receiving said liquid fissionable material from said collecting chamber; second passage means connected to said first passage means at the lower end thereof, said first passage means terminates at said connection with said second passage means said second passage means comprising a plurality of separate second passages adapted to receive liquid fissionable material from said first passage means, there being a plurality of said second passages for each said first passage, each said second passage being substantially smaller in cross-section than each said first passage. the total cross-sectional area of said second passages which are connected to a single said first passage is approximately equal to the cross-sectional area of said first passage. there being at least three in number of said second passages for each said first passage. a solid material contained within a portion of one of said passage means, said solid material being capable of being readily melted upon coming into contact with the heated fissionable material. a collecting chamber: a first passage means connected to said collecting chamber, said first passage means comprises at least one in number of first passages, said first passage means capable of receiving said liquid fissionable material from said collecting chamber; second passage means connected to said first passage means said second passage means comprising a plurality of separate second passages adapted to receive liquid fissionable material from said first passage means, there being a plurality of said second passages for each said first passage; each said second passage being substantially smaller in cross-section than each said first passage; the total cross-sectional area of said second passages which are connected to a single said first passage is approximately equal to the cross-sectional area of said first passage; there being at least three in number of said second passages for each said first passage; a solid material contained within a one of said passage means, said solid material being capable of being readily melted upon coming into contact with the heated fissionable material; said second passage means terminating into a base of loosely packed material, explosive charge means connected to said third passage means within said base, whereby ultimate dispersement of said fissionable material is accomplished by exploding of said explosive charge means and finely dispersing the fissionable material throughout said base. a collecting chamber; a passage means connected to said collecting chamber, said passage means comprising a plurality of spaced apart separate passages, said passage means capable of receiving liquid fissionable material from said collection chamber; and said passage means terminating into a base of loosely packed material, explosive charge means located within said base of loosely packed material, whereby ultimate dispersement of said fissionable material is accomplished by exploding of said explosive charge means and finely dispersing the fissionable material throughout said base. 2. The apparatus as defined in claim 1 wherein: 3. The apparatus as defined in claim 2 wherein: 4. The apparatus as defined in claim 3 wherein: 5. A dispersement apparatus for a mass of fissionable material comprising: 6. A dispersement apparatus for a mass of fissionable material comprising:
	claims	1. A method of operating a nuclear fission reactor, comprising:producing at least a portion of a traveling burn wave at a location relative to a nuclear fission module; andoperating a flow control assembly coupled to the nuclear fission module to modulate flow of a coolant in response to the location of the traveling burn wave relative to the nuclear fission module, the flow control assembly comprising a flow regulator subassembly having a first sleeve having a first hole and a second sleeve inserted into the first sleeve, the second sleeve having a second hole alignable with the first hole as the first sleeve translates in an axial direction relative to the second sleeve. 2. The method of claim 1, wherein operating the flow control assembly comprises modifying the flow regulator subassembly in response to an operating parameter associated with the nuclear fission module. 3. The method of claim 1, wherein operating the flow control assembly comprises reconfiguring the flow regulator subassembly according to a predetermined input to the flow regulator subassembly. 4. The method of claim 1, wherein operating the flow control assembly comprises achieving a controllable flow resistance via the flow regulator subassembly. 5. The method of claim 1, further comprising coupling a temperature sensor to the nuclear fission module and the flow regulator subassembly. 6. The method of claim 1, further comprising controlling flow of the coolant in response to the location relative to the nuclear fission module by operating the flow regulator subassembly according to when the burn wave arrives at the location relative to the nuclear fission module. 7. The method of claim 1, further comprising controlling flow of the coolant in response to the location relative to the nuclear fission module by operating the flow regulator subassembly according to when the burn wave departs from the location relative to the nuclear fission module. 8. The method of claim 1, further comprising controlling flow of the coolant in response to the location relative to the nuclear fission module by operating the flow regulator subassembly according to when the burn wave is proximate to the location relative to the nuclear fission module. 9. The method of claim 1, further comprising controlling flow of the coolant according to a width of the burn wave. 10. The method of claim 1, further comprising controlling flow of the coolant by operating the flow control assembly according to a heat generation rate in the nuclear fission module. 11. The method of claim 1, further comprising controlling flow of the coolant by operating the flow control assembly according to a temperature in the nuclear fission module. 12. The method of claim 1, further comprising controlling flow of the coolant by operating the flow control assembly according to a neutron flux in the nuclear fission module. 13. The method of claim 1, wherein:the nuclear fission module has a temperature dependent reactivity change; andthe flow control assembly controls the temperature dependent reactivity change within the nuclear fission module. 14. The method of claim 1, wherein producing at least a portion of a traveling burn wave comprises producing at least a portion of the traveling burn wave at a location relative to a nuclear fission fuel assembly. 15. The method of claim 1, wherein producing at least a portion of a traveling burn wave comprises producing at least a portion of the traveling burn wave at a location relative to a fertile nuclear breeding assembly. 16. The method of claim 1, wherein producing at least a portion of a traveling burn wave comprises producing at least a portion of the traveling burn wave at a location relative to a neutron reflector assembly. 17. A method of operating a nuclear fission reactor, comprising:producing at least a portion of a traveling burn wave at a location relative to a nuclear fission module; andcontrolling flow of a coolant in response to the location of the traveling burn wave relative to the nuclear fission module with a flow control assembly by relatively axially translating a first sleeve having a first hole and a second sleeve inserted into the first sleeve, the second sleeve having a second hole alignable with the first hole. 18. The method of claim 17, wherein relatively axially translating the first sleeve having the first hole and the second sleeve having the second hole comprises rotating a member coupled to the first sleeve. 19. The method of claim 17, wherein controlling the flow of coolant comprises controlling the flow of coolant in response to an operating parameter associated with the nuclear fission module.
052456420	abstract	A method for chemically controlling cobalt decontaminating the water cooling system of a water cooled nuclear fission reactor to reduce the radiation hazard to personnel.
	summary
048878855	abstract	Arrangements are disclosed for generating a well defined traveling wave beam substantially unaffected by diffractive spreading. In different embodiments, the beam can be an electromagnetic wave, particle beam, a transverse beam, a longitudinal beam such as an acoustic beam, or any type of beam to which the Helmholtz generalized wave equation is applicable. Pursuant to the teachings herein, a beam is generated having a transverse dependence of a Bessel function, and a longitudinal dependence which is entirely in phaser form, which results in a beam having a substantial depth of field which is substantially unaffected by diffractive spreading. In first and second disclosed embodiments respectively, optical and acoustical beams are generated by placing a circular annular source of the beam in the focal plane of a focussing means, which results in the generation of a well defined beam thereby because the far field intensity pattern of an object is the Fourier transform thereof, and the Fourier transform of a Bessel function is a circular function. In a third disclosed embodiment, a microwave beam is generated by transmitting a coherent microwave beam sequentially through a phase modulator, having a periodic stop function pattern, and a spatial filter, whose transmittance is the modulus of the Bessel function, to generate a microwave beam having a transverse Bessel function profile.. More specifically, several embodiments are disclosed of an integrated optical laser cavity and an integrated microwave maser cavity for increasing the efficiency of production of the laser or maser beam. The integrated laser and maser cavities are designed to generate directly from their own gain medium a Bessel-mode diffraction-free beam.
049845101	description	DESCRIPTION OF THE PREFERRED EMBODIMENT In FIGS. 1 to 8, a containment 1 is provided with a posting port 2 through which drums of radioactive waste materials can enter into the containment. The port 2 comprises an annular sphincter seal 3 and a removable lid or cover 4. The lid or cover 4 sealingly engages an inflatable seal 5 mounted in a lip 6 at the outer end of the port 2. Various detectors for controlling the sequence of operation are denoted by the reference numerals 7 to 11 respectively and the function of which will be described in the following description. A platform comprising a roller conveyor assembly 12 is located within the containment 1 below the port 2. A hoist 14 in the form of a fork cooperates with the conveyor assembly 12 such that the fork can pass between rollers 13 of the conveyor assembly. In FIG. 1 the hoist is shown in a retracted position. To enter a first drum into the containment, the hoist 14 is raised to its highest position, which will be indicated by position switch 10, the lid or cover 4 is removed to expose the port 2, and a drum 15 supported by a crane or hoist (not shown) is lowered through the port on to the raised hoist 14. This position is depicted in FIG. 2. The sphincter seal 3 engages the lower end of the drum 15. The detectors 7,8 and 9 can be infra-red switch devices and in FIG. 2 the drum interrupts signals from transmitters to receivers at the detection positions. With the drum 15 on the hoist 14, the hoist is then lowered to lower the drum 15 into the containment until the drum lid 16 is level with or just below the lip 6 as shown in FIG. 3. This is indicated by the detector 8 which initiates a signal to interrupt the hoist drive mechanism. The sphincter seal 3 remains at all times in engagement with the drum 15. During normal operation, a drum is always present in the port. This condition is only changed either at commencement of operation or if it becomes necessary to change and renew the sphincter seal. During normal operation a second drum 17, FIG. 4, is then lowered by the external crane or hoist to above the drum 15. The drum 17 on breaking the signal at the detector 7 initiates a signal to recommence the hoist drive mechanism. The hoist 14 retracts to lower the drum 15 on to the roller conveyor assembly 12 and the fully withdrawn position of the hoist is indicated by a signal from detector 11 which stops the hoist drive mechanism. The drum 17, still supported by the external crane or hoist is at the same time lowered into the containment 1 and stops when the top of the drum 15 clears the detector 9. This position is shown in FIG. 5. The sphincter seal 3 is now in engagement with the surface of the drum 17. In FIG. 6, the drum 17 is supported in the port by the external crane or hoist and the first drum 15 is moved off the roller conveyor assembly 12 and through an airlock door (not shown) into a work area within the containment. Thereafter, the airlock door is closed and the hoist 14 is raised again to its highest position as indicated by the detector 10. The drum 17 is lowered on to the hoist 14 and released from the external crane or hoist. This position is shown in FIG. 7. The drum 17 is then lowered on the hoist until the detector 8 is again actuated. The lid 4 can then be closed to form a secondary seal arrangement (FIG. 8). The drum 17 remains in this position, with the lid closed, until a further drum is ready for transfer through the port into the containment. The lid 4 is then reopened to allow the posting operation to take place. The sequence of operation is then repeated with the drum 17 and the further drum to be entered into the containment. The sphincter seal serves to reduce the opening between the interior of the containment and the atmosphere during posting operations such that the containment ventilation system can maintain an inward linear air flow, typically 1 m/sec, thus preventing back diffusion of airborne contamination from the containment. A preferred form of sphincter seal is shown in FIGS. 9 and 10. The seal 3 is an annular assembly comprising outer or upper and inner or lower brush seals 20 and 21, conveniently nylon brushes, between which are sandwiched rings of natural or synthetic rubber 22, each ring being divided into sectors. The rubber rings extend radially inwardly beyond the inner radius of the nylon brushes. A continuous elastic garter 23 is secured to the underside of the lowermost ring 22 projecting inwardly beyond the lower brush. As a drum is passed through the port the sphincter seal is deformed as indicated in FIGS. 1 to 8 but at all times maintains a positive contact with the drum over the periphery of the drum. The seal can accommodate changes in diameter of individual drums and also different sizes of drums. The garter 23 serves to resist deformation of the seal during use and ensures that the seal returns to its natural state after passage of a drum through the port.
060144181	description	The Figures the following reference numbers: 1 fuel rod 2 cladding tube 3 UO.sub.2 fuel pellet 4 end plug 5 spring 6 bead 7 heat affected zone 8 equiaxed structure 9 acicular structure 10 grain boundary DETAILED DESCRIPTION OF THE INVENTION The present invention may provide (1) A fuel rod for a light water reactor, comprising a cladding tube which comprises a zirconium alloy containing Nb and Fe; uranium oxide fuel pellets packed in said cladding tube; and end plugs comprising a zirconium alloy and closing both ends of said cladding tube, said cladding tube being sealed with said end plugs by TIG welding, wherein grain boundaries in each heat affected zone of said cladding tube adjacent to a bead formed by TIG welding of said cladding tube with said end plug have structural compositions including 4 to 30% by weight of Nb, and 0.9 to 20% by weight of Fe. The present invention may also provide (2) a method for manufacturing a fuel rod for a light water reactor, comprising: packing uranium oxide fuel pellets into a cladding tube which comprises a zirconium alloy containing Nb and Fe; closing both ends of said cladding tube with end plugs comprising a zirconium alloy; and sealing by TIG welding said cladding tube together with said end plugs, wherein: each heat affected zone of said cladding tube which is adjacent to a bead formed by TIG welding said cladding tube with said end plug is cooled at a rate of 70.degree. C./sec. to 5.degree. C./sec. According to the present invention, the zirconium alloy containing Nb and Fe for the cladding tube to be used in the fuel rod for a light water reactor has a composition including 0.6 to 2.0% by weight of Nb, 0.5 to 1.5% by weight of Sn, 0.05 to 0.3% by weight of Fe, and the balance being Zr and incidental impurities. Preferably, the zirconium alloy has a composition including 0.8 to 1.2% by weight of Nb, 0.8 to 1.1% by weight of Sn, 0.08 to 0.12% by weight of Fe, and the balance being Zr and incidental impurities. Conventional Zircaloy-2 (JIS H4751ZrNT802D) or Zircaloy-4 (JIS H475 1ZrNT804D) is used for the end plugs which close both ends of the cladding tube comprising the zirconium alloy containing Nb and Fe since the end plugs do not greatly affect the life span of the fuel rod for a light water reactor, even if they are corroded. A fuel rod for a light water reactor of the present invention comprises a cladding tube which comprises a zirconium alloy containing Nb and Fe; uranium oxide fuel pellets packed in the cladding tube; and end plugs comprising a zirconium alloy and closing both ends of said cladding tube, where the cladding tube is sealed with the end plugs by TIG welding. Grain boundaries in each heat affected zone of the cladding tube which is adjacent to a bead formed by TIG welding have compositions including 4 to 30% by weight of Nb and 0.9 to 20% by weight of Fe. In some cases, where Cr is contained in the alloy as an incidental impurity, Cr can also segregate at grain boundaries and can be detected. In the fuel rod for a light water reactor according to the present invention, for example, the following methods of cooling while controlling the cooling rate at 70.degree. C./sec. to 5.degree. C./sec. may be employed: (i) TIG welding is carried out while both areas adjacent to each portion to be welded are covered with a heat insulating material, and following sufficient cooling after completion of the welding, the heat insulating material is removed; or (ii) after welding, the cladding tube and end plugs are subjected to induction heating or heating by direct heating with electricity. When the heat affected zones due to welding are cooled at a cooling rate of greater than 70.degree. C./sec., the concentrations of Nb and Fe at the grain boundaries fall below 4% by weight and 0.9% by weight, respectively, and sufficient corrosion resistance cannot be achieved. On the other hand, when the heat affected zone due to welding is cooled at a cooling rate of less than 5 .degree. C./sec., further improvement in corrosion resistance cannot be achieved, since the concentrations of Nb and Fe at the grain boundaries in the heat affected zone formed by the welding do not exceed 30% by weight and 20% by weight, respectively, even with a much slower cooling rate. On the contrary, such a slow cooling rate causes the strength of the fuel rod to deteriorate. EXAMPLES Example 1 Zirconium alloy cladding samples were prepared which had dimensions of 10 mm in diameter and 0.6 mm in thickness, having a composition including 1.0% by weight of Nb, 1.0% by weight of Sn, 0. 1% by weight of Fe, and the balance being Zr and incidental impurities. Each zirconium alloy cladding sample was TIG welded under the conditions described below at its ends with zirconium alloy end plugs which had a composition including 1.5% by weight of Sn, 0.2% by weight of Fe, 0. 1% by weight of Cr, and the balance being Zr and incidental impurities. The cooling rate at the heat affected zones of each cladding sample was controlled by a method shown in Table 1. As a result, Samples 1 to 6 according to the present invention, Comparative Samples 1 and 2, and Conventional Samples 1 and 2 were manufactured, in which the concentrations of Nb and Fe at grain boundaries in heat affected zones were as shown in Table 1, respectively. TIG Welding Conditions: Current: 30 A PA1 Voltage: 15 V PA1 Welding rate: 500 mm/min. PA1 Cooling gas: 25 liter/min. He PA1 Voltage: 20 V PA1 Current: 100 mA PA1 Temperature: -40.degree. C. PA1 Solution: 5% perchloric acid-methanol Samples 1 to 6 according to the present invention, Comparative Samples 1 and 2, and Conventional Samples 1 and 2, which were zirconium alloy cladding samples having heat affected zones, were subjected to chemical milling in a nitric-hydrofluoric acid solution [HNO.sub.3 :HF:H.sub.2 O=45:5:50 (% by volume)] to a thickness 100 .mu.m, and were cut into disks having a diameter of 3 mm. Subsequently, the disks were subject to electrolytic milling under the conditions described below to prepare foil samples for examination by Transmission Electron Microscopy. Electrolytic Milling Conditions: The above-obtained foil samples for examination by Transmission Electron Microscopy from Samples 1 to 6 according to the present invention, Comparative Samples 1 and 2, and Conventional Samples 1 and 2 were examined with an accelerating voltage of 200 kV and a magnification of 50,000, and no precipitates of intermetallic compounds were found. Additionally, contents of Nb and Fe at the grain boundaries were measured by Energy Dispersive X-ray Analysis, with the results shown in Table 1. Samples 1 to 6 according to the present invention, Comparative Samples 1 and 2, and Conventional Samples 1 and 2 were placed in an autoclave, and subject to autoclave tests in purified water having a high temperature of 360.degree. C. for 120 days in order to examine color change in the heat affected zones of the zirconium alloy cladding samples. The results are shown in Table 1. TABLE 1 __________________________________________________________________________ Nb and Fe Concentrations at Appearance after Autoclave Grain Boundaries Test Cooling Rate (% by weight) (in 360.degree. C. Pure Water for Sample Type & No. (.degree. C./sec.) Controlling Method of Cooling Rate Nb Fe 120 Days) Remarks __________________________________________________________________________ Samples of the 1 68 Removal of the Chilling Block 4.3 0.95 Black -- Present Invention 2 45 Covering with a Heat-Insulating material 5.2 2.7 Black -- 3 30 Direct Heat with Electricity 6.3 4.8 Black -- 4 22 Induction Heating of the Cladding Tube 11 7.3 Black -- 5 15 Induction Heating of the Cladding Tube and 15e 10 Black -- End Plug Portions 6 6.0 Induction Heating of the Cladding Tube and 28e 25 Black -- End Plug Portion Comparative 1 3.8 Induction Heating of the Cladding Tube and 27e 23 Black Insufficient Samples End Plug Portions Strength 2 1.5 Induction Heating of the Cladding Tube and 25e 25 Black Insufficient End Plug Portions Strength Conventional 1 100 Natural Cooling 3.1 0.50 White (Peeled) -- Samples 2 80 Decreasing the Flow of Cooling Gas 3.3 0.65 White -- __________________________________________________________________________ As shown in Table 1, each heat affected zone of Conventional Samples 1 and 2, which were allowed to cool or were cooled with a cooling rate close to the natural cooling, turned white and had inferior corrosion resistance. On the other hand, each heat affected zone of Samples 1 to 6 according to the present invention and Comparative Samples 1 and 2, which were cooled while controlling the cooling rate at 70 to 5.degree. C./sec., and in which the concentrations of Nb and Fe at the grain boundaries in their structure were 4 to 30% by weight and 0.9 to 20% by weight, respectively, had a black color and had satisfactory corrosion resistance. Comparative Samples 1 and 2 with heat affected zones which cooled at a cooling rate of less than 5.degree. C./sec., were undesirably softened and had insufficient strength. As described above, the fuel rod for a light water reactor according to the present invention, which has improved corrosion resistance as compared to conventional rods, allows for highly efficient and highly reliable operation, and therefore, greatly contribute to the development of the atomic industry. Obviously, numerous modifications and variations of the present invention are possible in light of the above teachings. It is therefore to be understood that within the scope of the appended claims, the invention may be practiced otherwise than as specifically described herein. The priority document of the present application, Japanese Patent Application No. 08-211281, filed on Aug. 9, 1996, is hereby incorporated by reference.
051951215	abstract	An apparatus for modulating the intensity of an X-ray beam generated by an X-ray tube. The apparatus comprises a set of at least three blades made of a material opaque to X-ray. Each blade comprises a central hub and a pair of symmetrical wings extending away from the hub. All the blades have their hubs slidably mounted on radially projecting shafts symmetrically positioned about a rotor positioned adjacent the X-ray beam and driven at variable speeds. Matching pins and threads are provided on each pair of hub and shaft to cause the corresponding blade to pivot about the shaft on which it is mounted when the speed of the rotor increases and the blades are then moved in unison radially outwardly against the action of a return spring, because of the centrifugal force. The rotor, the blades and the shafts are positioned and sized to cause the blades to intersect the X-ray beam when the rotor is driven. This in turn causes the blades to modulate the intensity of the X-ray beam as a function of the angular position of the blades about their shafts, which allow more or less radiation to pass therebetween.
053295641	claims	1. In a nuclear reactor coolant system where a coolant pump receives coolant from a coolant tank and directs coolant through a coolant line to the reactor during normal reactor operation, a passive coolant system for removing decay heat from the reactor when the coolant pump is inoperative, said passive cooling system comprising: a. a plurality of coolant tanks in fluid communication with the coolant line via an inlet line and an exhaust line for each tank; b. a check valve in each inlet line; and c. flow control means in each exhaust line. 2. The passive cooling system of claim 1, wherein each of said flow control means provides for a different predetermined flow rate. 3. The passive cooling system of claim 1, wherein the flow rate of said flow control means is directly related to the decay heat rate of the nuclear reactor.
	abstract	The purpose of the present invention is to provide a method for manufacturing a three-dimensional structure, a method for manufacturing a scintillator panel, a three-dimensional structure, and a scintillator panel that enable the type and thickness of a substrate of the scintillator panel to be selected freely. The present invention provides a method for manufacturing a three-dimensional structure, by which a three-dimensional structure is obtained by forming a glass pattern on a base member and then separating the glass pattern from the base member.
	summary
	summary
	summary
	description	This is a Continuation of U.S. Ser. No. 11/614,199 filed on Dec. 21, 2006 entitled “BONE MINERAL DENSITY ASSESSMENT USING MAMMOGRAPHY SYSTEM” in the name of Huo et al., which issued as U.S. Pat. No. 7,746,976 which claims priority to U.S. Provisional Patent Application No. 60/755,233, entitled “BONE MINERAL DENSITY ASSESSMENT USING MAMMOGRAPHY SYSTEM”, provisionally filed on Dec. 30, 2005 in the name of Huo et al., both of which are incorporated herein. The invention relates generally to the field of mammography imaging system. More specifically, the invention relates to a system for assessing Radiation Absorptiometry (RA) based BMD (Bone Mineral Density) using a mammography x-ray imaging system. Osteoporosis is a skeletal disorder characterized by reduced bone strength. It can result in increased risk to fractures, height loss, hunched backs, and pain. Bone strength is a function of bone mineral density (BMD) and bone quality. It is believed that bone mineral density peaks about the age of 30 for both men and women, and then declines gradually. Some statistics have indicated that osteoporosis affects approximately 20 million people and is a cause of about 1.3 million fracture incidents in the United States each year. As such, screening for bone mineral density is often desired. Several common techniques have been used to measure bone mineral density, including bone puncture, radiation absorptiometry of single energy x-ray systems, DEXA (dual energy x-ray absorptiometry), and sonography. Bone puncture can be an accurate but invasive procedure, which involves the extraction of bone mass from spine area. This procedure carries risk. With regard to single energy x-ray systems, mineral loss in a person's bones can be estimated from a single energy x-ray image of a body part. In diagnosing and treating bone diseases, it is common to take radiographic images of the patient (e.g., skeletal features of the patient), then either read the images directly or perform software analysis on the images to extract information of interest. For example, in diagnosing or monitoring the treatment of osteoporosis, one might take x-ray images of selected skeletal bones, then perform computer analysis on certain image features to determine bone volume, bone length, bone geometric changes, bone strength conditions, bone age, bone cortical thickness, and bone mineral mass. Typically for reading and interpreting radiographic images directly, the treating physician will refer the patient to a radiologist, who can supervise both taking the radiographic image and interpreting the image to extract desired bone information, such as bone mass and bone contour irregularities. Alternatively, if the bone analysis is done, at least partially, by a computer analysis system, the x-ray images prepared by the radiologist may be sent back to the treating physician's computer site or to another computer site for computer analysis. DEXA is a device used by the hospitals to measure bone mineral density (BMD). In DEXA, two low-dosage x-ray beams with differing energy levels are aimed at the patient's spine, hip or whole body using conventional x-ray machines. The computer calculates the content of bone mineral density based on the relationship that different bones absorb different energy levels. Some consider DEXA to be accurate, but the apparatus is bulky and expensive and results in more radiation to the patients. U.S. Pat. No. 6,816,564 (Charles, Jr.) is directed to a technique for deriving tissue structure from multiple projection dual-energy x-ray absorptiometry. Sonography devices measure the bone mineral density of peripheral bones, such as heel, shin bone, and kneecap. But it is recognized that the bone mineral density in the spine or hip change faster than that in heel, shin bone, or kneecap. Thus sonography is considered by some to be not as accurate or sensitive as DEXA in the determination of bone mineral density. DEXA allows early detection of abnormal change in bone mass for its targets spine, hip, or whole body. However, sonography offers advantages of lower cost and radiation-free. U.S. Pat. No. 6,246,745 (Bi) describes a software system for determining bone mineral density from radiographic images of a patient hand obtained from conventional x-ray imaging system. US Patent Application No. 2005/0059875 (Chung) describes a biosensor and method for bone mineral density measurement. US Patent Application No. 2005/0031181 (Bi) is directed to a system and method for analyzing bone conditions using DICOM compliant bone radiographic images. U.S. Pat. No. 5,712,892 (Weil), commonly assigned, is directed to an apparatus for measuring the bone mineral content of an extremity. While such systems may have achieved certain degrees of success in their particular applications, there is a need for a system and method for bone mineral density screening, particularly wherein a medical professional can readily and locally (e.g., at their office location) generate a bone mineral density report. A suitable system would be easy to use, reduced in cost, yet provide sufficient accuracy. Preferred would be an on-site screening that can be utilized by physicians, radiologists, or other medical professionals. An object of the present invention is to provide an imaging system and method to acquire hand x-ray images suitable for bone mineral density (BMD) screening and analysis using a mammography x-ray imaging system. Any objects provided are given only by way of illustrative example, and such objects may be exemplary of one or more embodiments of the invention. Other desirable objectives and advantages inherently achieved by the disclosed invention may occur or become apparent to those skilled in the art. The invention is defined by the appended claims. According to one aspect of the invention, there is provided a mammography x-ray imaging system adapted to acquire hand images with sufficient image quality for the assessment of BMD by a computer-aided system. According to one aspect of the invention, the system includes an x-ray generator, x-ray source/target, filtration, x-ray detector, and a template for positioning the hand. According to another aspect of the present invention, there is provided a method of positioning the hand to obtain hand images with sufficient image quality for bone mineral density assessment. According to another aspect of the present invention, there is provided a preferred range of kVp (x-ray energy) and mAs (exposure) for a given target/filtration (built in a mammography x-ray image system) combination to obtain sufficient quality hand images when mammography screen/film systems are used as an image detector. According to another aspect of the present invention, there is provided a method of converting analog images to digital images for computer analysis. A film digitizer with a preferred dynamic range can be employed to convert analog images to digital images for the computer analysis. An image of a body extremity is acquired using a mammography x-ray system whereby a bone mineral density assessment can be performed on the image. The system for determining the bone mineral density of a body extremity includes: a support for supporting the body extremity; a detector for capturing an image of the body extremity; and an x-ray source adapted to project an x-ray beam through the body extremity toward the detector, the x-ray source having a voltage of no more than about 45 kVp and having a target/filter combination of rhodium/rhodium, molybdenum/molybdenum, molybdenum/rhodium, or tungsten/rhodium. The following is a detailed description of the preferred embodiments of the invention, reference being made to the drawings in which the same reference numerals identify the same elements of structure in each of the several figures. It is noted that the American Cancer Society recommends that women over the age of 40 years obtain annual mammograms. Millions of women have their annual screening mammograms each year at hospitals or breast imaging centers. Accordingly, Applicants have noted it would be desirable for women to have both their annual mammography screening and a bone mineral density screening done in one visit, at one location, and using one imaging system. Conventionally, extremities (e.g., hands and feet) are imaged using conventional x-ray system, which generates an x-ray beam adapted to capture both low and high-density objects (i.e., bone and soft tissue) on a detector (film or digital) that are designed with a wide dynamic range. In contrast, mammography imaging systems are configured for high contrast (i.e., narrow dynamic range) to image the soft tissue in the breast for the purpose of detection and diagnosis of breast cancer. The present invention is directed to a system and method for acquiring hand images using a mammography x-ray imaging system. A treating physician or a computer-aided system can then analyze the acquired images for bone mineral density (BMD) loss assessment. It is intended that the use of a mammography imaging system to acquire hand images for assessing BMD can improve the workflow and access for women to BMD exams, so as to reduce the cost and improve the efficiency of screening. Conventional x-ray imaging systems have been employed to image various human body parts (e.g., head, neck, chest, abdominal, and extremities) to detect and diagnose various diseases. Because of the bone structures and thick body part, high-energy x-ray is required to provide sufficient penetration. Also, a wide range of x-ray energies (for example, from 50 kVp-140 kVp, dependent on the selection of kVp) is available to provide a suitable x-ray photon energy level (kVps) when imaging different body parts. Tungsten targets are typically used in conventional systems to meet the needs of a wide-range of high energy x-rays. Generally, thicker and/or denser body parts require higher x-ray energy to provide sufficient penetration to achieve desired image quality while keeping the patient dose at minimum. For example, 50-60 kVps are typically employed for extremities, 70-90 kVp for hip and skull, and 100-130 kVp for chest. The signal strength (i.e., the amount of x-rays) reaching the detector can vary for a given x-ray energy. The weakest image signals are typically behind or within the dense and thick body parts (i.e., high attenuation), such as bone or abdomen. The strongest image signals reaching the detector are in the area of cavities or thin body part (i.e., low attenuation), such as the clear lung area or low-density soft tissue. However, there is a non-linear relationship between x-ray exposure (i.e., amount of x-rays) to the x-ray film and the film optical density (OD). Thus, neither underexposure (e.g., not enough penetration) or overexposure (e.g., too much x-rays penetrated the film) are desirable. FIG. 1 illustrates the relationship between the x-ray exposure (the amount of x-rays reaching the screen/film) and the film optical density (OD). The curve is usually called the H&D curve which characterize the uniqueness of a screen/film system in its response to x-ray exposure. Two curves are shown in FIG. 1: one for a screen/film mammography system and one for a conventional screen/film x-ray imaging system. The two curves illustrate a difference in latitude 102 between the conventional and mammography screen/film systems, referenced as 102-C and 102-M, respectively. More particularly, FIG. 1 shows characteristics curves for a mammography screen/film system (e.g., high contrast, narrow latitude) and a conventional screen/film system (e.g., low contrast, wide latitude). Latitude is defined as the wide linear range of the optical density over the exposure. That is, latitude refers to the range of relative exposure that will produce optical density within the accepted range for detection and diagnosis. Information captured on the shoulder and beyond is referred to as overexposed 104, while information captured on the toe is referred to as underexposed 103. The wider latitude in conventional screen/film reduces the likelihood of overexposure or underexposure of films on the shoulder and toe. Such overexposure and underexposure are considered to be undesirable image quality. When undesirable images occur, a retake of the image is required to capture sufficient information for detection and diagnosis. Underexposure of the conventional and mammography screen/film systems is shown in FIG. 1 as 103-C and 103-M, respectively. Overexposure of the conventional and mammography screen/film systems is shown in FIG. 1 as 104-C and 104-M, respectively. An x-ray film digitizer can be employed to convert an analog x-ray image to a digital image. Such x-ray film digitizers are well known. Still referring to FIG. 1, only the signals (exposures) within a limited range (latitude 102) are visible or recognizable by an x-ray film digitizer. Because of the wide range of densities in the body parts, the range of image signals and their strength which reach the film is considerably wide. Therefore, conventional x-ray films are design to provide wide enough latitude 102 to properly capture a wide range of image signals on the film. Mammography systems are designed to capture x-ray breast images, particularly for the detection and diagnosis of breast cancer. Since the attenuation or density differences in the different parts of breast tissues are small, mammography systems employ x-ray equipment and detectors specially designed to optimize breast cancer detection. Using low x-ray photon energies generally will provide better differential attenuation between the soft tissues than using higher energy x-ray photons (50 kVp and above). However, low x-ray energy has a high absorption and therefore delivers a relatively high dose. Screen/film systems used in mammography are also designed to maximize the contrast for the captured image signals and require a certain of amount of radiation to ensure sufficient image quality for the cancer detection task. Referring to FIG. 2 there is shown the H&D curves of a screen/film system for mammography and a screen/film system for conventional x-ray imaging system, illustrating the difference in contrast 201 between the conventional and mammography screen/film systems, referenced as 201-C and 201-M, respectively. More particularly, FIG. 2 shows characteristic curves for a mammography screen/film system (e.g., high contrast, narrow latitude) and a conventional screen/film system (e.g., low contrast, wide latitude). The graph illustrates the difference in film contrast between the two types of screen/film systems for an object with the same object contrast. FIG. 2 was obtained by projecting object 1 (a portion of aluminum step wedge) within the proper exposure range for the mammography system, while the entire step wedge was captured within the wide dynamic latitude of the conventional screen/film system. Minimizing the dose while providing sufficient image quality to enhance the low contrast detection imposes extreme requirements on mammographic equipment and detectors. Because of the risk of ionizing radiation, some prefer to minimize the dose and optimize the image quality. These concerns have led to the refinement of dedicated x-ray equipment, specialized x-ray tubes, compression devices, and/or optimized detector systems. The imaging requirements impact the design of the x-ray tube, peripheral mammographic equipments, and film/screen detectors. X-ray tubes designed specially for mammography provide can provide a nearly optimal x-ray spectrum for a good subject contrast of soft tissues while maintaining a radiation dose as low as possible. Some experiments have shown that for a tissue thickness having a 3-6 cm range (e.g., typical compressed breast thickness), preferred x-ray energies are typically generated through a molybdenum target with a kVp range between 24 to 32 kVp. A maximum tube voltage of the x-ray source for mammography is approximately 45 kVp. The x-ray source is defined by a target and filter combination (sometimes referred to as target/filter). Examples of target/filter combinations includes molybdenum/molybdenum, molybdenum/rhodium, rhodium/rhodium, and tungsten/rhodium. The contradictory requirements of high subject contrast and low radiation dose are difficult to accomplish, and indicate mono-energetic x-rays as the best choice. However, x-ray energy of the x-ray tube is poly-energetic. Referring to FIG. 3, there is shown a system in accordance with the present invention for obtaining hand images for BMD assessment using a mammography system. A source 301 emits x-ray energy 304 directed toward an object to be imaged (shown as fingers/hand in FIG. 3). A hand template 307 can be provided to properly position the object. One or more phantom/calibration/step wedges 308 can be positioned proximate the object made of a material (e.g., aluminum) to approximate the density variations of a human extremity. Template 307 and wedge 308 can be positioned by frame 306. Referring to FIG. 3, built-in filtration 302 is added to the x-ray tube to remove some of the low energy x-rays as the low energies contribute to tissue dose without contributing significantly to image formation. The highest x-ray energies in the x-ray beam are a function of the peak operating voltage (kVp) applied and added filtration at the x-ray tube port. A fixed amount of filtration (for example, molybdenum/rhodium) is built in the system to remove the low-energy x-rays. A rotating anode design is used for some mammographic x-ray tubes. Molybdenum targets are common, although tungsten is used in many tubes. A dual track molybdenum/rhodium target and molybdenum/rhodium filtration is used by a manufacturer. A combination of target and filtration determines the x-ray beam quality, which in turn determines the image quality (contrast and exposure level). The x-ray beam quality determined by the combination of the target and filtration in conjunction with the x-ray energy range (operating kVp range) provide a distinction between mammography x-ray systems and conventional x-ray systems, and this determines the difference in their applications. With regard to the image detector, the beam quality determines how much signals can penetrate through different body parts with different thickness, reaching the detector. The characteristics of a detector (FIG. 1 and FIG. 2) determine the way in which the signals are captured and presented in an analog and/or digital format by the detector. Detectors are customized to the need for conventional x-ray systems and for mammography x-ray systems. As was shown in FIG. 1, a conventional screen/film system has a wider latitude 102-C to capture the wide range of signals in strength for imaging head, neck, chest, abdominal and extremities. Hands can be imaged using conventional x-ray machines. The high x-ray energy in conventional x-ray machine provides sufficient penetration to see the details in the bone while the wide latitude of the screen/film system allows capture of both the soft tissues (low attenuation) and bone (high attenuation) in one image. As was shown in FIG. 1, information captured on the shoulder and beyond is referred to as overexposed (104-C and 104-M), while information captured on the toe is referred to as underexposed (103-C and 103-M). Both overexposure and underexposure are not desirable. The wider latitude in conventional screen/film reduces the likelihood of overexposure or underexposure of films on the shoulder and toe. Mammography screen/film systems have a higher contrast than conventional screen/film systems. This was illustrated in FIG. 2, where the difference in optical density on the vertical axis is used to measure the film. For example, an aluminum step wedge is an object having varying attenuation. The thicker the step wedge, the more x-rays are attenuated, and accordingly, the less x-rays reach the film. For a given object contrast (difference in exposure along horizontal axis in FIG. 2), the film contrast 201 for mammography screen/film is higher than that for conventional screen/film system. While the high contrast in mammography screen/film detectors assists to signify the small difference among the soft tissues in a breast, the narrow latitude 102 allows it to capture the information from a portion of the aluminum step wedge. It is difficult to capture both soft tissues and high-density bone structure on one single image. One can increase or decrease the x-ray energy and x-ray intensity to project the desired portion of the aluminum step wedge into the narrow latitude. For example, one can increase the x-ray energy to increase the exposure to the film under the thicker part of the aluminum, so that the thicker part of the aluminum got exposed properly within the narrow latitude. Note that when the thicker part of the aluminum gets exposed properly, the portion of aluminum at the thin end likely gets overexposed. This indicates that overexposure of soft tissue or underexposure of bone is likely to occur when a mammography screen/film detector is used. It is well known to use a cassette to hold a screen/film for x-ray imaging. If the screen/film employs a phosphor storage phosphor material (such as used for computed radiography), the storage phosphor screen/film can be disposed within a cassette for imaging. FIG. 3 generally shows a cassette 310 which can house computed radiography plate 311 (having a storage phosphor layer). Cassettes designed for mammography can include some particular attributes. Some cassettes are made of a low attenuation carbon fiber and/or have a single high definition screen used with a single emulation film. Because of the difference in the requirement for the x-ray beam quality between mammography and conventional x-ray image systems, and in the requirement for characteristics of screen-film detectors between the two systems, it is difficult to image hands using a low-energy (i.e., mammography) x-ray beam quality system to get sufficient penetration of the bone while not overexposing the soft tissues in the hand. For an accurate assessment of BMD using hand x-ray images, a sufficient amount of soft tissue needs to be visible as well as the detailed bone trabecular structures. Overexposing the soft tissue or under exposing the bone structure on the films can easily occur if proper techniques are not used to image the hands. These techniques include the right choice of target/filtration combination, additional filtration in-between the target to imaged object (hand), additional attenuation material in-between imaged object to the image receiver, speed and latitude of screen/film system. Since mammography is particularly designed for breast imaging, the limits or constraint set on how the system can be used does limit the choices of the kVp and mAs combinations along with other factors to capture the right exposure for hands. For example, 1) the maximum x-ray energy from mammography system is set at 45 kVp for rhodium target and 35 kVp for molybdenum target. This constriction limits the choices to use higher energy x-rays to get a good penetration of bone. 2) A low limit set on the output (x-ray intensity) for mammography x-ray units is rather high (4 or 5 mAs), it often causes overexposure to the soft tissue on hand image. 3) The distance from the source (target) to image receiver on mammography system is often fixed (65 cm), so the x-ray radiation cannot be lowered by increasing the distance, which is a way often used in conventional x-ray system to lower intensity of the x-ray radiation reach the film. A suggested distance between target and image receiver for imaging hands is 40 inches. As mentioned above, a high kVp is employed to obtain a good penetration of the hand. However, when kVp increases for a selected mAs setting, the radiation output (exposure to the film) increases as more x-rays are able to penetrate the object. To avoid overexposure of the soft tissue, it is desired to reduce the amount of the exposure to the film by reducing the mAs and increasing the distance from the source to the detector. As a result of these limitations on the high-end kVp, minimum mAs and the fixed distance from source to receiver on the mammography systems, when good penetration is obtained, soft tissue is often overexposed. Conversely, when soft tissue is appropriately imaged, sufficient penetration of bone cannot be reached. These limitations aggregated the problem in finding the right techniques to appropriately image the hand using mammography x-ray image systems and/or mammography screen/film detectors To address the various problems discussed above, Applicants have replaced the mammography screen/film detectors with digital detectors, that is, direct digital radiography (DR) and computed radiography (CR) designed for mammography. These detectors generally have a wide dynamic range of the pixel value over the exposure levels. With a combination of high x-ray energy (kVps) and mAs on the mammography machine, a suitable image quality for both soft tissue and bone detail has been obtained. Note that cassettes for mammography can be used to hold a CR screen when CR detectors are used. Further, the mammography screen/film system was replaced with low speed (<150) conventional screen/film detectors. For example, film designed for general radiography, such as Kodak X-sight G/RA, X-sight L/RA film, TMAT G/RA film or TMAT L/RA film or Insight film family. These detectors, as mentioned above, have wider latitude than mammography screen/film system. The screen/film were placed into a mammography cassette and positioned for imaging (for example, inserted into a bucky). Using the x-ray energy of a mammography system and the conventional screen/film system allowed Applicant to generate hand images with sufficient good image quality for BMD assessment. Further, mammography films can be replaced with conventional radiography films (with speed of slower than 400). Thus the combination of a mammography screen and a conventional radiography film was used to acquire hand images. Further, the screen/film configuration in terms of its position relative to incoming x-rays direction was investigated. In mammography, a single back screen configuration (i.e., placing the film between the x-ray source and the screen) is often used to maximize the image sharpness and the efficiency of the screen in converting absorbed x-ray energy to light. For imaging a hand using mammography, a single front screen configuration can be used to reduce the x-ray to light conversion efficiency of the screen to avoid overexposure to the hand tissue. That is, the screen is placed between the x-ray source and the film. Examples of a mammography screen/film system which can be used in the front screen configuration are Kodak MinR screen or MinR 2000 screen or MinR 2190 screen or MinR 2250 screen or MinR EV screens with MinR-L film or MinR 2000 film or MinR EV film. Screens designed for conventional radiography such as Lanex fine screen, Lanex medium screen, Lanex regular, Lanex Fast (Lanex screen family) or Insight screen family can replace the screen designed for mammography in either a back screen or front screen configuration. Some images acquired in the configurations described above have been reviewed using an available computer-aided BMD assessment system. (Note the computer-aided assessment system was originally designed to assess the BMD using the hand images acquired from conventional x-ray machines with the conventional screen-film system (U.S. Pat. No. 6,246,745)). The computer-aided assessment system relates to input images acquired using x-ray energies of 50 kVp and above and background optical density (OD) on the film of 1.0−/+0.1. The background OD obtained from the above configuration is higher than 1.0 because the constraints set on kVp, mAs, distance from the source to detector and the choice of the target/filtration combination. From the results, Applicants believe that low energy x-ray beams from mammography x-ray systems when combined with a detector with sufficient wide latitude in its H&D curve can generate hand images with adequate image quality for BMD assessment. Applicants investigated using mammography screen/film systems as detectors to capture hand images for BMD assessment recognizing that there are limitations on the mammography x-ray imaging systems and the property of mammography screen/film designed breast imaging, and that there are numerous combinations of multiple variables (choice of target/filtration, added material for additional filtration, kVps, mAs, screen/film combinations). Mammography screen/film system were categorized into two categories, high and low contrast. Hand phantom images with sufficient penetration of bone and sufficient soft tissue required by the software can be achieved with selections of target/filtration combination, kVp and mAs, positioning of hands, additional filtration, and types of screen/film. At 4 mAs setting when the images were acquired using a high-contrast screen/film (Kodak MinR EV system), the acceptable kVp range (which generates acceptable image quality for BMD assessment) is between 31 and 35 for molybdenum/molybdenum target-filtration combination, is between 30 and 34 for molybdenum/rhodium combination target-filtration combination, and is between 29 and 32 for rhodium/rhodium combination target-filtration combination. When increasing mAs setting from 4 mAs, the acceptable kVp range was shifted to a lower kVp range. Both the lower and upper bound kVp values reduced a rate of 1 or 2 kVp per mAs. This results since the radiation exposure increases as the mAs increases at a given x-ray energy setting (kVp). An acceptable kVp range may get smaller at a higher energy as there is minimum kVp setting for each target/filtration combination. The minimum kVp for molybdenum/rhodium is 24. The minimum kVp for rhodium/rhodium is 25. When a low-contrast mammography screen/film system was used (e.g., Kodak MinR L), the acceptable kVp range at each corresponding mAs setting is wider than that for the high contrast screen/film system. However, the minimum acceptable kVp at each mAs setting is similar to that of the high-contrast screen/film system. Thus minimum x-ray energy (kVp) is required, regardless of the type of screen/film combination, to have the penetration of bone required for BDM analysis. Further, additional filtration material can be placed between the hand and the x-ray source target to reduce the amount of x-ray radiation, thus avoiding overexposure for the soft tissue, especially at higher kVps. While this can reduce the output to avoid overexposure of soft tissue by adding filtration additional to the built-in filtration, this can further increase the penetration of the bone by using higher energy x-rays. A wider range of kVps can be employed to generate the hand images with sufficient image quality, thus increasing the robustness of its implementation. It is known that different mammography x-ray units are calibrated differently. In addition, the thickness and size of hands can vary. A wider range of kVps can increase the robustness of the system to generate images for accurate assessment of BMD by the computer system. The additional filtration materials can be aluminum of thickness between 0.02 to 12 mm, polymethyl methacrylate of thickness between 0.5 to 120 mm, copper 0.001 to 0.4 mm or other material with the thickness ranges that can provide the x-ray intensity attenuation from 5% to 99.9% of the x-ray intensity without additional filter material. Further, additional material can also be placed between the hand and the screen film system to further attenuate the amount of x-rays to avoid overexposures to the film. This can also be achieved by using thicker or high attenuation material for the front cover of the screen film cassette. Use of grid in the Bucky can be applied to reduce the x-ray exposure to the film. When hand images are acquired using conventional x-ray units along with a conventional screen/film system, a desktop scanner can be employed. For images acquired using mammography x-ray units, a x-ray film digitizer with a wide dynamic latitude can be used to read the wide range of information captured on the mammography film. The sufficient dynamic latitude is defined as the Dmin (minimum optical density) and Dmax (maximum optical density) recognizable by the digitizer. Dmin of 0.2 and Dmax of 4.0 have been recommended to capture the details required for the computer analysis. When acquiring the images, the patient stands on the chest wall side of the mammography system facing the gantry. One hand is positioned on a template with an aluminum step wedge placed proximate the thumb and the index finger (for example, refer to U.S. Pat. No. 6,246,745 (Bi)). The hand is positioned to lay flat on the template. As shown in FIGS. 3 and 5, frame 306 with the hand template 307 placed at the bottom and the additional filtration material placed on the top can be used to position the hand when additional filtration is required to get adequate image quality. Although, the added materials for additional filtration 305 is preferred to be placed in between anode/target and imaged object, another way to add filtration is to place the material on the breast compression plate attached to the mammography x-ray unit or simply use the compression plates. Another way is to mount the filter material on a support that can be attached to the mammography machine the same way as the compression plate does. Another way is to attach the add-on filtration 303 on the x-ray exit window of the x-ray tube. With reference to FIG. 4, additional filtration 303 can be added proximate x-ray source 301. Alternatively, as shown in FIG. 5, additional filtration 305 can be added proximate frame 306. Still further, as shown in FIG. 6, additional filtration 309 can be added to cassette 310. Combinations of these can also be employed. For example, FIG. 3 shows the use of added filtration 303, 305, and 309. The acquired hand images can be evaluated by a computer or a treating physician for BMD assessment purpose. The image quality can be evaluated by a computer system. The system uses a step wedge for a calibration purpose. The system can employ a test procedure to assess if sufficient image quality (sufficient soft tissue and adequate penetration of bone) is achieved. Other systems may have different requirements in image quality. However, the calibration for adequate penetration and/or sufficient soft tissue ensures the accurate assessment of BMD. The present invention provides a method to acquire digital hand images using mammography x-ray system for the purpose of BMD assessment. Thus, the present invention provides a method of acquiring hand x-ray images. According to one aspect of the present invention, the method comprises the steps of generating a digital or analog x-ray radiograph of human hand using mammography imaging system. A cassette and screen is used wherein having a MO/MO or MO/Rh target/filter combination (MO being molybdenum, and Rh being rhodium) with exposure level lower than 10 mAs. In one arrangement, a phosphor screen/film combination is inserted into a cassette to hold screen and film for x-ray imaging. According to another aspect of the present invention, there is provided a system to acquire hand images on a radiograph using a mammography x-ray system. The system includes x-ray generator with rotating targets and filtration materials designed for mammography, image receivers and additional filtration.
	claims	1. A heat exchanger for a passive residual heat removal system to cool a core of a nuclear reactor, the heat exchanger comprising:an emergency cooling tank in which water is received;a housing disposed inside the emergency cooling tank, the housing including an upper portion and a lower portion, wherein an exhaust port is formed in the upper portion of the housing, and through holes are formed at the lower portion of the housing; anda heat exchange tube disposed inside the housing,wherein the heat exchange tube includes a first member positioned corresponding to the upper portion of the housing and a second member positioned corresponding the lower portion of the housing, the first member being connected to a steam pipe through which steam generated from a steam generator of the nuclear reactor circulates, and a second member connected to both of the first member and a feed water pipe used to supply water to the steam generator provided in the nuclear reactor, and the first member has a shape different from that of the second member, andwherein the water is filled to a level of the lower portion of the housing and flows between the emergency cooling tank and the housing through the through holes, and air is filled above the water, so that the steam flowing the first member is air-cooled and the steam flowing the second member is water-cooled. 2. The heat exchanger of claim 1, wherein the first member includes a spiral tube, and the second member includes a straight tube. 3. The heat exchanger of claim 1, wherein the first and second members include spiral tubes, and the first member has spiral turns per length greater than those of the second member. 4. The heat exchanger of claim 1, wherein the first member includes a first spiral tube, and the second member includes a straight tube and a second spiral tube connected to the straight tube. 5. The heat exchanger of claim 4, wherein the first spiral tube has spiral turns per length greater than those of the second spiral tube. 6. The heat exchanger of claim 1, wherein the housing has a lower end. 7. The heat exchanger of claim 1, wherein a heat radiation fin is installed in the heat exchange tube.
	summary
	abstract	The invention makes possible to increase the degree of radiation focusing by the lens, to use particles of higher energies, and to increase the coefficients, depending on these factors, of the devices, the lens is used in. Thus the sublens 18 of the least degree of integration represents a package of the channels 5, which is growing out of joint drawing and forming the capillaries, which are laid in a bundle. The sublens of each higher degree of integration represents a package of sublenses of the previous degree of integration, which is growing out of their joint drawing and forming. The sublenses are growing out of performing the said operations at the pressure of the gaseous medium inside the channels being higher than the pressure in the space between the sublenses of the previous degree of integration and at the temperature of their material softening and splicing the walls. To produce the lenses a bundle of stocks (capillaries) in a tubular envelope is fed to the furnace (at the first stage) or stocks, produced on the previous degree, and the bundle is drawing from the furnace at the speed, exceeding the speed of feeding. The product is cut off on stocks for the next stage, and at the final stage the product is formed by varying the drawing speed, after what the parts with formed barrel-shaped thickenings are cut of.
	claims	1. A method for warming a rotational interface in an ion implantation environment, the method comprising:providing a scan arm configured to rotate about a first axis and an end effector rotatably coupled to the scan arm via a twist motor and configured to selectively secure a workpiece, wherein the end effector is further configured to rotate about a second axis having a bearing and a seal associated with the second axis and twist motor, wherein the first axis and second axis are positioned a predetermined distance apart;activating the twist motor;performing one of reversing the rotation of the twist motor after a predetermined time and reversing the rotation of the twist motor when the twist motor faults;determining whether the rotation of the end effector about the second axis is acceptable; andreciprocating the scan arm about the first axis when the rotation of the end effector is unacceptable, wherein inertia of the end effector causes a rotation of the end effector about the second axis. 2. The method of claim 1, wherein one or more of the bearing and seal comprise a ferrofluid. 3. The method of claim 1, wherein reversing the rotation of the twist motor after a predetermined time comprises successively incrementing a duration of the predetermined time. 4. The method of claim 1, wherein the rotation of the end effector about the second axis is determined to be acceptable when the end effector rotates a predetermined amount. 5. The method of claim 4, wherein the predetermined amount comprises a rotation about the second axis of approximately 30 degrees. 6. The method of claim 1, further comprising translating the scan arm in a direction generally perpendicular to the first axis, wherein reciprocating the scan arm about the first axis generally avoids an intersection of the end effector with a process medium. 7. The method of claim 6, wherein the process medium comprises an ion beam. 8. The method of claim 1, wherein reciprocating the scan arm about the first axis is performed concurrently with a period of time when the ion implantation is not being performed, therein maintaining a relatively low viscosity of the bearing and/or seal. 9. A method for warming a rotational interface in a chilled ion implantation environment, the method comprising:providing an ion implantation system configured to produce an ion beam;providing a workpiece scanning system configured to selectively pass a workpiece through the ion beam, wherein the workpiece scanning system comprises a scan arm configured to rotate about a first axis and an end effector rotatably coupled to the scan arm at a predetermined distance from the first axis, wherein the end effector is configured to selectively clamp the workpiece thereto, and wherein the end effector is configured to rotate about a second axis;chilling the end effector, wherein one or more of a bearing and a seal associated with the rotation of the end effector about the second axis are chilled, wherein one or more of the bearing and seal comprise a ferrofluid, and wherein chilling the end effector decreases an ability of the end effector to rotate about the second axis; androtating the scan arm about the first axis after chilling the end effector, wherein inertia of the end effector causes a rotation of the end effector about the second axis. 10. A method for warming a rotational interface in an ion implantation environment, the method comprising:providing a scan arm having an end effector rotatably coupled thereto via a twist motor and configured to selectively secure a workpiece, said end effector having a bearing and a seal associated with the twist motor, wherein one or more of the bearing and seal comprise a ferrofluid;activating the twist motor to rotate said end effector in a first direction; andselectively reversing the rotation of the twist motor after a predetermined time in order to maintain the rotatable coupling between said scan arm and said end effector, wherein activating the twist motor and reversing the direction of the twist motor are performed for a selected period of time for maintaining a predetermined viscosity of the ferrofluid. 11. A method for ion implantation, the method comprising:translating a workpiece across an ion beam via an end effector rotatably coupled to a scan arm;rotating the end effector about a first axis via a first axis motor, wherein a bearing positioned along the first axis rotatably couples the end effector to the scan arm, and a magnetic liquid seal positioned along the first axis further provides a protective barrier for the bearing; andproviding a predetermined amount of heat to the bearing and magnetic liquid seal via selectively rotating the end effector about the first axis, therein maintaining a minimum propensity of the end effector to rotate about the first axis, wherein selectively rotating the end effector about the first axis comprises a selective rotation and counter-rotation of the end effector about the first axis based, at least in part, on feedback from the first axis motor and a predetermined degree of rotation of the end effector about the first axis. 12. The method of claim 11, further comprising chilling the end effector, therein increasing a viscosity of the magnetic liquid seal, and wherein the predetermined amount of heat is increased to counteract the increase in viscosity of the magnetic liquid seal. 13. A method for scanning a workpiece in an ion implantation system, the method comprising:securing a workpiece to an end effector rotatably coupled to a scan arm via a twist motor, a bearing, and a seal positioned along a twist axis, wherein the seal generally provides a pressure barrier between an external environment and an internal environment associated with an internal region of one or more of the scan arm and end effector;translating the workpiece across an ion beam; andselectively rotating and counter-rotating the end effector about the twist axis, wherein a predetermined amount of heat is provided to the bearing and seal, therein increasing a propensity of the end effector to rotate about the twist axis;receiving feedback from the twist motor, wherein the selective rotation and counter-rotation of the end effector is based, at least in part, on the feedback from the twist motor; andreciprocating the scan arm about a first axis when the feedback from the twist motor indicates that the rotation and counter-rotation of the end effector is unacceptable, wherein the first axis and twist axis are separated by a predetermined distance, and wherein inertia of the end effector causes a rotation of the end effector about the twist axis. 14. The method of claim 13, wherein the seal comprises a magnetic liquid rotary seal assembly configured to provide a rotary, hermetic seal between the external environment and internal environment. 15. The method of claim 14, wherein the twist motor comprises a rotor configured to rotate about the twist axis and a stator that is generally fixed with respect to the scan arm, and wherein the magnetic liquid rotary seal comprises a ferrofluid disposed in an annular region between the rotor and stator. 16. The method of claim 13, wherein the end effector comprises a chilled electrostatic chuck configured to selectively electrostatically clamp the workpiece to a clamping surface thereof and to selectively cool the workpiece.
	claims	1. An optical system for extreme ultraviolet (EUV) lithography, comprisinga reflective optical element, including a substrate with a highly reflective coating emitting secondary electrons when irradiated with EUV radiation, anda source of electrically charged particles, which is arranged such that electrically charged particles are applied to the reflective optical element,wherein the source is a flood gun applying electrons to the reflective optical element as the only charge carrier compensation source. 2. The optical system according to claim 1, wherein the flood gun is arranged to apply the electrons only to a border region of the highly reflective coating of the reflective optical element. 3. The optical system according to claim 1, wherein the flood gun is arranged to apply the electrons to only the border and to cover completely the border of the highly reflective coating of the reflective optical element. 4. The optical system according to claim 1, wherein the flood gun is arranged to apply electrons to a surface of the reflective optical element with an angle larger than 45° to the surface normal. 5. The optical system according to claim 1, wherein the flood gun is a thermionic flood gun with a heating element, the heating element being a planar element. 6. The optical system according to claim 1, wherein the highly reflective coating is a multilayer system. 7. The optical system according to claim 1, further comprising at least one of: at least one further reflective optical element and at least one further electron source. 8. An illumination system for EUV lithography, comprising an optical system according to claim 1. 9. A projection system for EUV lithography, comprising an optical system according to claim 1. 10. An EUV lithography apparatus for EUV lithography, comprising an optical system according to claim 1. 11. An optical system for extreme ultraviolet (EUV) lithography, comprisinga reflective optical element, including a substrate and a reflective coating emitting secondary electrons when irradiated with EUV radiation, anda source of electrically charged particles, configured to apply the electrically charged particles to the reflective optical element,wherein the source consists essentially of a flood gun configured to apply electrons as charge carrier compensation for the emitted secondary electrons.
060977787	abstract	A gravity driven suction pump for a nuclear reactor condenser is described. The gravity driven suction pump is utilizes the potential energy of the condenser condensate to move condensible and noncondensible gases through the condenser. The gravity driven suction pump includes a drain line having a venturi section and a suction line extending from the condenser into the venturi section. The venturi section alters the flow velocity of condensate in the throat of the drain line and creates a pressure differential through the suction line to drain the noncondensible gases from the condenser.
	description	This application claims the benefit of U.S. Provisional Application No. 60/438,993, filed Jan. 10, 2003, the disclosure of which is hereby incorporated by reference in its entirety. Heightened security concerns worldwide have greatly increased the need for rapid and accurate detection and classification of controlled substances. Detection of controlled substances, with greater certainty, is especially critical at points of entry into countries, transportation hubs, sensitive facilities (e.g., nuclear power plants), and buildings. Securing airports from controlled substances such as explosives, nuclear material, drugs, pathogens, food and their component parts has been complicated by the large volume of luggage moved through airports each day. Currently, luggage is subjected to visual screening using x-ray and/or similar detection devices. This system relies on thousands of individual inspectors to detect all controlled substances and contraband and their component substances despite the efforts of individuals to disguise and hide such materials. However, detection and classification of explosives and certain other controlled substances, and their components, are not amenable to visual inspection either directly or through the use of x-ray machines. Furthermore, previous systems for detecting and classifying explosives and controlled substances are large, very expensive to operate, difficult and expensive to maintain, compromised in terms of speed and detection sensitivity, and result in a high rate of false positive detections. Various detection systems have been developed to address the need for mechanized detection of explosives and controlled substances. However, to be effective, such detection systems must employ powerful sources, which are quite large and emit potentially dangerous radiation. Attempts to shield workers and the general public from these radiation sources, while providing a reliable detection system of a sufficiently small size to have practical application, have not previously been successful. Accordingly, the radiation sources and the enclosures needed to shield operators and/or the general public from harmful radiation which have been developed to date are simply impractical for use in many facilities including airports and other sensitive locations. U.S. Pat. No. 5,144,140 (“the '140 patent”) is directed to an analyzer for detecting explosives having at least one source and several detectors having substantially the same spacing from the source. Col. 2, lines 3-5. According to the '140 patent, it is known that when nitrogen, a prime component of explosives, is irradiated with neutrons, it emits gamma rays at predetermined frequencies. Col. 1, lines 14-60. However, since clothing may contain nitrogen at levels that are not significantly different from explosives, analyzers may not be able to distinguish between clothing and explosives. Id. According to the '140 patent, optimizing the distance between the detector and source makes it possible to detect explosives in luggage to a “sufficiently high sensitivity.” Col. 1, line 64-col. 2, line 2. U.S. Pat. No. 5,076,993 (“the '993 patent”) discloses the use of high energy neutrons in systems to detect contraband to avoid the use of a large number of gamma detectors and to permit sequential interrogating of small volume elements of the object. U.S. Pat. No. 5,098,640 discloses the use of fast neutrons (e.g. 14 MeV) to induce prompt gamma rays from a target object. U.S. Pat. No. 4,882,121 also discloses the use of a 14 MeV neutron generator to irradiate an object suspected of containing an explosive to generate fast gamma photons which are analyzed to determine the nitrogen/oxygen ratio in the object. The ratio is compared to the ratio associated with the explosive to determine if the object contains an explosive. U.S. Pat. No. 5,606,167 discloses the use of a single neutron source to develop a total neutron cross-spectra of an object to identify elements of explosives. None of the prior systems for detecting explosives and controlled substances are designed to fit in an practically-sized enclosure, operate at a sufficiently low power level, and emit sufficiently low radiation levels to make the same suitable for use in public places, such as airports. Also, although the prior systems are capable of detecting particular explosives and substances, they do not reliably classify the data received to provide an indication of the type of substance detected. The present invention overcomes the limitations of the prior art by providing a relatively low power level system for detecting and classifying explosives and other controlled substances in a relatively small enclosure which effectively contains harmful radiation (e.g., stray neutron and gamma radiation) from reaching the environment outside of the enclosure. In a preferred embodiment, the system of the present invention for accurately detecting and classifying explosives and controlled substances includes a source/detector array including a plurality of sources and a plurality of detectors; a signal processor coupled to the source/detector array for processing data received from the detectors; a classifier coupled to the signal processor for classifying data received from the signal processor according to a plurality of algorithms; a maximal rejection classifier coupled to the classifier; and a declarative decision module coupled to the maximal rejection classifier for rendering a decision regarding the contents of the object. The exemplary apparatus of the present invention for detecting explosives and controlled substances preferably includes an enclosure, a shield layer disposed within the enclosure, a cavity disposed within the shield layer, a source/detection array disposed within the cavity, and a transport mechanism for moving objects through the cavity past the sources and detection array. Preferably, the cavity comprises one or more turns which preclude a straight line trajectory through the cavity. The enclosure can be in the shape of a rectangle, circle, triangle, square, or any other suitable shape. In one preferred embodiment, the shield layer prevents stray radiation from exiting the enclosure. In another preferred embodiment, the shield layer includes a shield layer (e.g., water, polyethylene) contained between the enclosure and the cavity. The cavity preferably has at least three connected segments. The cavity is provided with a plurality of relatively low intensity neutron sources for generating neutrons and a plurality of gamma detectors for detecting prompt gamma rays emitted by the object after irradiation with the neutrons. The neutron sources can be arranged in an array in close proximity to the object. In one embodiment, the neutron sources are disposed on at least two sides of the array. Preferably, the neutron sources irradiate an object with neutrons as the object is moved through the cavity past the detector array by the transport mechanism. The preferred method of the invention is to irradiate an object with neutrons, detect the gamma rays emitted by all substances contained within the object in response to the irradiation, determine the gamma counts, isolate the common eigen value signatures of the substances contained within the object, and use a maximal rejection hierarchy classifier to determine if a controlled substance is present in the object without interference by the presence of a confounding substance. In one preferred embodiment of the invention, the object is irradiated with relatively low intensity neutrons (e.g., each source of 107 neutrons/sec) from a plurality of neutron sources. The neutron sources are preferably pulsed simultaneously in order to maximize either the number of neutrons or intensity applied to the object. The neutron sources can also be pulsed sequentially to provide a spatial scan of the object. Explosives and other controlled substances detected and correctly classified by preferred methods of the invention include TNT, PETN, RDX, HMX, Ammonium Nitrate, Plutonium, Uranium, Drugs, and many potential confounders such as Nylon and Foods. For example, detected substances in the energy range of interest contain at least carbon, oxygen, and nitrogen having gamma counts in the following energy intervals: 4.43 MeV, 6.14 MeV, and 2.31 MeV. Other features and advantages of the present invention will become apparent when the following description is read in conjunction with the accompanying drawings. Referring to FIG. 1, a preferred system of the present invention is shown, including a source/detector array 100 having a plurality of sources 110 and a plurality of detectors 120. In one preferred embodiment, sources 110 are small pulse neutron generators capable of emitting neutrons at 14.7 MeV with an energy of about 107 neutrons/sec each. In another embodiment of the invention, the energy emitted by sources 110 can be varied depending on the desired level of sensitivity. Neutron generators can optionally emit low intensity neutrons in all directions from sources 110. Sources 110 emit neutrons which can penetrate an object 125 within source/detector array 100 (e.g., luggage). When neutrons penetrate object 125, they interact with atomic nuclei within substances contained within and on the outside of object 125 in two distinct ways: collision or absorption. In a collision process, the neutrons collide with an atom's nucleus and bounce off in an elastic or inelastic collision. In an absorption process, neutrons may enter the nucleus and be absorbed. In the absorption process, the resultant nucleus has excess energy from absorption of the neutron. In order to lower its excess energy, the nucleus emits “prompt” gamma radiation in manner characteristic of the particular material from which the nucleus is derived. For example, if the resultant nucleus is radioactive, it will decay by emission of particles and/or gamma radiation characteristic of that particular nuclide. The amount of gamma radiation emitted by the nucleus relates to the amount of energy absorbed by the nucleus. Detectors 120 can be any detector suitable for detecting gamma radiation (e.g., NaI(TI) or other suitable scintillation detector). Each detector determines the spectral density of prompt gamma radiation it receives producing a characteristic spectrum. In one embodiment, detectors 120 determine the spectral density of radiation received from the object in the energy range of about 0 to about 8 MeV. In a preferred embodiment of the invention, the spectral densities derived from each of detectors 120 are transmitted to signal processor 200. Signal processor 200 can isolate common signatures (e.g., eigen signatures) and independent signatures for the composition of the chemical substance of interest (e.g., explosives and/or controlled substances). The term “common signature” refers to energy levels characteristic of a known substance that can be isolated, for example, from spectral density data transmitted by detectors 120. The term “independent signature” refers to energy levels characteristic of an unknown substance that can be isolated, for example, from spectral density data transmitted by detectors 120. Preferably, signal processor 200 isolates common and/or independent signatures from spectral density data transmitted by each of detectors 120. In another preferred embodiment, signal processor 200 isolates common and/or independent signatures from spectral density data transmitted by one or more of detectors 120. In another preferred embodiment, the system provides an operator module 250 which can instruct signal processor 200 to search for and isolate signatures relevant to particular explosives or controlled substances and components thereof. For example, operator module 250 can be a computer terminal coupled to the signal processor for inputting parameters which define the explosives and/or controlled substances signal processor 200 will identify. Thus, if an operator wants to narrow the search parameters to search for anthrax and fissile material only, the operator can instruct the signal processor 200 accordingly using operator module 250. Preferably, the signature data isolated by signal processor 200 is transmitted to classifier 300. Classifier 300 preferably processes and analyzes the signature data in order to reduce the chance of misclassification of an explosive or other controlled substance through detection of confounders and/or background noise. The term “confounders” refers to substances that are not related to or incorporated in an explosive or a controlled substance. For example, a confounder may include a substance or material normally present in luggage (e.g., clothing, toiletries, and food) or a substance or material placed into the luggage deliberately to conceal or mask the presence of an explosive and/or a controlled substance. While prior explosive and controlled substance detection systems are capable of detecting the presence of components of explosives and/or controlled substances, such systems readily detect the presence of confounding material resulting in a high rate of false positive and/or false negative detection. The term “false positive” refers to incorrect detection of an explosive and/or controlled substance in an object while “false negative” refers to a failure to detect the presence of an explosive and/or controlled substance in an object. False positives result in greatly increased costs as objects may be needlessly manually searched, detained, or destroyed. False negatives may result in undetected explosives and/or controlled substances entering sensitive facilities resulting in increased danger to the public. Classifier 300 greatly reduces the incidence of false positives and false negatives by applying several analytical tools in parallel to the signatures isolated by signal processor 200. In one preferred embodiment, classifier 300 subjects the signatures to a combination of Principle Component Analysis, Wavelet Analysis, and Independent Component Analysis. Analytical tools (e.g., Principle Component Analysis, Wavelet Analysis, and Independent Component Analysis) can ensure that the feature vectors extracted from the signatures received from signal processor 200 are: (1) relevant to the classification task; and (2) eliminate signatures related to background noise or irrelevant substances or components. Preferably, classifier 300 significantly reduces or completely eliminates background noise. The combination of the signal processor and classifier can yield significant signal enhancement gain further reducing neutron source generator power and requiring less shielding. In a preferred embodiment, each signature received by classifier 300 from signal processor 200 is processed in parallel through one or more analytical tools (e.g., Principle Component Analysis, Wavelet Analysis, and Independent Component Analysis). The term “processed in parallel” refers to each signature being subjected to independent analysis by each analytical tool. Alternatively, each signature can be processed in series (i.e., by each analytical tool in turn). By using several analytical tools, classifier 300 greatly increases the probability that: (1) the system will not identify a confounder; and (2) the system will not miss a signature associated with the explosive and/or controlled substance material being sought. Analytical tools for use in the systems of the invention are described in, for example, the following references: S. Theodoridis, K. Koutroumbas, “Pattern Recognition,” Academic Press, 1999; S. Mallat, “A Wavelet Tour of Signal Processing,” Academic Press, 1998; A. Hyvarinen, J. Karhunen, E. Oja, “Independent Component Analysis,” John Wiley & Sons, 2001; S. Haykin, “Neural Network: A Comprehensive Foundation,” Prentice-Hall, 1999; and B. Scholkopf, C. Burges, A. Smola, “Advances in Kernal Methods: Support Vector Learning,” MIT Press, 1999. In another preferred embodiment, the several analytical tools used by classifier 300 result in one or more identifications of explosives and/or controlled substances. The resulting analysis from each analytical tool can be transmitted from classifier 300 to maximal rejection classifier 400 for determination of the identity of the components of the object. Preferably, maximal rejection classifier 400 subjects the analyses received from classifier 300 to a hierarchy of classifiers that incorporate neural network technology and other support vector machines. For example, maximal rejection classifier 400 can include an iterative algorithm for non-time sensitive target change detection for explosive material in a non-explosive object. Thus, the iterative algorithm can accurately distinguish small amounts of explosive material contained within non-explosive material while reliably rejecting confounders. In one preferred embodiment, the iterative algorithm can function in two distinct modes: running mode and learning mode. In running mode, wavelets procedures are used to identify high probability interest areas. FIG. 2 depicts an exemplary iterative algorithm in running mode. In reference to FIG. 2, signature is subjected to formation of a correlation function S1. Next, the algorithm performs wavelet decomposition of the correlation function and determines correlation levels S2. The signature signals constitute multiple peaks with various heights and features. These features are compared to a library of orthonormal signature functions S3 and their corresponding coefficients of known image and interference S4. Based on the comparison step S3, the signature can be matched to the signature for a known explosive or controlled substance S5. In another preferred embodiment, a deterministic correlation function is formed. In yet another preferred embodiment, a wavelet decomposition transform of the autocorrelation function can be formed to detect noise from the signature data and accentuate the peak of the autocorrelation function at the given scale. In this embodiment, additional contrast between different explosive and/or controlled substances can be achieved thereby decreasing the probability of error. FIG. 3 depicts an exemplary iterative function in learning mode. First, a correlation function of signature data is formed S6 followed by wavelet decomposition of the correlation function S7. The incoming signature function is compared with the signatures stored in the library of signatures S8. If the signature cannot be identified S9 (i.e., it is a new interference agent) by comparison to the library of signatures, a new eigen function corresponding to the new signature is formed S10 and a coefficient vector of the new data in new signal space is determined S11. This new information regarding the new signature is stored in the library of orthonormal signatures S12. In learning mode, the algorithm can sense, process, and add an unknown interference agent to the library of the algorithm. Utilizing these characteristic gamma radiations, and a knowledge of the unique chemical composition of explosives and other contraband materials, preferred systems of the invention can characterize materials contained within or on objects. The spectrum of gamma radiation detected by the gamma detectors in the cavity contain the characteristic energies resulting from the interactions of the neutrons and the nuclei of interest, e.g. nitrogen, oxygen and carbon. The data corresponding to the energy intervals of interest can be a function of the resolution and detector efficiency at the specific energy of interest. Using this information, each of detectors 120 can be calibrated to remove spectra background (e.g., radiation background) for each detector resulting in a greatly improved signal to noise ratio. Preferably, maximal rejection classifier 400 identifies which analysis conducted by classifier 300 is correct. For example, classifier 300 can provide an analytical result from each of the analytical tools used for analysis of the signatures provided by signal processor 200. Maximal rejection classifier 400 selects the correct analysis provided by classifier 300. This information is provided to declarative decision module 500. In one preferred embodiment of the invention, the decision is transmitted to operator module 250. Alternatively, the decision can be transmitted to a computer or network of computers. In another preferred embodiments, an alarm can be triggered by the decision, for example, if an explosive or controlled substance is identified. Referring to FIG. 4, a preferred apparatus of the present invention comprises an enclosure 150, a shield layer 155 disposed within the enclosure, a cavity 145 disposed within the shield layer 155, a neutron source/gamma ray detection array 100 disposed within the cavity 145, and a transport mechanism 140 for moving objects through the cavity 145 and past the source/detection array 100. Previous enclosures housing apparatus for detecting controlled substances have limited use because of the need to reduce or eliminate contamination of the public outside of the enclosure by the radiation (e.g., stray neutron and/or gamma radiation) generated during the detection process. These enclosures must be sufficiently large and contain sufficient shielding to prevent the escape of stray radiation. The need for large enclosures holding high energy density sources needed to detect small size substances significantly limits practical application of previous apparatus. The preferred enclosure of the invention can be significantly smaller due to the configuration of cavity and the neutron generator/gamma ray detectors in a manner which prevents escape of stray radiation. In a preferred embodiment, the cavity comprises one or more turns (e.g., turns 132 and 134) which preclude a straight line trajectory through the cavity. Since stray radiation cannot travel past these turns, the stray radiation cannot escape the apparatus via the cavity. The turns can be of any shape or configuration suitable for preventing the stray radiation from traveling in a straight line trajectory through the cavity. In one embodiment, the cavity comprises at least three connected segments. The segments are connected in any suitable order or at any suitable angle. Preferably, at least two of the segments are connected to at least another segments at an angle less than or equal to 90 degrees. Alternatively, at least two of the segments are connected to at least another segments at an angle more than 90 degrees. By reducing or eliminating travel of stray radiation through the cavity, the enclosure used in the preferred apparatus can be significantly smaller than the enclosures of previous apparatus. Enclosures for use in the invention can be of any suitable shape (e.g., rectangle, circle, triangle, square etc.). In one embodiment, the enclosure is rectangular. The rectangular enclosure is relatively compact, and, in one preferred embodiment, has dimensions of no more than 6 meters in length, 3 meters in width, and 3 meters in height, which is practical for installation in airports and other transportation facilities. Referring to FIG. 4, a shield layer 155 is provided between the cavity and the enclosure. The shield layer prevents stray radiation from exiting the cavity 145 and the enclosure 150. The preferred material in shield layer 155 preferably includes water, which is especially useful in blocking neutron rays. However, any suitable material for blocking neutron rays can be used in the shield layer. The apparatus of the invention preferably includes a source/detection array 100 for detecting the presence of explosives and controlled substances inside an object. Referring to FIG. 5, source/detection array 100 is disposed within cavity 145 and permits transport mechanism 140 to move object 125 through source/detection array 100. Source/detection array 100 can be of any suitable shape. Preferably, source/detection array 100 has four sides (top, bottom, right, and left) such that object 125 is surrounded on four sides when it enters source/detection array 100. Alternatively, source/detection array 100 can have one or more sides. Source/detection array 100 comprises a plurality of neutron sources 110 which can be arranged in close proximity to object 125 as it passes through source/detection array 100. Neutron sources 110 can be arranged so that an equal number of neutron sources 110 are disposed on at least two sides of detection array 100. In one preferred embodiment of the invention, neutron sources 110 are small pulse neutron generators capable of emitting neutrons at 14.7 MeV (each with an energy of 107 neutrons/sec). and the array 100 is provided with 10 neutron sources, five on each of two opposing sides. Alternatively, neutron sources 110 can be located at several locations along the cavity. Significantly, by using a plurality (e.g., ten in this embodiment) of less powerful (107) neutrons/sec sources, rather than one more powerful (1011) neutrons/sec source, the size of the present invention can be significantly reduced to a practical level, while maintaining required sensitivity to detect small size substances. Source/detection array 100 further comprises a plurality of gamma detectors 120 which can be arranged in close proximity to object 125 as it passes through detection array 100. In one embodiment, the gamma detector is a NaI(TI) or other suitable scintillation detector. Gamma detectors 120 can be arranged so that an equal number of gamma detectors 120 are disposed on at least two sides of detection array 100, around the respective neutron sources 110, or the detectors 120 can be provided on all four sides of the cavity. Preferably, gamma detectors 120 are sensitive to gamma photons emitted by carbon, nitrogen, oxygen, chlorine or any other chemical composition. In a preferred embodiment, detection array 100 comprises a mix of at least 100 gamma detectors 120. FIG. 5 shows a side cut-away view of source/detection array 100 and depicts object 125 on transport mechanism 140 inside source/detection array 100 and aligned with neutron sources 110 and gamma detectors 120. FIG. 6 shows a front view of object 125 on transport mechanism inside source/detection array 100 and moving towards turn 132. Referring to FIG. 4, in a preferred method of the present invention, an object is transported on transport mechanism 140 into cavity 145 in enclosure 150. Cavity 145 comprises one or more turns (e.g., turns 132 and 134) which preclude a straight line trajectory through cavity 145. As stated previously, precluding a straight line trajectory through cavity 145 prevents stray radiation from escaping from the open ends of cavity 145. Therefore, the size of the enclosure or facility needed to shield the user and/or the general public from gamma radiation is significantly reduced. The object is transported inside of detector array 100 which is disposed around transport mechanism 140 inside of cavity 110. FIG. 5 depicts object 125 inside of detector array 100. Neutron sources 110 generate low intensity neutron particles and irradiate object 125. The neutron particles interact with nuclei in target substances in the object and generate prompt gamma ray photons resulting from inelastic scattering of neutrons from the target substances. The gamma ray photons are detected by gamma detectors 120. A gamma count ratio for substances in object 125 may be determined and compared to known gamma count ratios for elements (e.g., carbon, nitrogen, oxygen, and chlorine) in explosives or other controlled substances. For example, the ratio of nitrogen to oxygen in the object following irradiation with neutrons may be characteristic of the ratio of nitrogen to oxygen in an explosive which is bombarded with neutrons. Thus, components of explosives have characteristic or signature ratio of various elements (e.g., nitrogen/oxygen, carbon/nitrogen, carbon/oxygen, nitrogen/carbon). If the gamma ray ratio emitted by an object is the same or similar to the signature gamma ray ratio for an explosive or other controlled substance, the object is likely to contain the same substance. Alternatively, selected peaks of the energy spectrum of the detected prompt gamma rays are analyzed to provide a fingerprint or signature characteristic of the material in the object responsible for scattering the neutrons. The energy spectrum of the detected gamma rays can be compared to known energy spectra for elements (e.g., carbon, nitrogen, oxygen, and chlorine) in explosives and other controlled substances. The energy spectrum emitted from nuclei bombarded with neutrons produces characteristic gamma radiation patterns. Particular spectral lines are associated with, for example, carbon, nitrogen, oxygen, and chlorine. See, e.g., U.S. Pat. No. 5,098,640, hereby incorporated by reference in its entirety. The use of a plurality of low intensity neutron sources, rather than one high intensity neutron source, not only reduces size of the shield, but also advantageously produces a lower amount of stray radiation, further improving the safety characteristics of preferred embodiments of the invention. In one embodiment, the neutron sources are pulsed simultaneously. The pulsing of a plurality (e.g., ten or more) of neutron sources at a lower energy density (107 vs 1011) will provide neutrons for detection purposes (e.g., sufficient ringing resonance to detect controlled substances) while reducing the overall risks associated with using a radioactive source. Pulsing the plurality of neutron sources simultaneously provides sufficient power to detect the presence of an explosive or other controlled substance. Once a particular explosive or other controlled substance is detected, the plurality neutron sources can then be flashed, i.e., sequentially activated, to provide a lower power scan which however generates spatial information so the operator can more accurately identify the location of the explosive or controlled substance in the object of interest. While the invention has been described in detail in connection with the exemplary embodiment, it should be understood that the invention is not limited to the above disclosed embodiment. Rather, the invention can be modified to incorporate any number of variations, alternations, substitutions, or equivalent arrangements not heretofore described, but which are commensurate with the spirit and scope of the invention. Accordingly, the invention is not limited by the foregoing description or drawings, but is only limited by the scope of the appended claims.
062018469	description	In FIG. 1, the cylindrical body 10 of uranium fits into the cup-shaped jacket 12 of aluminum. The jacket 12 may have, for example, a length of 9.5 inches, an outside diameter of 1.44 inches and an inside diameter of 1.368 inches, thus having a wall thickness of approximately 0.036 inches. It is necessary that the wall thickness of the jacket 12 be kept relatively small because of the fact that the neutron absorption of the jacket, and thus the loss of neutrons to the chain reaction of the neutronic reactor, increases with the quantity of aluminum present. The fissionable body 10 of uranium or other fissionable material may be, for example, 8 inches in length and, for example, 1.360 inches in diameter. It will be seen that the clearance between the fissionable body 10 and the jacket 12 in the example given is extremely small, being approximately 0.004 inches. The jacket 12 is preferably chamfered at the mouth as at 14, and the fissionable member 10 is preferably bevelled slightly at the edges as at 16 to allow convenient fitting of the fissionable member 10 into the jacket 12. The end cap 18 has approximately the same diameter as the fissionable member 10, and preferably has an axial boss 20 on the outer surface thereof for convenient handling. The elements described above and illustrated in the drawing do not in themselves constitute any portion of the present invention. In FIG. 1 there is illustrated, in addition to the elements already described, a cup-shaped sleeve 22 preferably of steel, into which the jacket 12 is designed to fit tightly. The sleeve 22 is likewise chamfered at its mouth as at 24 to facilitate insertion of the jacket 12 therein. In the method of the present invention, the jacket 12 is first inserted into the sleeve 22, the bottom of jacket 12 then resting at the bottom of the sleeve 22. Preferably, a ball, for example of steel (not illustrated in the drawing) is then inserted into the mouth of the jacket 12. The ball is of larger diameter than the mouth of the jacket 12, and thus pressure applied to the ball has the effect of slightly flanging the mouth of the jacket 12 against the inner surface of the sleeve 22, and thus effectively sealing any clearance which may exist between the jacket 12 and the sleeve 22 from the molten bonding agent into which this assembly is subsequently plunged. The assembly consisting of the jacket 12, slightly flanged at the mouth as described above, and the sleeve 22 is then plunged into a molten bath of an aluminum-silicon alloy. Preferably, the bath has a silicon content of 11.2 to 11.5 percent and has a temperature of 590 to 596.degree. C. The materials, composition, and properties of the bath are not in themselves the invention of the present inventor. It will be understood further that the teachings of the present invention are not limited to the particular bonding agent employed, nor, indeed, to any particular configuration or materials for the jacketed slug. Upon being plunged into the bath of bonding material, the jacket 12 fills with the bonding material, but its external surface does not come into contact with the bonding material because of the flanging above described. Thereupon, the uranium body 10, previously cleaned and prepared for bonding by methods which constitute no part of the present invention, is inserted into the jacket 12. Preferably, this operation is performed while the elements are completely submerged in the bath of molten bonding material. However, for convenience, the mouth of the jacket 12 encased in the sleeve 22 may be brought slightly above the surface of the bath in order that the end of the uranium body 10 may be successfully inserted into the mouth of the jacket 12, the material of the bonding bath being opaque. However, it is important that only a small portion of the uranium be inserted into the jacket before the whole assembly is plunged back into the bath, wherein the uranium body 10 is allowed to drop to the bottom of the jacket 12. Furthermore, if the mouth of the jacket 12 is so slightly withdrawn momentarily, the uranium body 10 should itself first be dipped into the bonding bath. As stated above, the preferable manner of performing the operation involves the insertion of the end of the body 10 into the jacket 12 while both elements are completely submerged. The procedure involving slight withdrawal of the mouth of the jacket 12 from the bath for purposes of visibility in commencing the insertion of the uranium body 10 has been found, however, to be permissible provided that it is done with great speed, and provided further that the uranium has first been wetted by the bonding agent by immediately previous dipping as above described. The uranium should be removed from the bath for such commencement of insertion simultaneously with the exposure of the mouth of the jacket. Upon completion of the insertion of the uranium body 10 into the jacket 12, while the assembly remains submerged, the cap 18 is inserted, being held as by tongs gripping the boss 20. The assembled piece is then removed from the bath and transferred with tongs to a bath of cold flowing water in a quench tank until the molten bonding agent is solidified. The slug may then be removed from the sleeve 22 by merely inverting the latter, since the interfaces between the jacket 12 and the sleeve 22 have not been exposed to the bonding bath. Thereupon, the end cap 18 may be machined down in a lathe to remove the boss 20 and any bonding agent which has adhered to the outer face thereof. It will be seen that the sleeve 22 serves the four-fold function of holding the jacket in position for slug insertion, preventing the bonding agent from fouling the outside of the jacket, supporting the thin-walled jacket to prevent its deformation, and transferring heat from the molten bath to the jacket. The method above described produces a unitary slug structure as illustrated in FIG. 2 of the drawing. It has been found that this method produces a bonding layer, designated by the numeral 26, which is completely free of faults caused by air gaps or impurities. The closure of the jacket may be further assured by the weld between the cap 18 and the jacket 12 at 30. The teachings of this invention shall not be deemed to be limited to the exact embodiment illustrated in the drawing and described above. Equivalent methods utilizing the teachings of this invention will be devised by persons skilled in the art.
	summary
055704026	claims	1. A support for a control rod drive housing in a boiling water reactor, comprising: first means for supporting a control rod drive in the case of a housing failure; second means for supporting said control rod drive in the case of a housing failure and shielding persons working under the reactor vessel from radiation, said second means being supported by said first supporting means, and wherein said second means can be raised and lowered between a non-support position where said control rod drive is not supported and a support position where said control rod drive is supported; and a radiation shield means separate from said second support means for shielding persons working under the reactor vessel from radiation, said radiation shield means being provided about the control rod drive above said second supporting means, said radiation shield means being removable from the second support means and supported by the second support means when the second support means is in its support position. means for supporting a control rod drive in the case of a housing failure; and means for shielding persons working under the reactor vessel from radiation, said shielding means comprising a cylindrical-shaped shield ring disposed about said control rod drive, said shield ring being supported by said supporting means. 2. A control rod drive housing support according to claim 1, wherein said first supporting means comprises a plurality of support members provided in rows on opposing sides of a lower portion of a plurality of control rod drives. 3. A control rod drive housing support according to claim 2, wherein said second supporting means comprises a plurality of support cups, each of said support cups receiving and shielding a lower portion of a respective control rod drive and supporting the respective control rod drive in the case of a housing failure. 4. A control rod drive housing support according to claim 3, wherein each of said support cups have an annular rim for supporting a respective control rod drive. 5. A control rod drive housing support according to claim 4, wherein said support cups each have first and second support tabs protruding from opposing sides thereof, said first and second support tabs being supported by said first supporting means to transfer the load of said support cups to said first supporting means. 6. A control rod drive housing support according to claim 5, wherein said support cups each have an upper support surface for supporting said radiation shield means. 7. A control rod drive housing support according to claim 3, wherein said radiation shield means comprises a plurality of shield rings, each of said shield rings being disposed about a respective control rod drive and shielding a portion of the respective control rod drive above the portion thereof shielded by a respective one of said support cups. 8. A control rod drive housing support according to claim 7, wherein said shield rings are supported by said support cups. 9. A control rod drive housing support according to claim 1, wherein said radiation shield means comprises a plurality of shield rings separate from said second supporting means, each of said shield rings being disposed about a housing of a respective control rod drive. 10. A control rod drive housing support according to claim 9, wherein said shield rings are generally cylindrical-shaped. 11. A control rod drive housing support according to claim 10, further comprising a hanger means provided on each of said shield rings for supporting the shield rings on a respective control rod drive when said second supporting means is lowered. 12. A control rod drive housing support according to claim 11, wherein said hanger means comprises a pin fixed to and extending radially inward from a side wall of each shield ring for engaging an upper surface of a respective control rod drive. 13. A control rod drive housing support according to claim 11, wherein vertical slots are formed in side walls of each of said shield rings, and further comprising seismic restraints disposed between adjacent control rod drives, said seismic restraints being received in said vertical slots. 14. A control rod drive housing support according to claim 13, wherein said vertical slots are open at an upper end thereof. 15. A control rod drive housing support according to claim 14, wherein said vertical slots comprise four vertical slots circumferentially-spaced about each of said shield rings. 16. A support for a control rod drive housing in a boiling water reactor comprising: 17. A control rod drive housing support according to claim 16, wherein said shield ring comprises a plurality of vertical slots, and further comprising seismic restraints disposed between adjacent control rod drives, said seismic restraints being received in said vertical slots. 18. A control rod drive housing support according to claim 17, wherein said vertical slots comprise four vertical slots circumferentially spaced about said shield ring. 19. A control rod drive housing support according to claim 17, wherein said shield ring has an inner diameter which is slightly larger than an outer diameter of a control rod drive housing so that said shield ring can be raised and lowered relative to the control rod drive housing. 20. A control rod drive housing support according to claim 19, further comprising a hanger means provided on said shield ring for supporting the shield ring on said control rod drive when said supporting means is removed. 21. A control rod drive housing support according to claim 20, wherein said hanger means comprises a pin fixed to and extending radially inward from a side wall of said shield ring for engaging an upper surface of the control rod drive housing. 22. A control rod drive housing support according to claim 16, wherein said supporting means comprises a support cup for receiving and shielding a lower portion of the control rod drive and for supporting the control rod drive in the case of a housing failure. 23. A control rod drive housing support according to claim 22, wherein said support cup has an upper support surface engaging and supporting said shield ring.
043748015	abstract	A nuclear reactor containing assemblies having two caps and a framework inside which fuel rods are disposed vertically is reloaded by first inverting the assembly to be removed into a vertical position permitting the then lower cap to be demounted. The spent fuel rods are then identified and removed in sets and deposited in a suitable storage area, and replacement rods are placed in sets in the framework of the fuel assembly. Finally, the lower cap is replaced and the assembly is conveyed into the reactor vessel at its new location.
047160128	description	DESCRIPTION OF THE PREFERRED EMBODIMENT FIG. 1 illustrates a typical pressurized water reactor (PWR) to which the invention has been applied. The reactor 1 includes an upright cylindrical pressure vessel 3 having an integral hemispherical lower head section 5 and a removable upper hemispherical head 7. The removable head 7 is secured to the pressure vessel by a plurality of bolts 9 which extend axially through an annular flange 11 on the head and into a confronting flange 13 on the upper end of the cylindrical pressure vessel 3. The cylindrical pressure vessel 3 also defines an annular, radially inwardly projecting support flange 15 near its upper end on which are supported the reactor internals identified generally by the reference character 17. The internals include a cylindrical core barrel 19 which is suspended inside the pressure vessel 3 by a radially, outwardly extending flange 21 around its upper edge which seats on the support flange 15. The core barrel 19 terminates at its lower end in a thick lower core support 23. Several angularly spaced sliding connections 25 between the lower core support 23 and the adjacent pressure vessel provide lateral support for the lower end of the core barrel while permitting free axial displacement due to differences in the coefficients of thermal expansion of the lower alloy pressure vessel 3 and the stainless steel core barrel 19. The reactor core 27 is mounted in the core barrel 19 between a lower core plate 29 and an upper core plate 31. The lower core plate 29 is secured to the side walls of the core barrel 19 while the upper core plate 31 is suspended by columns 33 from an upper core support 35. This upper core support 35, which is also referred to as the "top hat", is in turn suspended by a radially outwardly extending flange 37 from the support flange 15 on the inside of the pressure vessel 3 which, as discussed, also supports the core barrel 19. The flanges 37 and 21, on the upper core support and the core barrel respectively, are clamped down onto the flange 13 by the flange 11 on the hemispherical upper head as it is bolted onto the pressure vessel 3. An annular spring 39 between the flanges 37 and 21 takes up any slack created by manufacturing tolerances. Control rod assemblies 41 extending downward through the removable upper head 7, the upper core support 35, and the upper core plate 31, include control rods which are inserted into and withdrawn from the reactor to control reactivity. A secondary core support 43 is mounted under the core barrel 19. It includes a secondary core support base plate 45 suspended from the lower core support 23 by four columns 47 each supporting an axially crushable energy absorber 49. The energy absorbers 49 reduce the impact forces, and thereby preserve pressure vessel integrity, in the unlikely event that the core barrel suspension system should fail causing the core barrel 19 and reactor core 27 to fall onto the lower hemispherical head section 5. Conduits 51, known as instrumentation thimbles, extend upward through the lower hemispherical head section 5 of the pressure vessel, through the secondary core support base plate 45, the energy absorbers 49, the columns 47, and the lower core support 23 and into the reactor core 27. Additional instrumentation thimbles 51 extend upward through additional columns 47 suspended from the bottom of the lower core support and into the reactor core at selected locations across the core. Apertured horizontal plates 53 and 55 provide a rigid framework for supporting the columns 47. The enlarged views of FIGS. 2, 3 and 4 illustrate more clearly the details of the base plate 45 of the secondary core support 43. This base plate 45 is a square stainless steel plate with truncated corners and a large square aperture 57 through the center. It is mounted on the lower ends of the energy absorbers 49 which seat in the four counter bored apertures 58. The lower and side surfaces of the base plate 45 transition into a spherical surface 59 which conforms to the confronting surface 61 on the inner wall of the hemispherical lower head section 5 of the pressure vessel 3. The core barrel is suspended so that a radial gap 63 is defined by these spherical surfaces. The gap 63 is about 1.06 inches when the reactor is cold and narrows to about 0.5 inches when hot due to the differences in the coefficients of thermal expansion between the low alloy pressure vessel 3 and the stainless steel internals 17. In operation of the reactor 1, reactor coolant, in the form of light water, enters the pressure vessel 3 through inlet nozzle 65 and flows downward through the annular space 67 called the downcomer between the pressure vessel inner wall 69 and the core barrel 19. From the downcomer, it passes into the lower hemispherical head 5 where it reverses direction and flows upward through passages 71 in the lower core support 23 and through the reactor core 27, before passing out through outlet nozzle 73. The reactor coolant discharged through the outlet nozzle circulates in an external loop (not shown) where the heat absorbed from fission reactions in the reactor core is used to produce steam for generating electricity. While only one inlet nozzle 65 and outlet nozzle 73 each are shown in FIG. 1, a typical PWR has two to four primary loops each having similar inlet and outlet nozzles. Periodically, the reactor is shutdown for refueling. The head bolts 9 are removed and the hemispherical removable head 7 is lifted off. The top hat 35 is then lifted out bringing the upper core plate with it, thereby exposing the fuel assemblies in the core 27 which can be replaced and rearranged as desired. It is during this refueling process that debris could fall down into the lower head section. As also mentioned above, it is also possible for debris generated during failure of other hardware or repair work in the primary loops to be carried by reactor coolant into the reactor vessel where it too falls down into the lower hemispherical head section 5. It is possible for some of this debris to become lodged in the gap 63 between the secondary core support base plate 45 and the lower hemispherical head 5 when the reactor is cold and the gap is at its maximum width. As the reactor heats up, the debris prevents the gap from narrowing and instead the differential thermal expansion of the pressure vessel and the internals causes the core barrel to unseat from the support flange 15 against the annular spring 39. This is an undesirable condition which could lead to unacceptable vibration of the core barrel and the connected components as a result of the turbulent reactor coolant flow. In order to prevent this condition from occurring, a strainer member 75 is placed in the gap 63. The strainer member 75 is an annular stainless steel member having a central aperture defined by a thickened rim 77 which conforms to the peripheral configuration of the secondary core support base plate 45. The strainer member is mounted on the base plate 45 by welding the rim 77 to the side faces 81. Four inverted L-shaped stop members 83 welded to the center of each side of the rim 77 engage the top surface of the base plate 45 to set the vertical location of the strainer member. The main body of the strainer 75 is a planar section 85 which extends radially outward from the rim to the inner surface 61 of the lower hemispherical head 5 and is provided with apertures 87 through which reactor coolant, but not debris, may pass. Preferably, the planar section 85 is upwardly convex which increases the angle .theta. at which it contacts the surface 61. The strainer contacts the surface 61 when the reactor is in the cold state and it is resilient such that as the components heat up and the internals grow with respect to the pressure vessel so that the base plate 45 moves downward, the strainer flexes downward and maintains edge contact with the wall 61. Circular cutouts 89 in the strainer 75 some extending partially into the base plate 45, accommodate the instrumentation thimbles 51 which extend upward into the reactor core 7. To facilitate installation of the strainer member 75, it is preferably fabricated in two halves and welded in place along the diagonal 91. While specific embodiments of the invention have been described in detail, it will be appreciated by those skilled in the art that various modifications and alternatives to those details could be developed in light of the overall teachings of the disclosure. Accordingly, the particular arrangements disclosed are meant to be illustrative only and not limiting as to the scope of the invention which is to be given the full breadth of the appended claims and any and all equivalents thereof.
	claims	1. An electrostatic fusion device, comprising:a vacuum chamber;a potential well disposed in said vacuum chamber;a partial vacuum environment in said vacuum chamber containing fusion reaction ions;plural concentrically arranged collector cages surrounding said potential well, said collector cages being generally spherical in shape and of progressively increasing size from an innermost collector cage to an outermost collector cage and having progressively increasing voltage levels from said innermost collector cage to said outermost collector cage, said voltage levels corresponding to energy levels of protons emitted from a fusion reaction occurring within said potential well, and being connected to deliver an electrical current output at said collector cage voltage levels;plural electron source pathways respectively connected to said collector cages to deliver said electrical current output;a diverter wire associated with each of said collector cages;each diverter wire being of higher potential than its associated collector cage and being adjacent thereto at a location that is between said associated collector cage and said potential well;plural voltage sources respectively connected to said diverter wires, said voltage sources being of progressively increasing voltage from an innermost diverter wire to an outermost diverter wire and each being of higher potential than said collector cage associated with said diverter wire to which said voltage source is connected. 2. A fusion device according to claim 1 wherein said collector cages are series-connected to ground potential through corresponding charge storage devices. 3. A fusion device according to claim 2 wherein there are ten collector cages respectively maintained at voltage levels ranging from 1 MV to 10 MV in 1 MV increments, with a 1 MV collector cage being the innermost one of said collector cages and a 10 MV collector cage being the outermost one of said collector cages. 4. A fusion device according to claim 3 wherein each collector cage is series-connected to ground potential through a bank of “n” 1 MV-rated capacitors, where “n” is the MV potential of the collector cage. 5. A fusion device according to claim 4 wherein each said bank of capacitors associated with one of said collector cages is parallel-connected to banks of said capacitors associated with other of said collector cages. 6. A fusion device according to claim 1 wherein each diverter wire adjacent to a collector cage is connected to an adjacent collector cage of higher potential than the collector cage to which said diverter wire is adjacently located. 7. A fusion device according to claim 6 wherein there are several collector cages of different voltage, with the outermost collector cage having the highest voltage and being connected to a voltage potential source for fixing the voltage of said outermost collector cage. 8. A fusion device according to claim 7 wherein collector cage voltage levels are adjusted by controlling electrical current through said collector cages. 9. A fusion device according to claim 8 wherein said reaction ions are 3He ions and said fusion reaction is a 3He-3He reaction.
	claims	1. A device for system diagnosis of an aircraft comprising equipment, said device comprising:means for the monitoring of the equipment and means for emitting messages of observations (O1, O2, . . . , On) on a basis of effects produced by equipment of the system;means for the determination of a set of observations (Eobs) on a basis of the messages of observations (O1, O2, . . . , On) arising from the means for the monitoring, of a log of the messages of observations (O1, O2, . . . , On) and of a model representing a current state of the system, said means for the determination of a set of observations (Eobs) implementing temporal logic;means for the determination of indictments on a basis of observations (O′1, O′2, . . . , O′k) of the set of observations (Eobs) and of a behavioural model of the system, the indictments being logical relations between operating modes of equipment having produced effects;means for the determination of maintenance operations on the basis of the indictments; anda model explorer for computing logical relations (LR1, LR2, . . . , LRk) associated with the observations (O′1, O′2, . . . O′k) on a basis of the behavioural model of the system, the means for the determination of the indictments comprising means for the con'unction of the logical relations (LR1, LR2, . . . , LRk),wherein the con'unction of the logical relations (LR1, LR2, . . . , LRk) is carried out by means of a binary decision diagram and comprises:determination of a first binary decision diagram (BBD1, BBD2, . . . , BBDk) for each observation (O′1, O′2, . . . , O′k) of the set of observations (Eobs);computation of a second binary decision diagram (BDD′) by a conjunction operation on the first binary decision diagrams (BBD1, BBD2, . . . , BBDk); andcomputation of an indictment binary decision diagram by the development of the second binary decision diagram (BDD′). 2. The device for system diagnosis according to claim 1, wherein the means for the determination of the indictments comprises a model explorer allowing the direct exploration of the behavioural model of the system and the extraction of events in the form of sequences. 3. The device for system diagnosis according to claim 2, wherein the means for the determination of indictments is on board the aircraft, the determination of the indictments being performed during flight. 4. The device for system diagnosis according to claim 2, wherein the means for the determination of indictments is situated on the ground, the determination of the indictments being performed during the flight. 5. The device for system diagnosis according to claim 1, wherein a binary decision diagram comprises nodes, from which two branches sprout, and terminal leaves connected together by branches, each node representing a Boolean variable indicating the presence or the absence of an event, a branch being termed a terminal if it ends at a terminal leaf of value 0 or 1, or being termed intermediate if it ends at another node, a first branch representing the case where the event represented is absent, and a second branch representing the case where the event represented is present, the diagram comprising a single root node from which two branches, to which nodes or leaves are connected in cascade, sprout. 6. The device for system diagnosis according to claim 5, wherein the development of the second binary decision diagram (BDD′) is carried out by performing traversals of the second binary decision diagram (BDD′) from the root up to each of the terminal leaves of value 1, a traversal corresponding to a combination of events having given rise to the set of observations (Eobs). 7. The device for system diagnosis according to claim 1, wherein the determination of the logical relations (LR1, LR2, . . . , LRk) is carried out on the basis of a storage unit for associating a logical relation with each observation. 8. The device for system diagnosis according to claim 1, wherein the behavioural model comprises a component for each item of equipment of the system, a component of the behavioural model comprising at least:an input stream, an output stream, a state indicating the availability of the item of equipment and comprising:a fault mode, for indicating whether the component is defective or healthy and an operating mode for indicating whether the component is turned on, turned off or unavailable, and a logical relation between the input stream, the output stream and the state, and the output stream. 9. The device for system diagnosis according to claim 1, wherein the system under diagnosis comprises a plurality of processing units and is decomposed into various hierarchical levels grouping together these various processing units, a first hierarchical level being lower than a second level when the first level comprises component processing units of the processing units of the second level. 10. The device for system diagnosis according to claim 9, wherein the computation means of the device is implemented on two different hierarchical levels of the system comprising:a first level implementing means for the monitoring and means for the determination of a set of observations; anda second level, higher than the first level, implementing means for the determination of a set of observations on the basis of the sets of observations arising from the first level, means for the determination of indictments and means for the determination of the maintenance actions. 11. The device for system diagnosis according to claim 9, wherein the computation means of the device according to the invention is implemented on three different hierarchical levels of the system comprising:a first hierarchical level implementing means for the monitoring;a second level, higher than the first level, implementing means for the determination of a set of observations on the basis of the messages arising from the means for the monitoring of the first level and means for the determination of indictments; anda third level, higher than the second level, implementing means for the determination of a set of observations on the basis of the indictments arising from the second level, means for the determination of indictments and means for the determination of the maintenance actions. 12. The device for system diagnosis according to claim 11, wherein the first level groups together processing units of Shop Replaceable Unit type corresponding to electronic modules internal to an item of equipment and in that the second level corresponds to a maintenance application internal to the item of equipment.
046577211	summary
	abstract	A system for creating an image of the internal features of an object includes an X-ray source and detector array positioned to interpose the object between the X-ray source and the detector array. An X-ray beam is passed through the object along a first path. While passing through the object, the beam is successively filtered four times, each time with a different filter. The successive filtration of the beam results in the production of four electrical signals by the detector which are processed to create an image signal for the path. The process is repeated for a plurality of paths through the object and the resulting image signals are combined using traditional computer tomography techniques to produce an image of the object.
	claims	1. A collimator comprising: a panel defining an exit port through which an X-ray beam from an X-ray source emanates towards a detector; and one or more blades held between the panel and the X-ray source and covering at least a portion of the exit port, each of the one or more blades having a primary blocking member and a secondary blocking member with material densities sufficient to block X-ray radiation, the primary blocking member configured to shape the X-ray beam that is emanated through the exit port, the secondary blocking member secured in a fixed position relative to the primary blocking member between the primary blocking member and the panel to block scatter radiation from the X-ray beam from emanating through the exit port,wherein the primary blocking member has a substantially non-planar surface facing toward the X-ray source and the secondary blocking member has a substantially planar surface facing toward the X-ray source, andwherein each of the one or more blades includes a base that engages and holds both the primary blocking member and the secondary blocking member, the base having a material density that is less than the material densities of the primary blocking member and the secondary blocking member. 2. The collimator of claim 1, the base has a first side and a second side opposite the first side, the primary blocking member mounted along the first sine of the base, the respective secondary blocking member mounted along the second side of the base. 3. The collimator of claim 1, wherein the one or more blades includes a first blade and a second blade spaced apart from each other to define an aperture, the aperture having a depth that includes a first passage and a second passage, the first passage defined between respective inner edges of the primary blocking members of the first and second blades, the second passage defined between respective inner edges of the secondary blocking members of the first and second blades, the second passage disposed between the first passage and the exit port. 4. The collimator of claim 1, wherein at least one of the one or more blades is movable relative to the panel to adjust the portion of the exit port covered by the one or more blades for shaping the X-ray beam that is emanated through the exit port. 5. The collimator of claim 1, wherein the substantially non-planar surface of the primary blocking member of each of the one or more blades has a concave profile relative to the secondary blocking member of the respective blade such that a middle of the concave profile is closer to the secondary blocking member than each of a first end and a second end of the concave profile. 6. The collimator of claim 5, wherein the concave profile is a continuous curve extending from the first end of the concave profile to the second end. 7. The collimator of claim 5, wherein the concave profile is discontinuous and includes a plurality of segments disposed adjacent to one another between the first and second ends of the concave profile, each of the segments having a different angular orientation than an adjacent one of the segments. 8. The collimator of claim 1, wherein the substantially non-planar surface of the primary blocking member has a sawtooth pattern that includes a plurality of segments disposed adjacent to one another, the segments defining alternating peaks and valleys along a length of the substantially non-planar surface. 9. The collimator of claim 1, wherein the substantially non-planar surface of the primary blocking member has both linear segments and curved segments along a length of the substantially non-planar surface. 10. The collimator of claim 1, further comprising a housing that defines a cavity, the panel that defines the exit port disposed along a wall of the housing, the panel having a material density sufficient to block X-ray radiation, the one or more blades disposed within the cavity of the housing. 11. The collimator of claim 1, wherein the secondary blocking member is a planar plate that is stacked between the primary blocking member and the panel, and the substantially planar surface of the secondary blocking member is disposed underneath the substantially non-planar surface of the primary blocking member relative to the X-ray source. 12. The collimator of claim 1, wherein the primary blocking member includes one or more of lead, tungsten, or tungsten polymer. 13. A collimator comprising:a panel defining an exit port configured to allow an X-ray beam from an X-ray source to emanate therethrough towards a detector; andone or more blades held between the panel and the X-ray source and covering at least a portion of the exit port, each of the one or more blades comprising:a base having a first side facing the X-ray source;a primary blocking member mounted to the first side of the base, the primary blocking member having a material density sufficient to block X-ray radiation, the primary blocking member having a substantially non-planar surface facing toward the X-ray source that is configured to shape the X-ray beam that is emanated through the exit port; anda secondary blocking member mounted to the base in a fixed position relative to the primary blocking member and disposed between the primary blocking member and the panel, the secondary blocking member having a material density sufficient to block X-ray radiation, the secondary blocking member including a substantially planar surface facing toward the X-ray source that is configured to block scatter X-ray radiation from emanating through the exit port,wherein the substantially non-planar surface of the primary blocking member has a concave profile relative to the secondary blocking member such that a middle of the concave profile is closer to the secondary blocking member than each of a first end and a second end of the concave profile. 14. The collimator of claim 13, wherein the concave profile is a continuous curve extending from the first end of the concave profile to the second end. 15. The collimator of claim 13, wherein the concave profile is discontinuous and includes a plurality of segments disposed adjacent to one another between the first and second ends of the concave profile, each of the segments having a different angular orientation than an adjacent one of the segments. 16. The collimator of claim 13, wherein the substantially non-planar surface of the primary blocking member has both linear segments and curved segments along a length of the concave profile. 17. The collimator of claim 13, wherein the secondary blocking member is a planar plate that is mounted to a second side of the base that is opposite the first side. 18. The collimator of claim 13, wherein the one or more blades includes a first blade and a second blade spaced apart from each other to define an aperture, the aperture having a depth that includes a first passage and a second passage, the first passage defined between respective inner edges of the primary blocking members of the first and second blades, the second passage defined between respective inner edges of the secondary blocking members of the first and second blades, the second passage disposed between the first passage and the exit port. 19. A collimator comprising:a first blade and a second blade spaced apart from each other to define an aperture therebetween, the first and second blades configured to receive an X-ray beam from an X-ray source and to shape the X-ray beam via the aperture, at least one of the first and second blades being moveable relative to other blade to adjust a size of the aperture, each of the first and second blades comprising:a base having a first side facing the X-ray source and a second side opposite the first side;a primary blocking member mounted to the first side of the base; anda secondary blocking member mounted to the second side of the base and secured in a fixed position relative to the primary blocking member,wherein the primary blocking member and the secondary blocking member have material densities sufficient to block X-ray radiation, the primary blocking member having a substantially non-planar surface facing toward the X-ray source that shapes the X-ray beam, the secondary blocking member having a substantially planar surface facing toward the X-ray source that blocks scatter X-ray radiation,wherein the aperture has a depth that includes a first passage and a second passage, the first passage defined between respective inner edges of the primary blocking members of the first and second blades, the second passage defined between respective inner edges of the secondary blocking members of the first and second blades. 20. The collimator of claim 19, wherein the substantially non-planar surface of the primary blocking member of each of the one or more blades has a concave profile relative to the secondary blocking member of the respective blade such that a middle of the concave profile is closer to the secondary blocking member than each of a first end and a second end of the concave profile, and wherein the secondary blocking member is a planar plate.
	abstract	A system and method is provided for monitoring, gathering and aggregating performance metrics of a plurality of members configured as an entity. Configurable performance metric settings can be set at a first computer (e.g., a first member) and dynamically propagated to all members of the entity to establish performance metric configuration settings at each of the plurality of members. In one aspect of the invention, a system and method log performance metric data periodically at a predefined time period and resolution at a plurality of members for one or more performance metrics. The performance metric data values logged at the predefined time period and resolution are aggregated to data sets of at least one larger time period and resolution. Valid performance data values of similar time periods and resolutions are then gathered from the plurality of members and aggregated over the entity to provide a unified result set for the entity.
048266477	abstract	A mechanical spectral shift reactor comprises apparatus for inserting and withdrawing water displacer elements having differing neutron absorbing capabilities for selectively changing the water-moderator volume in the core thereby changing the reactivity of the core. The displacer elements may comprise substantially hollow cylindrical low neutron absorbing rods and substantially hollow cylindrical thick walled stainless rods. Since the stainless steel displacer rod have greater neutron absorbing capability, they can effect greater reactivity change per rod. However, by arranging fewer stainless steel displacer rods in a cluster, the reactivity worth of the stainless steel displacer rod cluster can be less than a low neutron absorbing displacer rod cluster.
	description	This application is a Continuation of U.S. patent application Ser. No. 12/434,131 filed May 1, 2009 and issued as U.S. Pat. No. 8,323,172 on Dec. 4, 2012 which claims priority to U.S. Provisional Application Ser. Nos. 61/049,843 filed May 2, 2008; 61/096,459 filed Sep. 12, 2008; 61/150,081 filed Feb. 5, 2009; and 61/166,369 filed Apr. 3, 2009, the disclosures of which are hereby incorporated by reference in their entireties. The present invention relates to low-dose rate (LDR) brachytherapy radiation treatment methods, systems and computer program products. Roughly 230,000 new cases of prostate cancer are expected in the U.S. this year. Typically 80-90% of these cases are relatively early stage disease for which various treatment options are available. Primary treatment options involving radiation include external beam radiation therapy, which uses high-energy x-ray beams that intersect the prostate from multiple angles, and brachytherapy, in which a radioactive source is introduced into the prostate itself. Typical brachytherapy techniques use so-called “seeds,” which are small (approximately 0.8×4.5 mm) cylinders that contain a radioactive element in a stainless-steel casing. A number of seeds, usually ranging from 80-120 seeds, are placed into the prostate using small gauge needles. The seeds can remain in place permanently while the emitted radiation decays over time. The common radioisotopes used in the seeds are iodine-125, palladium-103 and cesium-131. The goal of the radiation oncologist is to ensure that the total dose received by the cancer cells is sufficient to kill them. Seeds can be placed during an outpatient procedure in a single day and thus present an attractive treatment option for patients versus the many weeks required for external beam radiation therapy. Good candidates for brachytherapy seed therapy are typically patients having a PSA value ≦10, a Gleason score of ≦7 and low-stage disease (T1c or T2a); however, patients with more advanced stage disease may also be treated with brachytherapy. In some cases, patients (e.g., with more advanced disease) may be candidates for brachytherapy plus external beam therapy. The use of seeds has grown rapidly, and long-term survival data for LDR brachytherapy treatment of the prostate is typically good. In treating prostate cancer with brachytherapy seeds, it may be desirable to create a uniform radiation pattern within the prostate gland or within a region of the prostate gland. Computer code or treatment planning software can be used to select the number of seeds and their relative placement so that the desired radiation dose is achieved. This is a relative complex procedure because each individual seed is essentially a “point source” of radiation, and thus the radiation contributions from all of the seeds must be summed to achieve the total radiation dose. When the seeds are placed, great care is typically taken to ensure that they are arranged in the pattern specified by the treatment planning software. However, some deviation in seed placement may occur due to the divergence of needles as they are inserted. See Nath et al., Med Phys 27, 1058 (2000). A more problematic occurrence is the tendency of seeds to migrate once they exit the insertion needle. See Meigooni et al., Med Phys 31, 3095 (2004). It is not uncommon for seeds to migrate. In some cases, seeds may be caught in an efferent vessel and become embolized in the lung or excreted with urine. Gross movement of the seeds can create non-uniformities in the radiation pattern and thus potentially compromise the efficacy of therapy. In an attempt to mitigate the post insertion migration of brachytherapy seeds, various products have been developed. For example, the RapidStrand™ device from Oncura (Arlington Heights, Ill., USA) is a hollow suture material that contains conventional seeds in a “linked sausage” arrangement. The suture material subsequently dissolves away leaving the seeds implanted in the patient. However, the seeds are held by the suture for a time that allows for healing and better retention of the seeds. Various so-called “sleeves for seeds” are also available. Another device that is commercially available from IBA (Louvain-la-Neuve, Belgium) under the trade name Radiocoil™ is a coiled structure device that contains rhodium metal that is proton-activated to produce Pd-103. Accordingly, the radioactivity is emitted along the entire length of the device. Notably, the ability of the radiation oncologist to achieve the highest accuracy in therapy planning is hampered by the discrete nature of the current “seed” radiation sources due to their limited size and anisotropic radiation patterns. The tendency of seeds to move when placed in or near prostatic tissue is a problem that, while not invalidating this excellent form of therapy, creates a non-ideal situation for planning (e.g., requiring revalidation of the placement by CT scan). For example, migration of seeds to the lungs can result in incidental lung doses that are not favorable. According to some embodiments of the invention, methods of forming a low-dose-rate (LDR) brachytherapy device include depositing a solution comprising a soluble form of a radioactive material on a substrate. The soluble form of the radioactive material is converted to a water-insoluble form of the radioactive material on the substrate. A medical device is formed from the substrate and the water-insoluble form of the radioactive material. According to some embodiments of the invention, a low-dose-rate (LDR) brachytherapy device includes a substrate having a micropattern thereon. The micropattern includes spaced-apart regions having a water-insoluble form of a radioactive material thereon. According to some embodiments of the invention, methods of forming a low-dose-rate (LDR) brachytherapy device include depositing a solution comprising a soluble form of a radioactive material on a substrate using a solenoid dispensing system having a controlled pressurized fluid source. The soluble form of the radioactive material is converted to a water-insoluble form of the radioactive material on the substrate. A medical device is formed from the substrate and the water-insoluble form of the radioactive material. According to some embodiments of the invention, methods of forming a low-dose-rate (LDR) brachytherapy devices include depositing a solution comprising a soluble form of a radioactive material on a bioabsorbable, polymer substrate. The soluble form of the radioactive material is exposed to a plasma treatment for a time sufficient to convert the soluble form of the radioactive material to a water-insoluble form of the radioactive material. A medical device is formed from the substrate and the water-insoluble form of the radioactive material. According to some embodiments of the invention, methods of forming a low-dose-rate (LDR) brachytherapy device include depositing a radioactive shielding layer on a bioabsorbable, polymer substrate. A solution including a soluble form of a radioactive material is deposited on the radioactive shielding layer opposite the substrate. The soluble form of the radioactive material is exposed to a plasma treatment for a time sufficient to convert the soluble form of the radioactive material to a water-insoluble form of the radioactive material. A medical device is formed from the substrate and the water-insoluble form of the radioactive material. The bioabsorbable, polymer substrate includes a region that is substantially free of the radioactive material. A portion of the region that is substantially free of the radioactive material is cut to customize a shape of the substrate for implantation. The present invention now will be described hereinafter with reference to the accompanying drawings and examples, in which embodiments of the invention are shown. This invention may, however, be embodied in many different forms and should not be construed as limited to the embodiments set forth herein. Rather, these embodiments are provided so that this disclosure will be thorough and complete, and will fully convey the scope of the invention to those skilled in the art. Like numbers refer to like elements throughout. In the figures, the thickness of certain lines, layers, components, elements or features may be exaggerated for clarity. The terminology used herein is for the purpose of describing particular embodiments only and is not intended to be limiting of the invention. As used herein, the singular forms “a,” “an” and “the” are intended to include the plural forms as well, unless the context clearly indicates otherwise. It will be further understood that the terms “comprises” and/or “comprising,” when used in this specification, specify the presence of stated features, steps, operations, elements, and/or components, but do not preclude the presence or addition of one or more other features, steps, operations, elements, components, and/or groups thereof. As used herein, the term “and/or” includes any and all combinations of one or more of the associated listed items. As used herein, phrases such as “between X and Y” and “between about X and Y” should be interpreted to include X and Y. As used herein, phrases such as “between about X and Y” mean “between about X and about Y.” As used herein, phrases such as “from about X to Y” mean “from about X to about Y.” Unless otherwise defined, all terms (including technical and scientific terms) used herein have the same meaning as commonly understood by one of ordinary skill in the art to which this invention belongs. It will be further understood that terms, such as those defined in commonly used dictionaries, should be interpreted as having a meaning that is consistent with their meaning in the context of the specification and relevant art and should not be interpreted in an idealized or overly formal sense unless expressly so defined herein. Well-known functions or constructions may not be described in detail for brevity and/or clarity. It will be understood that when an element is referred to as being “on,” “attached” to, “connected” to, “coupled” with, “contacting,” etc., another element, it can be directly on, attached to, connected to, coupled with or contacting the other element or intervening elements may also be present. In contrast, when an element is referred to as being, for example, “directly on,” “directly attached” to, “directly connected” to, “directly coupled” with or “directly contacting” another element, there are no intervening elements present. It will also be appreciated by those of skill in the art that references to a structure or feature that is disposed “adjacent” another feature may have portions that overlap or underlie the adjacent feature. Spatially relative terms, such as “under,” “below,” “lower,” “over,” “upper” and the like, may be used herein for ease of description to describe one element or feature's relationship to another element(s) or feature(s) as illustrated in the figures. It will be understood that the spatially relative terms are intended to encompass different orientations of the device in use or operation in addition to the orientation depicted in the figures. For example, if the device in the figures is inverted, elements described as “under” or “beneath” other elements or features would then be oriented “over” the other elements or features. Thus, the exemplary term “under” can encompass both an orientation of “over” and “under.” The device may be otherwise oriented (rotated 90 degrees or at other orientations) and the spatially relative descriptors used herein interpreted accordingly. Similarly, the terms “upwardly,” “downwardly,” “vertical,” “horizontal” and the like are used herein for the purpose of explanation only unless specifically indicated otherwise. It will be understood that, although the terms “first,” “second,” etc. may be used herein to describe various elements, components, regions, layers and/or sections, these elements, components, regions, layers and/or sections should not be limited by these terms. These terms are only used to distinguish one element, component, region, layer or section from another region, layer or section. Thus, a “first” element, component, region, layer or section discussed below could also be termed a “second” element, component, region, layer or section without departing from the teachings of the present invention. The sequence of operations (or steps) is not limited to the order presented in the claims or figures unless specifically indicated otherwise. As used herein, the term “globule” refers to a discrete volume of material. Globules of material can be deposited on a substrate or in a micro-well, for example, using a micro-syringe pump or micro-pipette according to embodiments of the present invention. In some embodiments, the volume of material in globule can be controlled, for example, with an accuracy of better than 10%. Typical sizes of globules are between 5 and 500 nanoliters. In particular embodiments, the globule size is between 30 and 200 nanoliters. In some embodiments, the globules can be spaced apart by about 500-1000 μm. Radiation treatment devices and fabrication methods are discussed in U.S. Application Ser. Nos. 60/823,814, filed Aug. 29, 2006; 60/847,458, filed Sep. 27, 2006; 60/926,349 filed Apr. 26, 2007; and Ser. No. 11/846,075, filed Aug. 28, 2007, the disclosures of which are hereby incorporated by reference in their entireties. As illustrated in FIG. 1, a low-dose-rate (LDR) brachytherapy device can be formed by depositing a solution including a soluble form of a radioactive material on a substrate (Block 10). The soluble form of the radioactive material (e.g., a water-soluble radioactive material) is converted to a water-insoluble form of the radioactive material on the substrate (Block 20). A medical device can be formed from the substrate and the water-insoluble form of the radioactive material (Block 30). Although embodiments according to the invention are discussed herein with respect to converting a water-soluble form of a radioactive material into a water-insoluble form, it should be understood that the solution deposited on the substrate at Block 10 may include solvents in addition to or in place of water, such as HCl. Thus, the soluble form of the radioactive material may be water-soluble or soluble in a solvent other than water without departing from the scope of the invention. The water-insoluble form of the radioactive material can be formed by various techniques, including plasma decomposition (FIG. 1; Block 22), chemical precipitation (FIG. 1; Block 24), thermal decomposition (FIG. 1; Block 26) and/or combinations thereof. For example, as shown in FIGS. 2A-2D, a substrate 50 is provided with microwells 52 therein. The substrate 50 can be a polymer substrate such as a nylon substrate. The selection of the polymer will be within the skill of one in the art. A volume of radioactive salt solution 54A is deposited in the microwell 52 as shown in FIG. 2B. The radioactive salt solution 54A can be deposited using a microsyringe, pipette or a solenoid dispensing system as described herein. The radioactive salt solution can be [Pd(NH3)4]Cl2 dissolved in ammonium hydroxide (NH4OH) and/or PdCl2 dissolved in HCl; however, other suitable solutions can be used. The salt solution 54A is optionally dried to provide a dried salt 54B as shown in FIG. 2C. The salt 54B is then decomposed into a water-insoluble radioactive material 54C as shown in FIG. 2D, for example, by chemical, thermal or plasma decomposition as described herein. For example, palladium salt, such as [Pd(NH3)4]Cl2, can be thermally decomposed at temperatures of between about 220-300° C. and/or plasma decomposition can be performed, e.g., using a 50-150 mTorr (or greater) oxygen plasma for at least about 15-30 minutes at a power setting of 200 watts. PdCl2 can be decomposed at about 675° C., and [Pd(NH3)4]Cl2 can be decomposed at about 290° C. In some embodiments, a radio marker RM can be included in one of the wells 52 (e.g., a gold marker) to increase visibility of the device during medical imaging. Although a well pattern on the substrate 50 is illustrated in FIGS. 2A-2D, any suitable substrate can be used. As shown in FIGS. 2E-2G, in some embodiments, the substrate 50′ can have a generally planar or smooth surface. The radioactive solution 54A′ can be deposited in a spaced-apart pattern on the substrate 50′. The radioactive solution 54A′ can be evaporated as shown in FIG. 2F to form a solution 54B′. The solution 54B′ is then decomposed or converted into a water-insoluble radioactive material 54C′. The substrate 50, 50′ and water-insoluble radioactive material 54C, 54C′ of FIGS. 2D and 2G can be coated with a suitable biocompatible coating and formed into a medical device, such as a elongated strand or a generally planar brachytherapy device. For example, an elongated strand substrate 50″ with a water-insoluble radioactive material 54C″ is shown in FIG. 2H. An elongated substrate 50′″ with a water-insoluble radioactive material 54C′″ in a two dimensional, spaced-apart pattern is shown in FIG. 2I. It should be understood that elongated (linear) strand substrates and planar substrates can be formed with microwells, such as the wells 52 shown in FIGS. 2A-2D, or on a substrate without a microwell pattern, such as a generally smooth surface as shown in FIGS. 2E-2G. In some embodiments shown in FIGS. 2A-2D, the solution 54A includes Pd(NH3)4Cl2, and dilute HCl (hydrochloric acid) is added to the solution 54A prior to the drying step shown in FIG. 2C. Without wishing to be bound by any one theory, when HCl is added to Pd(NH3)4Cl2, a Pd(NH3)2Cl2 precipitate is formed. A precipitant can be at least partially formed before the evaporation of the solution 54A, which can reduce uneven deposits of the salt after evaporation (e.g., ring patterns), which can affect the decomposition time in the plasma. Accordingly, adding dilute HCl to the solution 54A can lead to a more uniform dispersion of particles in the dried salt 54B of FIG. 2C, and HCl can also facilitate complete and rapid decomposition, for example, in a subsequent plasma process. In some embodiments, a deposition device, such as an isolated solenoid dispensing system (such as is available from Innovadyne Technologies, Inc., Santa Rosa, Calif. (U.S.A.)) can be used to deposit a volume of Pd(NH3)4Cl2 (e.g., 75 ml) into a well 52 followed by another volume (e.g., 25 nl) of dilute HCl. Adding HCl may produce ammonium chloride (NH4Cl); however, NH4Cl may be decomposed in the subsequent oxygen plasma and may be removed at the end of the processing. In some embodiments, the substrate 50 is treated with a plasma (such as oxygen plasma) before the radioactive salt solution 54A is deposited in FIG. 2B. In particular embodiments, the plasma treatment may be performed at a pressure of 75-100 mTorr with a power setting of 230 Watts for 2-5 minutes, such as for about 3 minutes. The amount of plasma pre-treatment can be modified to provide sufficient dispersal of the salt residue within the well 52. Without wishing to be bound by any one theory, it is believed that plasma surface treatment can function to create surface roughening and/or “activation” of surface binding sites that favor relatively even dispersal of the salt residue within the well 52 and reduce clumping. If the plasma process is performed for an excessive amount of time, the solution 54A can spread out of the well and thus reduce containment of the palladium metal within the well 52. It should be understood that the plasma pre-treatment process described above can be performed on any suitable polymer substrate. Although the techniques for converting the solution 54A into the water-insoluble radioactive material 54C is described above with respect to FIGS. 2A-2D, it should be understood that such techniques can be applied to substrates having other configurations, including the substrate of FIGS. 2F-2I. It should be understood that a radiation detector can be used to test the radiation on the resulting medical device. Devices according to embodiments of the present invention can be used as low dose radiation brachytherapy devices, such as for prostate cancer, lung cancer and/or breast cancer. In some embodiments, the water-insoluble form of the radioactive material can be formed by exposing the substrate and the water-soluble form of the radioactive material to plasma to thereby decompose the water-soluble form of the radioactive material to a water-insoluble form. In some embodiments, a hydrogen or oxygen plasma is used, typically at atmospheric pressure or in a partial vacuum. Although any suitable substrate material and/or radioactive material can be used, in particular embodiments, a polymer substrate can be used. The water-soluble form of the radioactive material can be a salt of Pd-103, such as tetraaminopalladium chloride. For example, a 50-75 mTorr oxygen plasma was found to decompose a dried residue of tetraaminopalladium chloride in about 30 minutes at a power setting of 200 watts. Elemental analysis of the post-plasma residue confirmed that only Pd metal remained. Plasma decomposition of non-radioactive water-soluble materials are described, for example, in Korovchenko et al., Catalysis Today 102-103 (2005) 133-141, and in U.S. Pat. No. 6,383,575, the disclosures of which are hereby incorporated by reference in their entireties. A brachytherapy medical device can be formed, for example, by enclosing the substrate and the water-insoluble form of the radioactive material (e.g., the substrates 50, 50′, 50″ and 50′″) with a biocompatible material. In some embodiments, polymer microwells as described herein can be used to receive drops or globules of the water-soluble radioactive material in a spaced apart pattern. Alternatively, the water-soluble radioactive material can be deposited on a substantially planar substrate, e.g., in a spaced apart pattern. After the water-soluble form of the radioactive material is converted to a water-insoluble form, a polymer sheet can be adhered on the substrate to laminate the substrate and the water-insoluble form of the radioactive material. Elongated portions of the substrate can be cut or separated to thereby form a brachytherapy strand or a planar sheet. The brachytherapy strand can be positioned in a biocompatible tube, and the tube can be filled, e.g., with a curable thermoplastic resin such as epoxy, and cured such that the radioactive material is sealed. Shielding materials can be added, for example, on one side of the device, to provide reduced irradiation on a side of the device as desired using a radiation treatment plan. Exemplary dimensions of micropatterned wells are about 250 μm wide, around 300 μm deep, and about 650-1500 μm long. Other dimensions may be about 100-400 μm wide, around 100-500 μm deep and about 650-1500 μm long. The spacing between the wells can range between about 100 μm to about 250 μm. In some embodiments, the radioactive material is selectively deposited (e.g., in globules) on the micropattern to provide non-uniform and/or discontinuous radiation pattern. Examples of suitable substrates include a suture, such as a monofilament suture, or other biodegradable or non-biodegradable material that is biocompatible and can be implanted in a patient, such as a silicon, glass or metal fiber. Biodegradable materials include, but are not limited to, polydioxanone, polylactide, polyglycolide, polycaprolactone, and copolymers thereof. Copolymers with trimethylene carbonate can also be used. Examples are PDS II (polydioxanone), Maxon (copolymer of 67% glycolide and 33% trimethylene carbonate), and Monocryl (copolymer of 75% glycolide and 25% caprolactone). Non-biodegradable materials include nylon, polyethylene terephthalate (polyester), polypropylene, expanded polytetrafluoroethylene (ePTFE), glass and metal (e.g. stainless steel), metal alloys, or the like. In some embodiments, a low-dose radiation (LDR) brachytherapy device is formed by determining a radiation profile for the device based on a patient radiation treatment plan and depositing a radioactive material on the device in a pattern. The radioactive material can include a molecularly dispersed radioisotope. The pattern can include a plurality of spaced-apart, discrete globules, each globule having a respective volume of the radioactive material. In particular embodiments, a water soluble radioactive material in a solution is deposited on the substrate, and a water-insoluble form or precipitate of the radioactive material is formed on the substrate by chemical precipitation, plasma treatment and/or thermal decomposition. A solution having a water soluble radioactive material dispersed therein can provide a known amount of radiation because the radioactive material can be evenly dispersed in the solution. Therefore, the amount of radiation deposited on the substrate is proportional to the volume of solution deposited. However, a water soluble radioactive material can present containment and/or leakage problems in medical devices because water soluble materials may leach into the body. According to embodiments of the present invention, the water soluble form of the radioactive material is converted to a water-insoluble precipitate or form, e.g., by chemical precipitation, by plasma treatment or by thermal decomposition. For example, the water-insoluble form of the radioactive material can be formed by thermal decomposition by heating the substrate, for example, at a temperature above a decomposition temperature of the radioactive material. The radioactive solution can include a palladium salt, and the decomposition temperature of the palladium salt can be about 290° C. for Pd(NH3)2Cl2. Certain polymer materials, such as silicon, can withstand temperatures around 290° C. or higher, and therefore, can be used to form the substrate. However, the decomposition temperature of PdCl2 is 675° C. In particular embodiments, a plurality of spaced-apart, hydrophilic regions are formed on a hydrophobic region of the substrate, and the solution is deposited on some of the plurality of spaced-apart, hydrophillic regions. Accordingly, the solution can adhere to the hydrophillic regions. In particular embodiments, the substrate is silicon and the plurality of spaced-apart, hydrophilic regions are silicon dioxide. In certain embodiments, a silicon substrate (optionally including a hydrophilic region of silicon dioxide) can be fixed or adhered to a polymer core. In other embodiments, forming a water-insoluble precipitate or form of the radioactive material includes chemically forming the precipitate using a precipitation solution. For example, the solution can be deposited in a plurality of spaced-apart wells on the substrate. A suitable precipitation solution can be added to the wells to chemically precipitate the radioactive material. The remaining solution can be removed, for example, by drying. In some embodiments, the radioactive material is palladium-103 (Pd-103). For example, a solution including a water-soluble form of Pd-103 can include [Pd(NH3)4]Cl2 and/or PdCl2. In particular embodiments, the solution can be [Pd(NH3)4]Cl2 dissolved in ammonium hydroxide (NH4OH) and/or PdCl2 dissolved in HCl. For example, the water-insoluble form can be formed by adding sodium borohydride (NaBH4) to the [Pd(NH3)4]Cl2 and/or PdCl2 to chemically precipitate water insoluble Pd-103. The molar concentration of NaBH4 can be at least twice the molar concentration of palladium ion. The sodium borohydride (NaBH4) can be buffered in NaOH, e.g., to stabilize the sodium borohydride and/or substantially prevent the sodium borohydride from breaking down in the water solution. It is noted that sodium borohydride generally decomposes in pure water and produces hydrogen gas and sodium borate (NaBO2); however, sodium hydroxide can lower the pH of the solution and reduce this decomposition of the sodium borohidride. In particular embodiments, a 20% (by weight) solution of NaOH and a 7.6% solution (by weight) of sodium borohydride can be used. In certain embodiments, a solution of about 2.4% of sodium borohydride and as little as 1% NaOH can be used for a 56 mg/ml PdCl2 solution. An excess of moles of sodium borohydride versus moles of palladium can be desirable so that substantially all of the palladium is precipitated (a molar ratio of at least 2, as noted above, may be sufficient). Specifically, the ratios chosen for the precipitation solution can be selected for concentrations of palladium around 56 mg/ml or more of PdCl2 (palladium chloride, which can be used as a precursor for either acid or base forms of the solution). In some embodiments, the molar concentration of NaOH is as low as feasible so as to provide buffering against rapid hydrolysis of sodium borohydride. For example, the molar concentration of NaOH can be about 0.25. In certain embodiments, the substrate includes aluminum and the solution can be deposited on the aluminum substrate to chemically precipitate water insoluble Pd-103. In other embodiments, aluminum is added to the solution to chemically precipitate water insoluble Pd-103. In some embodiments, the radioactive material comprises 1-125 and can be provided as NaI (sodium iodide). The water soluble solution can include NaI dissolved in NaOH. AgNO3 can be added to the NaI to chemically precipitate AgI. Substrates according to embodiments of the present invention can be coated with a biocompatible coating, such as a polyurethane sleeve, for example, having a thickness greater than 150 micrometers. In some embodiments, a plurality of hydrophillic regions are spaced apart by hydrophobic regions, and the water insoluble precipitate of the radioactive material is on at least some of the plurality of hydrophillic regions. For example, the substrate can be formed of silicon and the plurality of spaced-apart, hydrophilic regions can be silicon dioxide, which are optionally affixed to a polymer core. In particular embodiments, the radioactive material can include two radioisotopes having respective decay profiles. Accordingly, the spatial pattern and the at least two different decay profiles can provide a spatiotemporal radiation profile that can be fabricated to implement a radiation therapy plan for an individual patient. For example, ratio of two or more isotopes can be modified to achieve a time-varying radiation profile and can be used to increase the radiobiological effectiveness of the LDR therapy. In some embodiments, different isotopes of the same element can be used. In some embodiments, the output of conventional radiation therapy planning software or other suitable radiation therapy plans can be used to determine the spatial and/or temporal radiation profile for a device. An example of radiation therapy planning software is VariSeed™ from Varian, Inc. (Palo Alto, Calif.). The device can be fabricated using calculated amounts of radioactive material, such as a radioactive material, that can be dispersed in a spatial pattern along a length of an elongated LDR device and/or using a mixture of two or more isotopes to achieve an appropriate temporal profile. The radiation therapy plan and spatial and/or temporal radiation profile of the device can take into account the effects of post-implantation edema, e.g., by adding extra length to the device and/or increasing the radioactivity of the proximal and distal ends of the device that can be implanted at an outer boundary of the tumor or organ. In particular embodiments, the device can include a filament that can extend outside the patient after implantation. The filament can have sufficient tensile strength to allow a physician to pull the brachytherapy device in the proximal direction to readjust the position of the device after placement. Once final positioning is achieved, the filament can be severed close to the skin surface. In particular embodiments, computer program products can be used to determine a pattern of radioactive portions and non-radioactive portions of a device and/or mixture(s) of radioisotopes to create a spatial and/or temporal radiation profile when implanted in the patient and/or to control the fabrication of the brachytherapy device. Brachytherapy devices can be provided that include a polymeric material having a chemically distributed therapeutic isotope throughout. In specific embodiments, processing techniques can be used to form radioactive materials, e.g., to form polymeric fibers of the requisite dimensions for use in LDR (low-dose-rate) brachytherapy. Exemplary materials are discussed herein. However, any suitable radioactive material, including radioactive materials or materials that can become radioactive through irradiation, can be used. FIG. 3 illustrates an exemplary data processing system that can be included in devices operating in accordance with some embodiments of the present invention. As illustrated in FIG. 3, a data processing system 116, which can be used to carry out or direct operations includes a processor 100, a memory 136 and input/output circuits 146. The data processing system can be incorporated in a portable communication device and/or other components of a network, such as a server. The processor 100 communicates with the memory 136 via an address/data bus 148 and communicates with the input/output circuits 146 via an address/data bus 149. The input/output circuits 146 can be used to transfer information between the memory (memory and/or storage media) 136 and another component, such as a deposition controller, beam controller or irradiation device for selectively patterning a brachytherapy device with radiation or radioactive material. These components can be conventional components such as those used in many conventional data processing systems, which can be configured to operate as described herein. In particular, the processor 100 can be commercially available or custom microprocessor, microcontroller, digital signal processor or the like. The memory 136 can include any memory devices and/or storage media containing the software and data used to implement the functionality circuits or modules used in accordance with embodiments of the present invention. The memory 136 can include, but is not limited to, the following types of devices: cache, ROM, PROM, EPROM, EEPROM, flash memory, SRAM, DRAM and magnetic disk. In some embodiments of the present invention, the memory 136 can be a content addressable memory (CAM). As further illustrated in FIG. 3, the memory (and/or storage media) 136 can include several categories of software and data used in the data processing system: an operating system 152; application programs 154; input/output device circuits 146; and data 156. As will be appreciated by those of skill in the art, the operating system 152 can be any operating system suitable for use with a data processing system, such as IBM®, OS/2®, AIX® or zOS® operating systems or Microsoft® Windows®95, Windows98, Windows2000 or WindowsXP operating systems Unix or Linux™. IBM, OS/2, AIX and zOS are trademarks of International Business Machines Corporation in the United States, other countries, or both while Linux is a trademark of Linus Torvalds in the United States, other countries, or both. Microsoft and Windows are trademarks of Microsoft Corporation in the United States, other countries, or both. The input/output device circuits 146 typically include software routines accessed through the operating system 152 by the application program 154 to communicate with various devices. The application programs 154 are illustrative of the programs that implement the various features of the circuits and modules according to some embodiments of the present invention. Finally, the data 156 represents the static and dynamic data used by the application programs 154, the operating system 152 the input/output device circuits 146 and other software programs that can reside in the memory 136. The data processing system 116 can include several modules, including a radiation treatment planning module 120, a radiation profile control module 124, and the like. The modules can be configured as a single module or additional modules otherwise configured to implement the operations described herein for planning a radiation treatment plan, determining a spatial and/or temporal radiation profile for a device and/or controlling the deposition of radioactive material or other materials described herein (such as a precipitation solution) on a device to form a desired radiation pattern. The data 156 can include radiation data 126, for example, that can be used by the radiation treatment planning module 120 and/or radiation profile control module to design and/or fabricate a brachytherapy device. The radiation profile control module 124 can be configured to control a deposition device 125. While the present invention is illustrated with reference to the radiation treatment planning module 120, the radiation profile control module 124 and the radiation data 126 in FIG. 3, as will be appreciated by those of skill in the art, other configurations fall within the scope of the present invention. For example, rather than being an application program 154, these circuits and modules can also be incorporated into the operating system 152 or other such logical division of the data processing system. Furthermore, while the radiation treatment planning module 120 and the radiation profile control module 124 in FIG. 3 is illustrated in a single data processing system, as will be appreciated by those of skill in the art, such functionality can be distributed across one or more data processing systems. Thus, the present invention should not be construed as limited to the configurations illustrated in FIG. 3, but can be provided by other arrangements and/or divisions of functions between data processing systems. For example, although FIG. 3 is illustrated as having various circuits and modules, one or more of these circuits or modules can be combined, or separated further, without departing from the scope of the present invention. As shown in FIG. 4A, a deposition device 125 is controlled by a deposition controller (e.g., the radiation profile control module 124 via the I/O circuits 146 of FIG. 3) to form the radioactive portions 62 of the device 60. In particular, the radioactive material can be deposited in a plurality of spaced-apart, discrete globules. Each globule of radioactive material can include a particular volume of the material so that the pattern of radioactive material provide a desired radioactive profile. The globules can have relatively precisely deposited volumes between 5 and 500 nanoliters or between 10 and 200 nanoliters. Two or more radioisotopes can be used to provide a desired decay profile, which can vary along a length of the device. In some embodiments, the volume of the radioactive material can be calculated by the radiation treatment planning module 120 and/or radiation profile control module 124 of FIG. 3. Without wishing to be bound by theory, the amount of radioactive material is generally directly related to the amount of radiation emitted. For example, twice the amount of a radioactive material will generally result in twice the amount of radiation being emitted. Accordingly, in some embodiments, precision deposition can be used to deposit desired amounts of radioactive material to achieve a particular radiation profile. The deposition device 125 can be a micropipette, a microsyringe pump, or other suitable extrusion and/or deposition device, such as an Ultramicrosyringe II by World Precision Instruments, LTD, Stevenage, Hertfordshire, England. The deposition device 100 can deposit a volume of material with an accuracy of within 10% or less of the calculated volume. Although embodiments according to the present invention are described herein with respect to the deposition device 125, any suitable deposition device can be used. In some embodiments, a syringe pump can be used to aspirate a sample solution, and then a digitally controlled gas pressure system can be used to expel a controlled volume of liquid. For example, an isolated solenoid dispensing system, such as is available from Innovadyne Technologies, Inc., Santa Rosa, Calif. (U.S.A.) can be used. Such liquid dispensing solutions typically include a dispensing path that dispenses solution via an orifice, a rapidly actuating solenoid dispensing valve and a controlled pressurized liquid or fluid source. A hybrid valve can connect the dispensing path alternatively to a syringe so that the solution can be drawn into the dispensing path via the orifice when the hybrid valve is connected to the syringe. After filling the dispensing path with solution, the hybrid valve connects the dispensing path to the controlled pressurized liquid source. The pressurized liquid source enters the dispensing path, and a corresponding volume of the solution is displaced so that it exits the dispensing orifice. The amount of the pressurized liquid source entering the dispensing path can be controlled by a micro-solenoid valve and a digital pressure regulator. As the fluid leaves a dispensing orifice of the dispensing device, the velocity and energy of the fluid displacement as the fluid is displaced from the device enables the surface tension of the dispensed liquid to separate the dispensed liquid from the device. Accordingly, such devices can be referred to as “non-contact” devices because contact with the dispensing surface is not required to separate the liquid droplets from the device. Exemplary deposition devices and deposition techniques for depositing controlled volume droplets are described, for example, in U.S. Patent Publication Nos. 20070155019, 20030170903, 20030167822 and 20030072679 and in U.S. Pat. Nos. 7,135,146; 6,983,636 and 6,852,291; the disclosures of which are hereby incorporated by reference in their entireties. In some embodiments, commercially available devices, such as the isolated solenoid dispensing systems described above, can be configured to accept microtiter plates; such devices can be adapted to deposit a solution including the radioactive material on a substrate or a plurality of substrates positioned on a cassette as described with respect to FIGS. 4B-4C. When the deposition process is completed, the remaining radioactive solution can be expelled into a recycling container and the syringe tip can be cleaned in water to reduce the formation of salt crystals in or on the syringe tip to reduce or eliminate clogging. The volume of the solution can be about 75 nanoliters; however, the concentration of radioactive material/drop volume can be selected (e.g., 50-100 nanoliters) to provide a desired radioactivity (higher or lower around the central value). In some embodiments, the radioactive material is in a salt solution, such as Pd(NH3)4Cl2. One technique for controlling the concentration of the radioactive material in a volume of solution is to evaporate a solution leaving the salt residue, which is then weighed before being reconstituted into a solution using an appropriate volume of solvent (e.g., water and ammonium hydroxide) to provide the desired radioactivity per volume. The ammonium hydroxide can be present at 10% or greater (e.g., 28-30%) for a radioactive salt to dissolve. The solution can be concentrated to correspond to a range of clinically desired levels of radioactivity, and the volume of the solution drop can also be modified to enhance or reduce the activity in a given globule or well. The radioactive material can be formed using a radioactive precursor material so that it is radioactive at the time that it is deposited on the device 60 or, in some embodiments, the radioactive material can be in an inactive form during deposition and can be irradiated (e.g., by neutron bombardment) after it is deposited and/or cured to provide a radioactive device. In some embodiments, as shown in FIG. 4B-4C a plurality of substrates 60 can be positioned on a cassette 66. The cassette 66 includes grooves, and the substrates 60 are positioned in the grooves. The deposition device 125 of FIG. 4A can be used to deposit radioactive material or other materials such as a precipitation solution on the substrates 60. For example, as shown in FIG. 4B, the deposition device 125 is a micropipette having a reservoir 125A, a plunger 125B and a micropipette needle 125C. The plunger 125B pushes a desired amount of material through the needle 125C and deposits globules of the material at desired positions on the substrates 60. FIG. 4D is a schematic drawings of a cassette 66 for depositing radioactive material on the substrates. In some embodiments, the substrates 60 can be positioned on the substrate cassette 66, and a biocompatible coating or tube can be positioned in grooves on the cassette 66 or on another cassette so as to be aligned with the substrates such that the substrates can be pushed or urged into a biocompatible tube. Exemplary techniques for chemical precipitation to provide a water insoluble radioactive material will now be described. As shown in FIG. 5A, a substrate 200 includes a plurality of microwells 202. A deposition device 250A, such as a micropipette or micro syringe, deposits a radioactive material 204 in the form of a solution in the microwell 202. For example, the solution can be tetraamine palladium chloride in ammonium hydroxide or palladium chloride in hydrochloric acid. The volume of the solution of radioactive material 204 can be calculated to match the desired amount of radioactivity in the well. In FIG. 5B, a chemical precipitation solution 206 is deposited in the well on the radioactive material 204 by a deposition device 250B, such as an ink jet deposition device. For example, the precipitation solution 206 can be a mixture of sodium borohydride and sodium hydroxide in sufficient amounts for a reaction to occur to precipitate out the palladium metal, which is then insoluble. The solution 206 and any remaining amounts of the solution containing the radioactive material 204 can be allowed to dry. In FIG. 5C, the wells 202 are filled with a sealant 208, such as medical grade epoxy. The sealant 208 can then be cured, for example, by thermal or UV curing based on the type of sealant used. As shown in FIG. 5D, the substrate 200 can then be cleaned with a cleaning solution 212 using known techniques to remove any exposed radioactive material. The radioactive material 204 is in a water insoluble state and sealed by the sealant 208 to reduce or prevent leakage into the body. As shown in FIGS. 5E-5G, the substrate 200 can also be inserted into a sheath 214 to further reduce the risk of radiation leakage. As shown in FIG. 6A-6C, a radiographic marker 216, such as a gold wire, can be affixed to the ends of the substrate 200 so that the device can be more readily imaged. The radiographic marker 216 can include notches 216A for allowing sealant 208 to be injected into the sheath 214 by a sealant injector 260. As shown in FIGS. 7A-7B, the ends of the resulting device can be trimmed (FIG. 7B) and a plug 218, such as a polymeric plug, can be inserted on the ends of the device for further sealing and containment of the radioactive material 204. In some embodiments, radiographic markers can be placed directly in the wells 202. According to some embodiments of the present invention, a radioactive material in a solution can be converted to a water-insoluble form by thermal decomposition. In particular embodiments, a silicon substrate or other suitable material that can withstand the thermal processing steps can be used. For example, as shown in FIGS. 8A-8C, a carrier 300, such as a glass or ceramic carrier, includes a plurality of holders 302 for holding a plurality of substrates 310. As shown in FIG. 8C, vacuum holes 300H can be drilled in the carrier 300 to enable immobilization of the substrates 310. In particular embodiments, the substrates 310 are silicon. The density of silicon is ˜2.3 g/cc and thus is denser than polymers (which are all roughly 1 g/cc). For a 20 KeV photon (slightly less than the primary Pd-103 photons), the mass attenuation coefficient is roughly 5.3, and the mass attenuation coefficient for water is 0.81. Accordingly, a 100 μm (about 4 mils) thick layer of silicon can lead to a 5% attenuation of the photons (whereas a 300 μm layer of polymer would be about 2.5% attenuation). Small silicon chips can be even thinner than 4 mils. In some embodiments, the solution can be converted to a water-insoluble form at relatively low temperatures. For example, when Pd(NH3)4Cl2 solution dries thoroughly it forms Pd(NH3)2Cl2, which can be thermally decomposed at about 290° C. leaving palladium metal. This processing temperature is consistent with certain polymers, such as silicone, and thus presents a format whereby a polymer can be used as the substrate for the conversion of Pd salt into water insoluble Pd metal. As shown in FIGS. 8D-8F, a deposition device 350, such as a microsyringe or micropipette, can be used to deposit a radioactive material 304 in the form of a water soluble solution on the substrates 310 (FIGS. 8D-8F). Silicon is hydrophobic, and therefore, the solution may bead up. In some embodiments, the silicon surface is substantially free of other layers or materials. However, surface treatments and/or other material/layers can be used. In some embodiments, a silicon dioxide layer is on the silicon surface in a pattern. The silicon dioxide layer is hydrophillic, and therefore can provide a surface or platform for the radioactive material solution. The volume of the solution can be about 50 nanoliters or about 0.46 mm in diameter; however, the concentration of radioactive material/drop volume can be selected (e.g., 75 nanoliters, or 50-100 nanoliters) to provide a desired radioactive activity. In some embodiments, the volume of the material 304 deposited on the substrates 310 can vary, e.g., to create a variable and/or customized radiation pattern. The radioactive material 304 is then dried (FIG. 8F). For example, the radioactive material can be a palladium compound, such as tetraaminopalladium chloride. However, the resulting dried radioactive material 304 in FIG. 8F is in a water soluble form. As shown in FIG. 8G, the radioactive material 304 on the carrier 300/substrate 310 is thermally decomposed at a sufficient temperature (e.g., above about 650° C., e.g., for palladium chloride to form a water insoluble precipitate of the radioactive material. In some embodiments in which the substrates 310 are silicon substrates, a silicon compound, such as palladium silicide, can be formed when the radioactive material 304 is heated on the substrate 310. As shown in FIG. 8H-8I, a protective coating 306, such as epoxy, can then be applied to the substrate 310. A placement device 360, such as a “pick-and-place” machine, can move the substrates 310 with the water insoluble radioactive material 340 thereon to an adhesive 322 (such as epoxy) on a polymer carrier core 320. The core 320 can be formed of a suitable material that can be more flexible than the substrates 310. As shown in FIGS. 8J-8N, the polymer core 320 can be inserted into a sheath 330 (FIG. 8J), filled with a sealant 322 by a sealant injector 370 (FIG. 8K), cured and/or trimmed (FIG. 8L). Moreover, radiographic markers 380 can also be added to the polymer core 320 (FIG. 8M), e.g., for enhanced imaging visibility. For example, a radio-opaque metal, such as gold, can be used as a marker 380 for increased x-ray visibility. In some embodiments, the polymer core 320 can be cut or sized to a desired length for implantation (FIG. 8N). In particular embodiments, the core 320, sealant 322, and/or sheath 330 can be formed of a biodegradable polymer so that over time, the core 320, sealant 322, and/or sheath 330 degrades leaving the substrates 310 and radioactive material 304 implanted in the body. Additional embodiments according to the present invention are shown in FIGS. 9A-9G. As shown in FIG. 9A, a polymer substrate 402 is for the deposition of a plurality of spaced-apart, water-soluble solution droplets of radioactive material 404S. The material 404S can be deposited using various techniques described herein, such as using a syringe or micropipette. In some embodiments, the water-soluble radioactive material can be a salt of Pd-103, such as tetraaminopalladium chloride or palladium chloride. For example, the polymer substrate 402 can be 2 mil nylon 6,6 that is optionally coated with a thin layer of hydrophobic material, such as Epotek 302-3M (which may be thinned and spun onto the polymer substrate 402). Alternatively, the sheet could be PTFE, which is hydrophobic and gives rise to smaller dried residues. The water-soluble solution of radioactive material 404S is then decomposed to form a non-soluble form of the radioactive material 404NS as shown in FIG. 9B. For example, if a palladium salt is used as the material 404S, the polymer substrate 402 can be exposed to hydrogen or oxygen plasma to decompose the salt into palladium metal. Although embodiments according to the invention are described above with respect to a plasma decomposition of palladium salt into non-soluble palladium metal, it should be understood that other suitable techniques and/or radioactive materials can be used. For example, in some embodiments, a precipitation solution can be deposited on the material 402S in FIG. 9A so that the water-soluble material 402S chemically precipitates to a water-insoluble form to provide the material 404NS of FIG. 9B. As shown in FIG. 9C, a polymer sheet 406 can be adhered or laminated to the substrate 402 to form a laminated structure L as shown in FIG. 9D. As further shown in FIG. 9D, the laminated structure L can be singulated or cut or sized, for example, with a laser cutter, along cut lines C. As shown in FIG. 9E, the resulting laminated core 410, which includes the spaced-apart regions of the water-insoluble material 404NS. The laminated core 410 can be further coated or enclosed so that the water-insoluble material 404NS is sealed so as to substantially prevent leakage of the insoluble, radioactive material. For example, the laminated core 410 can include a region without the material 404NS for attaching to a thread, such as a nylon thread 412. As shown in FIG. 9E, the thread 412 is used to position the core 410 into a tube, such as a PTFE mold tube 414 with a carbothane sleeve 416. As shown in FIG. 9F, the tube 414 is then filled with epoxy 418. The epoxy 418 is cured, and the sleeve 416 and the epoxy 418 are removed from the sleeve 416 to form a sealed device as shown in FIG. 9G. Although FIGS. 9A-9G are described above with respect to a polymer sheet 406 that is cut, it should be understood that any suitable configuration can be used. Although embodiments according to the present invention have been described herein as a string or elongated member (e.g., the polymer core 320) that is implantable in the body, it should be understood that other implantable devices can be formed using the techniques described herein. For example, the substrates 310 can be inserted into a conventional metallic brachytherapy seed structure or in a polymeric brachytherapy seed. In some embodiments, implantable devices can be provided that are sized and configured to provide brachytherapy to a particular organ or region of the body. For example, radioactive material can be implanted on a polymeric sheet and implanted in the patient. As illustrated in FIGS. 10-11, a generally planar, 2D radioactive sheet 600 is shown. The sheet 600 includes a substrate 602 (which can be formed using a biodegradable or bioabsorbable material), a radiation shielding layer or gold layer 604 on the substrate 602, and a radioactive material 606 on the gold layer 602. As shown in FIG. 12, the sheet 600 can optionally include perforations 608 and/or cut marks 610. The perforations 608 can reduce the amount of substrate material implanted in the body and/or increase fluid exchange between the two sides of the substrate 602, and the cut marks 610 can be used to allow a medical professional to cut and customize a sheet for implantation in the body. Although described with respect to the sheet 600 in FIGS. 10-11, it should be understood that other configurations can be used. For example, the gold layer 604 is shown as a continuous strip; however segmented gold areas under the radioactive material 606 can also be used. Other radioactive shielding materials can be used for the gold layer 604. In some embodiments, the gold layer 604 is omitted; however, the gold layer 604 can provide radioactive shielding such that the radiation is reduced on one side of the substrate to provide substantially emissions in one direction. Reducing the radiation emitted on the side of the device adjacent the gold layer 604 and opposite the radioactive material 606 can reduce damage to health tissue in certain applications, such as lung cancer, so that the radioactive side of the sheet 600 is implanted adjacent a cancerous site. In some embodiments, the substrate 602 can be imprinted with a well pattern and the gold layer 604 and radioactive material 606 can be deposited therein. The density and/or size of the radioactive material 606 deposited on the sheet 600 can be controlled as described herein to provide a desired radioactive profile for the sheet 600. In some embodiments, a top sheet of bioabsorbable material can be laminated with bioabsorbable adhesive or otherwise affixed over the radioactive material 606 to provide a generally sealed source. In addition, the substrate 602 can include regions that are substantially free of the radioactive material 606 to provide, for example, a border for surgical attachment. Radiomarkers such as gold squares can be added at various places on the substrate 602 to facilitate radiographic imaging and/or for dosimetry assessment after the implantation is completed. The gold layer 604 can be deposited on the substrate 602 by shadow masking or using appliqués. The perforations 608 can be formed before or after the deposition of the gold layer and/or radioactive material 606. The radioactive material 606 can be deposited using the techniques described herein. For example, a palladium (Pd) salt solution can be deposited, such as with a precision deposition system (e.g., an isolated solenoid dispensing system, such as is available from Innovadyne Technologies, Inc., Santa Rosa, Calif. (U.S.A.)). The solution can be dried and/or decomposed into a water-insoluble form using the chemical, plasma and/or thermal decomposition techniques described herein. In some embodiments, the gold layer 604 can be coated with an additional thin polymer layer to provide a pre-treatable surface prior to Pd solution deposition. In addition, the order of the radioactive layer 606 and the gold layer 604 can be reversed and/or additional substrates or coating layers may be used. For example, as illustrated in FIG. 12A, the radioactive material 606 can be deposited on a substrate 602A as described herein. As shown in FIG. 12B, a gold layer 604 can be deposited on another substrate 602B. The two substrates 602A, 602B may be laminated together to provide substantially uni-directional radiation emissions as shown in FIG. 12C such that radiation is reduced on the side of the device adjacent the substrate 602A. Embodiments according to the present invention will now be described with respect to exemplary lung cancer brachytherapy treatment. Lung Cancer Treatment: Background: Lobectomy is the standard of care for patients diagnosed with early stage non-small cell lung cancer (NSCLC). However, patients affected by this disease frequently have compromised pulmonary function, which can be clearly documented with pre-operative pulmonary function studies. See M. A. Beckles, S. G. Spiro, G. L. Colice, and R. M. Rudd, “The physiologic evaluation of patients with lung cancer being considered for resectional surgery,” Chest, vol. 123, pp. 105S-114S, January 2003; J. A. Bogart, E. Scalzetti, and E. Dexter, “Early stage medically inoperable non-small cell lung cancer,” Curr Treat Options Oncol, vol. 4, pp. 81-8, February 2003; and C. T. Bolliger, “Evaluation of operability before lung resection,” Curr Opin Pulm Med, vol. 9, pp. 321-6, July 2003. As such, these patients can be predicted to be unsuitable for standard lobectomy. Multiple studies, including a prospective randomized clinical trial from the Lung Cancer Study Group (LCSG) (R. J. Ginsberg and L. V. Rubinstein, “Randomized trial of lobectomy versus limited resection for T1 N0 non-small cell lung cancer. Lung Cancer Study Group,” Ann Thorac Surg, vol. 60, pp. 615-22; discussion 622-3, September 1995.), have demonstrated the inferiority of sublobar resection alone, showing local failure rates as high as 22%. See T. A. d'Amato, M. Galloway, G. Szydlowski, A. Chen, and R. J. Landreneau, “Intraoperative brachytherapy following thoracoscopic wedge resection of stage I lung cancer,” Chest, vol. 114, pp. 1112-5, October 1998.; R. J. Landreneau, D. J. Sugarbaker, M. J. Mack, S. R. Hazelrigg, J. D. Luketich, L. Fetterman, M. J. Liptay, S. Bartley, T. M. Boley, R. J. Keenan, P. F. Ferson, R. J. Weyant, and K. S. Naunheim, “Wedge resection versus lobectomy for stage I (T1 N0 M0) non-small-cell lung cancer,” J Thorac Cardiovasc Surg, vol. 113, pp. 691-8; discussion 698-700, April 1997; R. Santos, A. Colonias, D. Parda, M. Trombetta, R. H. Maley, R. Macherey, S. Bartley, T. Santucci, R. J. Keenan, and R. J. Landreneau, “Comparison between sublobar resection and 125Iodine brachytherapy after sublobar resection in high-risk patients with Stage I non-small-cell lung cancer,” Surgery, vol. 134, pp. 691-7; discussion 697, October 2003. Additionally, the LCSG demonstrated superior pulmonary function at 12 and 18 months post sublobar resection and brachytherapy compared to the control group treated by lobectomy alone. See R. Santos, A. Colonias, D. Parda, M. Trombetta, R. H. Maley, R. Macherey, S. Bartley, T. Santucci, R. J. Keenan, and R. J. Landreneau, “Comparison between sublobar resection and 125Iodine brachytherapy after sublobar resection in high-risk patients with Stage I non-small-cell lung cancer,” Surgery, vol. 134, pp. 691-7; discussion 697, October 2003. Some investigators have reported on the multi-week fractionated delivery of post-operative external irradiation following sublobar resection. The addition of external beam radiotherapy has the increased risk of pulmonary fibrosis and radiation pneumonitis as well as a decreased incidence of local control. Patient compliance and completion of total prescribed therapy are frequent problems with high dose external beam radiotherapy. Second line primary therapy in this patient cohort is external irradiation alone, which has been shown to be inferior in terms of local control and overall survival by 15-20% or more. Therefore, patients treated with a non-surgical option have outcomes significantly inferior to surgically managed patients. Sublobar resection complimented by the intraoperative placement of a Vicryl® mesh substrate impregnated with 125I ribbons affixed in a parallel planar array designed to deliver a dose of between 100 and 120 Gy to 0.5 cm from the perpendicular plane of the implant has been investigated. See R. Santos, A. Colonias, D. Parda, M. Trombetta, R. H. Maley, R. Macherey, S. Bartley, T. Santucci, R. J. Keenan, and R. J. Landreneau, “Comparison between sublobar resection and 125Iodine brachytherapy after sublobar resection in high-risk patients with Stage I non-small-cell lung cancer,” Surgery, vol. 134, pp. 691-7; discussion 697, October 2003; and T. A. d'Amato, M. Galloway, G. Szydlowski, A. Chen, and R. J. Landreneau, “Intraoperative brachytherapy following thoracoscopic wedge resection of stage I lung cancer,” Chest, vol. 114, pp, 1112-5, October 1998. Since implementation, more than 400 patients have been treated with this technique. The local failure rates have been shown to be approximately 1-3% in properly selected patients. The radiation is delivered in a very uniform manner with 100% patient compliance. Additionally, the delivery of radiotherapy is immediate and constant over the permanent time frame of the implant. This precludes the approximately 6-week course of external radiotherapy, decreasing time of treatment, inconvenience to the patient and overall cost. The morbidity associated with this procedure is low with no increase in infection or morbidity in patients treated with brachytherapy versus those treated with sublobar resection alone. One incidence of fatal vacular rupture developed in a patient whose dose was supplemented with external irradiation (8). However, this is now relatively contraindicated and the brachytherapy procedure is currently not performed in patients who require post-operative external irradiation. These include patients with positive mediastinal nodes proven by frozen section at the time of surgery. The American College of Surgeons Oncology Group (ACOSOG) has instituted protocol Z4032 “A randomized phase III study of sublobar resection versus sublobar resection plus brachytherapy in high risk patients with non-small cell lung cancer (NSCLC), 3 cm or smaller”. “ACOSOG protocol Z4032.” One of the potential problems with brachytherapy in this circumstance is a lack of a reliable and consistent delivery substrate which can be custom fitted to the individual patient. 125I ribbons come pre-formed with 1.0 cm spacing between the seeds and 10 seeds per ribbon. This yields an effective treatment length of 9 cm. Many times this length is unnecessary and a shorter length would be preferred. In addition, there is generally a 1.0-1.5 cm spacing between each of the parallel ribbons of the 125I implant and a substrate with a more conforming size would be preferred; if one existed. Although the technique has worked well in many patients, a significant number of implants are aborted intraoperatively as the constructed mesh would be in contact with major vascular structures (excluding the aorta) or major bronchial structures. A substrate which could be custom tailored and fitted intraoperatively would provide a marked improvement in the safe and limited delivery of radiation. Additionally, one in which the radiation could be effectively delivered in one direction could significantly limit the potential for unnecessary exposure to normal organs or tissues at risk. These approaches would potentially increase the numbers of patients amenable to the limited surgical approach and potentially decrease the risks associated with the procedure from an implant that was larger than necessary. Precision deposition of radioisotope in solution can be performed as described herein on a substrate sheet that is sized and configured for implantation adjacent a patient's lung. The soluble isotope is converted into an insoluble form as described herein. Several water soluble precursor formats could be used, including tetraaminepalladium chloride: Pd(NH3)4Cl2. This material can be obtained commercial with a nominal specific activity and activity per unit volume of solution. The material is first evaporated, leaving behind the palladium salt, and then reconstituted to a concentration that yields the targeted activity per unit volume needed for the desired air kerma strength of the finished device. That is, the desired precision in apparent activity is achieved by controlling the volume and activity per unit volume of a molecularly dispersed compound of the Pd-103 isotope. There are several commercial dispensing systems capable of producing drops sizes in the 100 nl range. This process uses a microsyringe pump assembled with an x-y-z stage to allow for precision placement of drops onto a substrate. Once dried, the 4-ammonia Pd salt loses two ammonia molecules to become Pd(NH3)2Cl2. This compound will decompose, thermally, at about 290° C. See L. A. Solov'yov, A. I. Blokhin, M. L. Blokhina, I. S. Yakimov, and S. D. Kirik, “Powder diffraction study of the crystal structure of trans-Pd(NH3)2Cl2,” Journal of Structural Chemistry, vol. 38, 1997. Although certain polymers, like silicone, will withstand this temperature, some will not and this can be limiting in terms of a choice of materials. The 2-ammonia salt can also be decomposed in an oxygen plasma and this is the technique can also be used to create the subject devices. A planar-type reactor is used with a power level of ˜200 watts. The oxygen pressure in the chamber is typically 100 mTorr. The yellowish salt of palladium visually turns black (non-water soluble Pd metal) in the plasma. As noted below, assays of the material after treatment confirm that the salt has been decomposed (chlorine and ammonia go off as gases) and only metallic Pd remains. It is also possible to use an optical spectrometer to record the evolution of the gas species, thus creating an endpoint detection mechanism for process monitoring. Once cured, the devices can then be trimmed to final length and are then ready for dosimetric evaluation and sterilization (e.g., in ethylene oxide (ETO)). It should also be noted that gold radiographic markers can be inserted into one or more of the nylon wells to enable visualization in clinical x-ray imaging runs (gold is not placed in wells that contain pd metal). Because of the extended length of this source, a well chamber with a long “sweet spot”, the IVB-1000, is used so that the air kerma strength of the entire source can be acquired. Additionally, the source can be moved along a cylindrical aperture in the chamber to assure that the activity per unit length is within tolerance. Note that it is possible with this method to vary the activity per unit length, which opens the possibility of a truly customized source. The IVB-1000 well chamber will be maintained in calibration with reference to a VAFAC (12) device. Monte Carlo simulations of this source can also be performed. To check for sealed-source integrity, a source is soaked in water (ISO 9978:1992(E)) for 4 hours at 50° C. and then the soak solution is assayed. In the test results summarized below, a scintillation fluid is added to the soak solution and then it is placed in an integrating sphere for event counting. The ISO metric is activity less than 5 nCi after this assay. Devices were evaluated according to the ANSI N43.6-2007 standards to ensure that physical damage caused by various factors would not result source leakage (specifically, the device is designated ANSI 07C22XX1X (X)). Results: Biocompatibility Testing: Toxikon (Bedford, Mass.) evaluated non-radioactive analogs of finished devices. Obviously, radioactive devices would create significant tissue damage, thus masking any effects of the materials (though the devices were “aged” by administering 7 kGy of gamma radiation prior to animal testing). Tests were completed for: cytotoxicity, sensitization, irritation, and systemic toxicity and all results were negative for induced changes relative to controls. Long-term (subcutaneous) implantation tests were performed for 2, 4, 13, and 26 weeks with histopathology being performed after explantation. There were no changes recorded relative to controls. Also, an in-house image study of the explanted devices showed no apparent degradation of the polyurethane or epoxy materials. Therefore, it was concluded that the materials are safe for permanent implantation in the current device configuration. Materials Analysis: Development of the oxygen plasma process progressed through the use of a non-radioactive mimic of the Pd salt solution. This allowed for the use of standard assay tools to hone the process parameters. In particular, an x-ray emission analyzer (EDS) was used to obtain elemental concentrations of the processed salt. As shown below, well processed samples only had Pd remaining whereas incompletely processed material gave a measurable chlorine peak (the assay was not really sensitive to ammonia, nitrogen). This tool was used to establish proper drying and process time values so that consistent conversion of the salt to metal was obtained. It should be noted that the effective density of the converted Pd metal residue is significantly lower than solid, sintered metal. This is because the dried crystals, initially, have significant amounts of trapped water and further the crystals are highly dendritic. When the salt is decomposed, the Pd grains appear to be sub-micron in size and are loosely held in a spongy matrix. In-house testing showed that there is minimal self-attenuation of the “sponge,” thus simplifying the task of achieving the target activity. Monte Carlo Simulation: The device geometry was evaluated using the MCNP5 code. As noted above, the effective density of the unconsolidated Pd metal residue is much lower than that of the pure metal. The simulation assumed a thin layer of metal in the well with an effective density of a few percent of dense metal value. This simulation will be re-run with different dispersions of the metal to check for consistency. Because of the close placement of the wells, the source will be indistinguishable from a purely linear source. The MC predictions can be evaluated with empirical measurements, e.g., using a VAFAC electrometer. Sealed Source Testing: The basic sealed-source test conditions were described above. The devices were subjected to ANSI standard testing, as well as certain special tests, as listed below: A temperature test from −40° C. (20 min) to 80° C. (1 hour). This is to simulate changes in temperature beyond which the finished devices would ever experience. An external pressure test from 3.6 psi (vacuum) to atmospheric (simulate possible air travel). A special impact test involving a “heel crush” to simulate the device being stepped on by a user (155 pounds for 10 seconds). A special puncture test using forceps to simulate excessive gripping of the device by a user (5 pounds of force using serrated metal tweezers). A special bending test. The device is bent over a cylinder 5D in diameter, where D is the device diameter. A non-ANSI test under autoclave conditions to assess whether any leakage of the isotope occurs. A non-ANSI test wherein the device is cut in half (in the middle of a well) to assess the potential for contamination should the device be accidentally cut. A non-ANSI test in which each of the devices is soaked again for 1 week (at room temperature), after which the soak solution is again assayed for any contamination. All of the tested samples were checked, as described above, for evidence of leakage. The only sample that showed a reading above background was the one that was purposely cut in half through a well containing Pd metal (roughly 35 nCi of leakage, still quite small). Most probably the soaking action loosened Pd grains that simply fell out into the solution. All of the samples were then soaked for an additional period of 1 week at room temperature in saline and the scintillation study was performed again. As before, only the purposely cut sample showed any signal above background (this time ˜10 nCi). Therefore, the data indicate that the sealing method is effective. Linear Sources: A linear, polymeric source can provide the following attributes: Polymer composition reduces intersource attenuation (13) A linear source can achieve the same DVH as seeds with ˜20% less apparent activity (14) The dose fall-off has a different length dependence RE seeds, which should provide greater immunity to dose distortion, e.g. from edema or needle misplacement (15, 16) The activity per length can be changed as desired Placement of markers can be optimized for the application Precision definition of activity per unit length Straightforward assembly resulting in “stranded” sources Physically robust materials Low cost-of-goods due to simple processing steps and significantly less Pd-103 required. The flexibility of the production method can be extended to a 2-dimensional source type so as to provide a type-match for post lung resection brachytherapy. Applicability to 2D Bioabsorbable Source: The deposition system described herein can be used for a flat, 2-dimensional substrate, and a desired activity per unit area can be established The plasma oxidation method can be used irrespective of the distribution of the Pd salt precursor (as long as the deposit size is not too large, which can extend processing time beyond acceptable limits) and irrespective of the polymer used for the substrate. A 2D radioactive sheet can be formed with sealed source integrity on a bioabsorbable substrate, The deposition pattern and/or shape of the device can be matched to the desired therapy plan values. Monte Carlo (MC) modeling will be performed using N-Particle 5 (MCNP5) transport code with updated DLC-146 photon cross sections. Where possible, specific atomic compositions of the materials will be used to ensure an accurate end result. The two primary goals under the modeling task are: 1) evaluate the isodose profile of the 2D sheet and 2) provide relative dose estimates (as a function of location) so as to mesh with the empirical measurements to be performed. The linear source aspects already investigated for the 1D source will be taken advantage of in the 2D design. Various Pd metal dot spacings will be evaluated. In some embodiments, the user can physically trim the 2D source sheet in the operating room as needed based on the patient's specific anatomy after the resection is performed, which can not be feasible with current seed-based arrangements. For this to be achieved safely, cut lines will be indicated on the sheet so as to avoid slicing through the Pd metal regions themselves. The effect of such cuts will be evaluated using MC modeling as well. About fifteen to twenty microns of gold will attenuate the primary Pd-103 x-rays by 90% or more. Therefore, a design that incorporates gold discs of this thickness under the sites where Pd metal is placed will create a significant anisotropy in the dose on one side of the 2D sheet versus the other. The appeal of this approach is to mitigate dose delivered to collateral tissue while maintaining adequate, dose to the surgical site. Using standard photolithographic patterning, patterning gold in the necessary array of dots is readily achievable. Gold is well known to be biocompatible and the thickness needed will not be visible on diagnostic x-rays. The exact size of the gold dot relative to the Pd metal dot needed to achieve this effect will be established with MC modeling. Sample devices can be made and empirical dose measurements can be conducted, for example, using non-absorbable substrates (e.g., nylon 6,6) to construct sample devices for evaluation with a lung phantom. Empirical dose measurements can be made at selected points in the phantom to provide quantitative validation of the Monte Carlo model predictions. Fabrication: For efficiency, 2D sources will be made with a nylon substrate for dosimetric tests while candidate bioabsorbable materials are evaluated in parallel. The techniques described herein to produce the linear sources will be applied to a 2D substrate. Specifically, the substrate will be affixed to a plastic cassette and a pattern of Pd salt dots will be applied using the microsyringe deposition system. After drying, the salt will be decomposed in oxygen plasma in situ. (For the anisotropic test device, gold dots will be pre-patterned on the substrate prior to the application of Pd material.) Previously, the endpoint (full conversion to Pd metal) was determined empirically by running the plasma process at varying times and using EDS analysis to search for salt remnants (specifically a chlorine peak). An optical emission spectroscopy system can be used for endpoint detection. This system, produced by Ocean Optics (Dunedin, Fla.), will enable the process to be monitored over time and stopped when the decomposition products are exhausted. For the purposes of these tests, the Pd metal regions will be covered with a thin layer of medical grade epoxy to seal them. This is effectively how the 1D devices were sealed and validated. These materials will be input into the MC model to ensure a proper match to the empirically derived dose readings. Additionally, gold squares of sufficient thickness to be seen with diagnostic CT will be added to the 2D source so as to enable post-implantation identification of the source location for use in clinical dosimetry (generally ˜30 days after the surgery). Dosimetry: In this task the Monte Carlo calculations will be validated in a phantom following the American Association of Physicists in Medicine Task Group Report 43 (AAPM-TG43) dosimetry protocol. Because the depth of interest in the lung brachytherapy procedure is 0.5 cm, we propose to measure the relative longitudinal dose distribution at that distance in water and lung equivalent phantom material. The Pd103 rectangular substrate sheet will be sandwiched between two slabs of 0.5 cm thick material that are equivalent to water in the first case and lung in the second case in terms of dose deposition from keV-range photons. One radiochromic film sheet will then be placed on each side of the setup, to measure the dose distribution on both sides of the Pd103 sheet. An additional 2 cm of water or lung equivalent material will be added on top of each film to account for backscatter dose. The films will then be processed, digitized and analyzed using the RIT software (Radiological Imaging Technology, Colorado Springs, Colo.) to obtain a relative dose distribution. The radial distribution relative to the depth of 5 mm will be measured in an acrylic slab phantom (PTW type 2962, New York, N.Y.) using a parallel plate chamber (PTW model 23342). The measurements will be made at distances of 1 to 50 mm above the center of the Pd103 sheet, in 2 mm increments. Additionally an absolute dose measurement will be made with the parallel plate chamber at a point 0.5 cm above the center of the Pd103 sheet. All previously described measurements will be repeated for the anisotropic source sheet, which has a higher emission on one side by attenuating the Pd103 using patterned gold dots on the one side. Measurements will also be performed in a similar manner using the standard I-125 vicryl mesh of the same size as the Pd103 patch to allow comparison of dose distributions. Several candidate bioabsorbable materials can be evaluated for biological half-life versus sealed-source containment needs. Common bioabsorbable materials Materials: Copolymers and homopolymers of glycolic acid (GA) and L-lactic acid (LA) have generally excellent toxicological histories when used in medical implants. The focus for the current evaluation will be on copolymers using a blend of these two base materials. Vicyrl (Polyglactin 910), for instance, is formed with a 90:10 GA-to-LA blend. Another material described in the literature (17) for orthopedic uses is a mixture of 18:82 GA-to-LA blend to achieve longer-term stability in the body. 90:10 of PA:LApoly-GApoly-LA(vicryl)Period until polymer60-90 days3-5years70 daysmass becomes zero(at 37° C. in saline)Period until tensile14-21 days6-12months21 daysstrength of polymerdrops by 50% (at37° C. in saline) Various commercially available materials can be used, In addition to Vicryl, there are many other materials that have been investigated such as: Atrisorb, Resolut, Lactosorb, etc. Pd-103 has a 17 day half-life and thus at 60 days, the source strength is down over 90%. For 1-125, the half-life is 60 days and thus it takes 200 days for the source strength to decay by 90%. Thus, the task of balancing radioactive half-life with bioabsorbability may favor Pd-103. However, 1-125 has thus far been used exclusively for brachytherapy after sub-lobar lung resection. In terms of radiobiological differences between the isotopes, the most extensive comparative data is from use in the prostate gland (18). Despite theoretical arguments (using the linear-quadratic model) that would favor Pd-103 for more rapidly dividing cell types, there is no clear evidence of differences in outcomes. Further, the anisotropic sheet approach can allow for a net reduction of dose rate to surrounding tissues and organs. Accelerated Life Tests and Sealed-Source Integrity: Accelerated life testing in physiologic saline at temperatures in the 45-50° C. range will be undertaken. Typically this temperature range will provide a 2.5-3.5 times increase in the hydrolytic degradation of bioabsorbable polymers (19). The tests will investigate structural as well as sealed-source integrity. For the former, a non-radioactive analog of the Pd salt precursor will be used so that measurements of dissolution using HPLC (high performance liquid chromatography can be performed without the risk of contaminating equipment. Identically arranged test fixtures will incorporate Pd-103 so that soak samples can be taken and added to a scintillation fluid for counting purposes (to determine the extent of any radiation leakage into the soak solution). Smooth layers (at least locally under the Pd material) of the substrate material are needed to contain the palladium grains that form after plasma oxidation. Thus a mesh, per se, may not be useful. Though all of the polymers of interest will be etched by the oxygen plasma, the rate of material removal is small (microns). Devices sealed with medical grade epoxy will be used as controls, as those materials will not degrade over the time scale of interest and have been proven to work successfully in our process to maintain sealed-source integrity. The following sequence will be used for each test arm (degradation and sealed-source): Soak test structure at elevated temperature in physiologic saline for 3-7 days (based on a determination of degradation acceleration at the test temperature) Remove the test structure (unless it is no longer self-supporting) and place it in fresh soak solution Assay the old soak solution by HPLC or in a counting rig as appropriate Plot these data to obtain a rough kinetics profile of physical and sealed-source degradation with time (referred to 37° C.). The foregoing is illustrative of the present invention and is not to be construed as limiting thereof. Although a few exemplary embodiments of this invention have been described, those skilled in the art can readily appreciate that many modifications are possible in the exemplary embodiments without materially departing from the novel teachings and advantages of this invention. Accordingly, all such modifications are intended to be included within the scope of this invention as defined in the claims. Therefore, it is to be understood that the foregoing is illustrative of the present invention and is not to be construed as limited to the specific embodiments disclosed, and that modifications to the disclosed embodiments, as well as other embodiments, are intended to be included within the scope of the appended claims. The invention is defined by the following claims, with equivalents of the claims to be included therein.
	description	This application claims the priority of U.S. Provisional Patent Application Ser. No. 61/521,040, filed Aug. 8, 2011, which application is incorporated by reference into the present application in its entirety. 1. Technical Field This disclosure relates to methods and apparatus for processing and analyzing a microscopic sample, in particular, methods and apparatus for such processing and analyzing inside a charged-particle instrument such as a focused ion-beam microscope (FIB) or scanning electron microscope (SEM). 2. Background FIB processes for lamella creation for transmission electron microscope (TEM) sample preparation have long been used in the semiconductor industry, including the use of nanomanipulators for FIB in situ lift-out (INLO) sample preparation, where the sample is lifted from some substrate, such as a semiconductor wafer. In addition to thin lamella, INLO samples have been adapted to other types of geometries such as wedges and micropillars, and they have been adapted to INLO-TEM types of in situ specimen analysis performed directly within the chamber of the charged-particle microscope. Examples of these analysis types include STEM, EDS, and EBSD, to name a few. The traditional INLO sample preparation is comprised of three main steps: 1) by induced-beam CVD deposition, glue, static, or other attachment means, attaching a nanomanipulator end-effector such as a fine probe tip to the sample destined for analysis, 2) while attached and supported by the end-effector, lifting the sample out or away from its original position, and 3) attaching the lifted sample to a new substrate or holder, such as a TEM grid, for completion of processing such as sample shaping by the ion beam or for inspection by various analytical means (such as the previously mentioned STEM, EDS, EBSD, or TEM analysis), performed either in situ or ex situ to the original charged-particle microscope chamber. There are three primary reasons INLO samples are attached to a secondary support prior to further processing and analysis. First, this provides a means to easily manipulate a sample into different orientations to achieve the desired processing or analysis result based on using the degrees of freedom of the charged-particle beam microscope stage. Second, by placing the sample on a support directly connected to the microscope stage, additional drift or vibration effects outside those of the stage can be avoided. Third, the secondary holder or support provides an easy way to handle the sample for storage or when transferring the sample between different instruments. The imaging requirements of charged-particle beam instruments are demanding, especially when imaging at the upper resolution limits in the range of angstroms to a few nm. Even the smallest vibration or drift can interfere with processing and analysis. In some cases, to achieve the desired performance, every unnecessary accessory on a charged-particle beam microscope is removed, as each accessory adds its own amount of drift and vibration to the entire chamber. Nanomanipulators, being accessories to these charged-particle beam instruments, have their own characteristic drift and vibration. These disturbances must be minimized to achieve the desired results if the processing and analysis are to occur while the sample is held by the end-effector of the nanomanipulator. In practice, the sample is placed on a secondary support after the orientation steps are completed to provide stability against drift and vibration during processing, imaging and analysis. The lack of a solution to perform processing and analysis while on the end-effector means valuable instrument time is consumed with the third INLO step of attaching the lifted sample to a secondary support, such as a holder. What is needed is a means to accomplish processing and analysis after the second lift-out step of attaching the lifted sample to the end-effector of the nanomanipulator without the need for attachment of the sample to a secondary holder. This application uses the term “FIB” or “charged-particle beam instrument” generically for any kind of instrument using one or more radiation beams to assist chemical vapor-deposition procedures, etch, image or lift-out specimens in a vacuum. These terms as used here thus include instruments using ion beams, electron beams, other charged-particle beams, or light energy, such as a beam of laser light, or any combination of these beams. Unless otherwise stated, the terms “end-effector”, “probe tip” or “tip” refer to any part of a manipulator apparatus intended to be attached to a specimen for lift-out or manipulation and are equivalent in this disclosure. A suitable nanomanipulator system is the AutoProbe® 300, manufactured by Omniprobe, Inc. of Dallas, Tex. In the Omniprobe apparatus, the end-effector is typically a fine tungsten needle probe tip. A solution to the requirement for reduced drift and vibration is provided by moving the end-effector holding the lifted sample until it touches a support structure that is fixed to the microscope stage (100), and continuing to move the end effector against the support until a sufficient pushing force is obtained, where no discernible vibration is observed. With this method, no vibration is observed even at magnifications of up to 400,000 times, and accurate processing and analysis can be performed while the sample is held on the tip. The flow chart in FIG. 3 shows the steps of the method. The method may start optionally with the step 310 of Pre-Collection Sample Orientation. Before collection of the sample (120) from the substrate (not shown) with the nanomanipulator tip (110), the support, such as the microscope stage (100) supporting the substrate, can be positioned so that after the sample (120) has been collected by the nanomanipulator tip (110), the position of the sample (120) on the nanomanipulator tip (110) is optimized for subsequent processing, or is suitable for final orientation using the nanomanipulator after sample (120) collection. In step 320, the Collect Sample step of the process, the sample (120) is collected from the substrate by attaching the sample (120) to the nanomanipulator tip (110). This attachment can be made, for example, with static electric attraction, an adhesive material, mechanical gripping, or by material deposition using charged-particle beam assisted material deposition with a chemical vapor, or charged-particle beam induced redeposition from a neighboring solid material. The means of attachment are not limited to the above examples. Once collected, the nanomanipulator can be used to translate the sample (120) to a suitable location for later orientation and preparation. At step 330, the sample (120) may optionally be oriented. After the sample (120) has been collected onto the nanomanipulator tip (110), the nanomanipulator tip (110) can be positioned to optimize subsequent processing using the nanomanipulator's X, Y and Z orthogonal axes, rotation about the nanomanipulator tip (110) axis, and pitch (tilt) of the nanomanipulator tip (110) about a pitch axis that is perpendicular to the nanomanipulator tip (110) axis. At step 340, the tip is stabilized. FIGS. 1-3 show the process. After the sample (120) has been collected, and optionally the orientation of the sample (120) on the nanomanipulator tip (110) has been optimized, the nanomanipulator tip (110) can be mechanically stabilized by bringing the nanomanipulator tip (110) into physical contact with a stabilizing support (130). This can be accomplished by translating the nanomanipulator tip (110) to the stabilizing support (130), moving the stabilizing support (130) to make contact with the nanomanipulator tip (110) or a combination of both movements. At step 345, the sample (120) is viewed to determine if any vibration of the sample (120) is discernible. If so, control returns to step 340 to continue to move the nanomanipulator tip (110) against the support (130); if not, then the sample (120) is ready to be prepared at step 350. The stabilizing support (130) can be a passive straight edge (150), or, in another embodiment (140), a passive straight edge with a V notch (160) as shown in FIG. 2, or a more complicated shape designed to passively minimize vibration and drift of the sample relative to the charged-particle beam, including shapes that mechanically capture the nanomanipulator tip (110). Also, the stabilizing support (130) can provide active drift or vibration minimization using feedback control from the charged-particle beam image or from a sensor or sensors attached to the nanomanipulator, to the sample stage (100) or to the charged-particle beam microscope. At step 350 a decision is made to either prepare the sample (120) or analyze it at that time. The sample preparation (step 360) can include changing the sample (120) shape or changing the sample (120) properties, such as by annealing the sample (120) using heat. The particle beam can be used to reshape the sample (120) to optimize it for subsequent analysis. The reshaping can include elimination of material by ion beam milling, for example. The reshaping might also include adding material with charged-particle beam assisted deposition, for example, if a particular shape is required for subsequent analysis. In step 370, the sample (120) is analyzed. Once prepared, the sample (120) can be analyzed immediately in the charged-particle beam microscope, such as with scanning transmission electron microscopy (STEM) or energy dispersive X-ray analysis (EDS), or the sample (120) can be removed and taken to a separate instrument for analysis, such as to a transmission electron microscope (TEM). None of the description in this application should be read as implying that any particular element, step, or function is an essential element which must be included in the claim scope; the scope of patented subject matter is defined only by the allowed claims. Moreover, none of these claims are intended to invoke paragraph six of 35 U.S.C. Section 112 unless the exact words “means for” are used, followed by a gerund. The claims as filed are intended to be as comprehensive as possible, and no subject matter is intentionally relinquished, dedicated, or abandoned.
	abstract	A grid (13) for supporting nuclear fuel pencils (3) for a nuclear fuel assembly (1) comprising a peripheral belt (17), the peripheral belt (17) comprising on at least one of its edges (35, 37) guide fins (33) is disclosed. The edge (35, 37) of the peripheral belt (17) has between the adjacent guide fins (33) recesses (39) towards the inside of the grid (13). The invention is applicable, for example, to pressurized water reactors.
	claims	1. residual heat removal ventilation system for spent fuel dry storage facility of nuclear power plant, characterized in that it is disposed in a spent fuel building, and the spent fuel building comprises a storeroom, an operating room, and a ventilation equipment room sequentially arranged from bottom to top; a ventilation heat shield cylinder is disposed in each silo of the storeroom comprising a buffer storage region, a buffer storage region cold air intake chamber and a buffer storage region hot air removal chamber separated by a buffer storage region cold and hot air chamber separator are disposed at an upper part of the buffer storage region, wherein, the buffer storage region hot air removal chamber is located above the buffer storage region cold air intake chamber;a buffer storage region cold air pipe is connected to an upper part of the buffer storage region cold air intake chamber, the buffer storage region cold air pipe extends from the operating room to the ventilation equipment room, and a buffer storage region cold air inlet connected with the buffer storage region cold air pipe is disposed on the wall of the operating room, such that cold air enters into the buffer storage region cold air intake chamber from the buffer storage region cold air pipe, enters into a cold air channel of silos of the buffer storage region through the buffer storage region cold air intake chamber, enters into the ventilation heat shield cylinder from an air inlet of the ventilation heat shield cylinder, and enters into the buffer storage region hot air removal chamber through an air outlet of the ventilation heat shield cylinder; a buffer storage region first heat removal pipe is connected to an upper part of the buffer storage region hot air removal chamber, the buffer storage region first heat removal pipe extends from the operating room to the ventilation equipment room, and connect with a buffer storage region ventilation heat removal vent on the top of the ventilation equipment room through a first air valve;a buffer storage region second heat removal pipe, an air cooling equipment, a buffer storage region heat removal fan and a bypass air pipe are disposed in the ventilation equipment room; the buffer storage region second heat removal pipe, the air cooling equipment, and the bypass air pipe are connected in parallel with the buffer storage region first heat removal pipe respectively; the buffer storage region second heat removal pipe is connected with the buffer storage region ventilation heat removal vent through a second air valve; a third air valve and a fourth air valve are connected to the inlets of the air cooling equipment and the bypass air pipe respectively, the outlets of the air cooling equipment and the bypass air pipe are both connected into an inlet of the buffer storage region heat removal fan, and an outlet of the buffer storage region heat removal fan is connected to the buffer storage region first heat removal pipe and the buffer storage region second heat removal pipe through a fifth air valve and a sixth air valve respectively. 2. The residual heat removal ventilation system for spent fuel dry storage facility of nuclear power plant of claim 1, characterized in that the storeroom further comprises a second-period intermediate storage region isolated from the buffer storage region, a second-period intermediate storage region cold air intake chamber and a second-period intermediate storage region hot air removal chamber, which are separated by a second-period intermediate storage region cold and hot air chamber separator, are disposed at an upper part of the second-period intermediate storage region, and the second-period intermediate storage region hot air removal chamber is located above the second-period intermediate storage region cold air intake chamber;a second-period intermediate storage region cold air pipe is connected to an upper part of the second-period intermediate storage region cold air intake chamber, the second-period intermediate storage region cold air pipe extends from the operating room to the ventilation equipment room, and a the second-period intermediate storage region cold air inlet connected with the second-period intermediate storage region cold air pipe is disposed on the wall of the operating room, such that cold air enters into the second-period intermediate storage region cold air intake chamber through the second-period intermediate storage region cold air pipe, enters into a cold air channel of silos of the second-period intermediate storage region through the second-period intermediate storage region cold air intake chamber, enters into the ventilation heat shield cylinder from an air inlet of the ventilation heat shield cylinder, and enters into the second-period intermediate storage region hot air removal chamber through an air outlet of the ventilation heat shield cylinder; a second-period intermediate storage region heat removal pipe is connected to an upper part of the second-period intermediate storage region hot air removal chamber, the second-period intermediate storage region heat removal pipe extends from the operating room to the ventilation equipment room, and connect with a second-period intermediate storage region ventilation heat removal vent on the top of the ventilation equipment room through a seventh air valve. 3. The residual heat removal ventilation system for spent fuel dry storage facility of nuclear power plant of claim 2, characterized in that the storeroom further comprises several first-period intermediate storage regions separated from the buffer storage region and the second-period intermediate storage region;a first-period intermediate storage region cold air intake chamber and a first-period intermediate storage region hot air removal chamber separated by a first-period intermediate storage region cold and hot air chamber separator are disposed at an upper part of the first-period intermediate storage region, and the first-period intermediate storage region hot air removal chamber is located above the first-period intermediate storage region cold air intake chamber; a first-period intermediate storage region cold air pipe is connected to an upper part of the first-period intermediate storage region cold air intake chamber, the first-period intermediate storage region cold air pipe extends from the operating room to the ventilation equipment room, and a first-period intermediate storage region cold air inlet connected with the first-period intermediate storage region cold air pipe is disposed on the wall of the operating room, such that cold air enters into the first-period intermediate storage region cold air intake chamber through the first-period intermediate storage region cold air pipe, enters into a cold air channel of silos of the first-period intermediate storage region through the first-period intermediate storage region cold air intake chamber, enters into the ventilation heat shield cylinder from an air inlet of the ventilation heat shield cylinder, and enters into the first-period intermediate storage region hot air removal chamber through an air outlet of the ventilation heat shield cylinder; a first-period intermediate storage region heat removal first pipe is connected to an upper part of the first-period intermediate storage region hot air removal chamber, the first-period intermediate storage region heat removal first pipe extends from the operating room to the ventilation equipment room, and connect with a first-period intermediate storage region ventilation heat removal vent on the top of the ventilation equipment room through an eighth air valve;a first-period intermediate storage region heat removal second pipe, a first-period intermediate storage region heat removal fan and a ventilation pipe are disposed in the ventilation equipment room; the first-period intermediate storage region heat removal second pipe and the ventilation pipe are connected in parallel with the buffer storage region first heat removal pipe, respectively; the first-period intermediate storage region heat removal second pipe is connected with the first-period intermediate storage region ventilation heat removal vent through a ninth air valve; the ventilation pipe is connected with the first-period intermediate storage region heat removal fan, and an outlet of the first-period intermediate storage region heat removal fan is connected to the first-period intermediate storage region heat removal first pipe through a tenth air valve and to the first-period intermediate storage region heat removal second pipe through an eleventh air valve. 4. The residual heat removal ventilation system for dry storage of spent fuel of nuclear power plant of claim 3, characterized in that the buffer storage region heat removal fan and the first-period intermediate storage region heat removal fan both comprise two working heat removal fans and one backup heat removal fan. 5. The residual heat removal ventilation system for spent fuel dry storage facility of nuclear power plant of claim 3, characterized in that the buffer storage region is separated from the first-period intermediate storage region by a first partition, and the first-period intermediate storage region is separated from the second-period intermediate storage region by a second partition. 6. The residual heat removal ventilation system for spent fuel dry storage facility of nuclear power plant of claim 3, characterized in that the number of the buffer storage region cold air inlet, the second-period intermediate storage region cold air inlet, the first-period intermediate storage region cold air inlet, the buffer storage region ventilation heat removal vent, the second-period intermediate storage region ventilation heat removal vent and the first-period intermediate storage region ventilation heat removal vent is multiple. 7. The residual heat removal ventilation system for spent fuel dry storage facility of nuclear power plant of claim 3, characterized in that an operating room roof is disposed at the top of the operating room, and a supporting beam is disposed below the operating room roof; the buffer storage region cold air pipe, the buffer storage region first heat removal pipe, the buffer storage region second heat removal pipe, the first-period intermediate storage region cold air pipe, the first-period intermediate storage region heat removal first pipe, the first-period intermediate storage region heat removal second pipe, the second-period intermediate storage region cold air pipe and the second-period intermediate storage region heat removal pipe are all disposed at the same side, wherein, the buffer storage region cold air pipe, the buffer storage region first heat removal pipe, and the buffer storage region second heat removal pipe are disposed in correspondence to the buffer storage region, the first-period intermediate storage region cold air pipe, the first-period intermediate storage region heat removal first pipe, and the first-period intermediate storage region heat removal second pipe are disposed in correspondence to the first-period intermediate storage region, and the second-period intermediate storage region cold air pipe and the second-period intermediate storage region heat removal pipe are disposed in correspondence to the second-period intermediate storage region. 8. The residual heat removal ventilation system for spent fuel dry storage facility of nuclear power plant of claim 1, characterized in that a silo mouth floorslab is disposed at the top of the storeroom, a silo mouth corresponding to each silo is disposed on the silo mouth floorslab, and each of the silo mouth is provided with a silo plug accordingly; a bottom extremity of the ventilation heat shield cylinder is fixed on the floor of the spent fuel building, and a top of the ventilation heat shield cylinder extends to the outlet of the silo under silo mouth floorslab; an air inlet of the ventilation heat shield cylinder is arranged circumferentially along the bottom of the ventilation heat shield cylinder; the silo guide-rails extending from the floor of the spent fuel building to the silo mouth floorslab are fixed with the internal side of the ventilation heat shield cylinder, the silo guide-rails are used for the guidance of the storage canister in the hoisting process, and the silo guide-rails are fixed to the silo outer wall by lateral supporting; and the ventilation heat shield cylinder and the silo guide-rails are both segmented. 9. The residual heat removal ventilation system for spent fuel dry storage facility of nuclear power plant of claim 1, characterized in that multilayer anti-seismic floorslabs are disposed between the silos and the outer wall of the spent fuel building. 10. The residual heat removal ventilation system for spent fuel dry storage facility of nuclear power plant of claim 1, characterized in that a temperature measuring room and a transport room are disposed between the spent fuel storeroom and the spent fuel building outer wall; the temperature measuring room is located under the ground, and the transport room is located above the ground; several through-wall storage canister temperature measuring apparatuses are disposed on internal wall of the temperature measuring room and the transport room.
	claims	1. A method for identifying a presence of a substance in an unknown sample, comprising:photodissociating the sample into one or more portions including a NO molecule, the NO molecule having an electron in a first-vibrational excited state of an electronic ground state;employing laser-induced fluorescence to induce fluorescence of the NO molecule; anddetecting the florescence of the NO molecule to thereby identify the presence of the substance. 2. The method of claim 1, wherein the step of photodissociating comprises photodissociating the sample into a plurality of NO molecules each NO molecule having an electron in a first-vibrational excited state and a distinct rotational state, and further wherein the step of employing laser-induced fluorescence includes inducing fluorescence of the NO molecules in distinct rotational states. 3. The method of claim 1, wherein the step of detecting comprises distinguishing the substance from a presence of at least one material containing a NO portion. 4. The method of claim 3, wherein at least one material containing a NO portion is at least one of an atmospheric NO-containing compound, an inorganic NO-containing compound, an inorganic nitrate, and fertilizer. 5. The method of claim 1, wherein the substance comprises a portion of an explosive compound. 6. The method of claim 5, wherein the explosive compound comprises at least one of 2,6-dinitrotoluene, 2,4,6 trinitrotoluene, pentaerythritol tetranitrate, hexahydro-1,3,5-trinitro-1,3,5-triazine, and cyclotrimethylenetrinitramine. 7. The method of claim 1, wherein the step of photodissociating the sample comprises photodissociating a solid material. 8. The method of claim 1, wherein the step of photodissociating the sample comprises exposing the sample to a wavelength of light that differs from a wavelength of light used for employing laser-induced fluorescence. 9. The method of claim 1, wherein the steps of photodissociating and employing laser-induced fluorescence are performed using light energy from a single source. 10. The method of claim 1, wherein the step of employing laser-induced fluorescence comprises exciting the NO molecule using light having a wavelength of about 236.2 nm. 11. The method of claim 1, wherein the method is employed in a stand-off mode. 12. The method of claim 11, wherein the step of detecting the fluorescence comprises collecting the fluorescence using a detector positioned at least about 50 cm from the sample. 13. The method of claim 1, wherein the substance comprises urea nitrate. 14. The method of claim 1, wherein the step of detecting comprises detecting fluorescence at a wavelength of about 226 nm. 15. The method of claim 1, wherein the method is carried out in ambient conditions. 16. A method for identifying a presence of a substance in an unknown solid sample, comprising:photodissociating the solid sample into one or more portions including a NO molecule;employing laser-induced fluorescence to induce fluorescence of the NO molecule; anddetecting the florescence of the NO molecule to thereby identify the presence of the substance. 17. The method of claim 16, wherein the NO molecule has an electron in a first-vibrational excited state of an electronic ground state.
063339579	summary	BACKGROUND OF THE INVENTION The present invention relates to nuclear fuel bundles of a type having water rods serving as supports between a lower tie plate and an upper tie plate and particularly relates to a tool kit for fabricating, inspecting and handling the nuclear fuel bundle. In a recent nuclear fuel bundle design by the assignee of the present invention, mechanical support for the fuel bundle in an unchanneled condition is provided by a pair of central water rods connected to the lower tie plate rather than by conventional tie rods. The water rods are typically threaded and screwed into the lower tie plate and tie bars are fixed at the upper ends of the water rods for attachment to a lifting tool. The upper ends of the tie bars terminate in heads having one or more flats designed for proper orientation and fit into similarly machined complementary openings in the lift tool. The tie bar heads have a D-flat configuration, although other non-circular cross-sectional configurations can be utilized. Preferably, at least one flat is used. BRIEF SUMMARY OF THE INVENTION In this new fuel bundle, the two water rods lie in close proximity to one another in the fuel bundle. As a consequence, their relative lengths from the face of the lower tie plate to the underside of the tie rod ends are required to meet tight tolerances. For example, the length tolerance for the pair of water rods in a particular bundle preferably lies within 0.020 inches of one another to prevent undesirable tilting of the bundle during vertical handling or transport. The water rod assemblies are adjustable in length. This is accomplished by providing for relative axial movement of the water rod and the tie bar which, when adjusted, provide a water rod assembly of fixed predetermined length. To adjust the length, however, it is necessary to first rotate the water rod assembly to engage the threaded lower end plug of the water rod into the lower tie plate and then thread the tie bar into the water rod. The adjustment is further complicated by the requirement for at least one tabbed water rod assembly. That is, at least one water rod assembly requires a plurality of generally radially projecting tabs at axially spaced positions therealong for engagement with the spacers of the fuel bundle whereby the tabs prevent the spacers from disengaging from the water rod assemblies and sliding up or down the assemblies. These tabs lie only on one side of the water rod assembly and must lie in a predetermined orientation in final assembly. This requirement enables adjustment of the water rod in only one turn increments. Thus, the first adjustment to the overall length of the water rod assembly can be accomplished by a rotation of the water rod within a single rotation so that the water rod is one turn or less from bottoming out on the lower tie plate. The secondary adjustment of the tie bar is also restricted to a minimum of a single rotation because in final position, the D-flat on the upper end of the tie bar must be in a designated fixed rotational orientation. Additionally, the threads at the end plugs of the water rods and the lower ends of the tie bars are of different pitch. Consequently, by mutual adjustment of the water rod and tie bar, the tabbed water rod length can be set within a specified range, for example, 0.040 inches. More particularly, the adjustment of the orientation of the tie rod and water rod relative to one another involves first rotating a threaded washer at the junction of the tie rod and water rod and moving a lower axially tabbed lock washer from locking relation with an end plug at the upper end of the water rod. Thus, by releasing the threaded washer and rotating it upwardly along the tie bar, the tie bar can be screwed farther into the upper water rod end plug by rotating it at least one full turn or multiples thereof before the lower lock washer is reseated in the water rod upper end plug and the threaded washer is tightened against it. Likewise, if the tie rod is to be rotated out of the upper water rod end plug to lengthen the water rod assembly, the threaded washer is threaded down on the tie bar. To accomplish this, the threaded washer must first be raised to permit the lower lock washer's axial tab to disengage from the upper water rod end plug. The adjustment in the lengths of the water rod assemblies requires with respect to the tabbed water rod setting the length within a predetermined range, i.e., .+-.0.040 inches, and within one turn of bottoming out on the lower tie plate. It is also necessary to properly orient the water rod tabs and the tie bar D-flats. With respect to setting the length of the untabbed water rod, it is necessary to adjust its length within the predetermined tolerance, i.e., .+-.0.040 inches and also to set its length within a predetermined tolerance, e.g., .+-.0.020 inches of the tabbed water rod assembly length. It is also necessary to properly orient the tie bar D-flat. Special tools and gauges are therefore necessary to accomplish these actions and in accordance with a preferred embodiment of the present invention are provided as part of a tool kit. The tools and gauges of the tool kit include a socket, a flat orientator wrench, a threaded washer wrench, orientator gauges, orientator cap and a tie bar simulator. The first five tools and gauges are used in setting and maintaining the lengths and orientation of the water rods. The last device, i.e., the tie bar simulator is used in checking the alignment of the D-flat holes in the lift bars and lift tools. The tools and gauges of the tool kit overcome significant problems in fabricating and handling the new fuel bundle design. For example, problems associated with locking and unlocking the threaded washer especially in a bundle with some of the fuel rods installed, rotation of the tie bars to adjust their lengths and orientation of the water rod tabs and the tie bar D-flats in a bundle where some or all of the fuel rods installed are overcome by the present invention. Additionally, the tool kit provides tools and gauges for checking the alignment of the two tie bar D-flat ends over maximum, minimum and nominal dimensions, maintaining the relative orientation of the tie rod D-flats during handling and shipping and checking the orientation of the machined D-flat holes in the lift bars and tools prior to shipment. In a preferred embodiment according to the present invention, there is provided a tool kit for adjusting the length of a water rod assembly in a nuclear fuel bundle having a lower end plug for threaded engagement in a lower tie plate, the water rod assembly including a pair of water rods, a pair of tie bars having flats at upper ends thereof and releasable locking subassemblies for respectively securing the tie bars and water rods to one another, forming joints therebetween, comprising first and second gauges each having a gauge body with first and second openings generally complementary in shape to the respective upper ends of the tie bars having the flats, the openings having parallel axes, margins of each gauge body defining the openings being offset from one another in the direction of the axes, the openings of the second gauge being reduced relative to the openings in the first gauge body whereby the angular orientation of the water rods about their respective axes is adjustable upon application of the first and second gauges to the tie rods. In a further preferred embodiment of the present invention, there is provided a tool kit for adjusting the length of a water rod assembly in a nuclear fuel bundle having a lower end plug for threaded engagement in a lower tie plate, the water rod assembly including a pair of water rods, a pair of tie bars having flats at upper ends thereof and releasable locking subassemblies for respectively securing the tie bars and water rods to one another forming joints therebetween. The tool kit comprises a pair of wrenches having wrench heads including openings respectively complementary in shape to the flats at the upper ends of the tie bars and reference points on the wrench heads for alignment with one another upon engagement of the tie bar upper end in the wrench head openings to align the water rod assemblies with one another rotationally about their respective axes; and at least one gauge having a gauge body with first and second openings generally complementary in shape to the respective upper ends of the tie bars having the flats, the openings having parallel axes, margins of the gauge body defining the openings being offset from one another in the direction of the axes whereby angular orientation of the water rods about their respective axes is adjusted.
	claims	1. A computer-implemented method for processing a composite signal, the computer-implemented method comprising:forming, by a first block of a simulatable block diagram model, a first composite signal from an ordered set of a first signal and a second signal, the first signal being associated with one or more attributes and the second signal being associated with one or more attributes, the first signal being associated with at least one attribute that is different than the attributes associated with the second signal, the first composite signal preserving the attributes associated with the first signal and the second signal;inputting the first composite signal into a second block of the simulatable block diagram model; andprocessing the first composite signal in the second block, the processing including:performing an operation on the first signal,performing the operation on the second signal, andoutputting one or more output signals from the second block, a value of the one or more output signals being based on a result of performing the operation on the first signal and the second signal. 2. The method of claim 1, wherein the attributes of the first signal or the second signal include one or more of a name, data type, numeric type and dimensionality associated with the first signal or the second signal. 3. The method of claim 1, wherein:the one or more output signals have dimensionality,the first signal and the second signal of the first composite signal have dimensionality, andthe dimensionality of the one or more output signals is different than the dimensionality of the first signal or the second signal of the first composite signal. 4. The method of claim 1, wherein one of the first signal or the second signal of the first composite signal is a second composite signal. 5. The method of claim 4, wherein there is no circular relationship between the first composite signal and the second composite signal. 6. The method of claim 1, wherein the first composite signal includes a third signal, the third signal being identical to the first signal or the second signal. 7. The method of claim 1, wherein the first block is a multiplexer block having two or more inputs and one or more outputs. 8. The method of claim 7, wherein the multiplexer block generates a linked list having a plurality of nodes, a node of the linked list representing the first signal or the second signal of the first composite signal. 9. The method of claim 7, wherein at least one of the two or more inputs of the multiplexer block is a second composite signal. 10. The method of claim 1, wherein the first composite signal of the simulatable block diagram model is displayed on a display device using visual cues such that the first composite signal is distinguishable from a non-composite signal. 11. The method of claim 1, wherein the simulatable block diagram model includes a demultiplexer block having one or more input ports and one or more output ports, the method further comprising:receiving the first composite signal at the one or more input ports of the demultiplexor block;extracting at least one of the first signal and the second signal from the first composite signal; andoutputting at least one of the extracted first signal and the extracted second signal from the demultiplexor block at one or more output ports of the demultiplexor block. 12. The method of claim 1, wherein the simulatable block diagram model includes a viewer block, and wherein the method further comprises:receiving the first composite signal at an input port of the viewer block; andproviding, on a graphical user interface associated with the viewer block, a graphical visualization of a hierarchical organization of the first composite signal. 13. The method of claim 12, wherein the graphical visualization includes a list that lists the first signal and the second signal of the first composite signal. 14. A storage medium storing computer executable instructions for processing a composite signal, the storage medium storing one or more instructions for:forming, by a first block of a simulatable block diagram model, a composite signal from an ordered set of a first signal and a second signal, the first signal being associated with one or more attributes and the second signal being associated with one or more attributes, the first signal being associated with at least one attribute that is different than the attributes associated with the second signal, the composite signal preserving the attributes associated with the first signal and the second signal;inputting the composite signal into a second block of the simulatable block diagram model; andprocessing the composite signal in the second block, the processing including:performing an operation on the first signal,performing the operation on the second signal, andoutputting one or more output signals from the second block, a value of the one or more output signals being based on a result of performing the operation on the first signal and the second signal. 15. The storage medium of claim 14, wherein the first block is a multiplexer block that generates a linked list having a plurality of nodes, a node of the linked list representing the first signal or the second signal of the composite signal. 16. The storage medium of claim 14, wherein the composite signal of the simulatable block diagram model is displayed on a display device using visual cues such that the composite signal is distinguishable from a non-composite signal. 17. The storage medium of claim 14, wherein the simulatable block diagram model includes a demultiplexer block having one or more input ports and one or more output ports, the method further comprising:receiving the composite signal at the one or more input ports of the demultiplexor block;extracting at least one of the first signal and the second signal from the composite signal; andoutputting at least one of the extracted first signal and the extracted second signal from the demultiplexor block at one or more output ports of the demultiplexor block. 18. The storage medium of claim 14, wherein the simulatable block diagram model further comprises a viewer block coupled to the composite signal, the storage medium further storing one or more instructions for:receiving the composite signal at an input port of the viewer block; andproviding, on a graphical user interface associated with the viewer block, a graphical visualization of a hierarchical organization of the composite signal. 19. The storage medium of claim 18, wherein the graphical visualization includes a list that lists the first signal and the second signal of the composite signal. 20. A system comprising:a processor for:forming, by a first block of a simulatable block diagram model, a composite signal from an ordered set of a first signal and a second signal, the first signal being associated with one or more attributes and the second signal being associated with one or more attributes, the first signal being associated with at least one attribute that is different than the attributes associated with the second signal, the composite signal preserving the attributes associated with the first signal and the second signal;inputting the composite signal into a second block of the simulatable block diagram model; andprocessing the composite signal in the second block, the processing including:performing an operation on the first signal,performing the operation on the second signal, andoutputting one or more output signals from the second block, a value of the one or more output signals being based on a result of performing the operation on the first signal and the second signal.
	abstract	An X-ray optical transmission grating of a focus-detector arrangement of an X-ray apparatus is disclosed, for generating projective or tomographic phase contrast recordings of a subject. In at least one embodiment, the grating includes a multiplicity of grating bars and grating gaps arranged periodically on at least one surface of at least one wafer, wherein the X-ray optical transmission grating includes at least two sub-gratings arranged in direct succession in the beam direction.
	summary
	abstract	A system for measuring flow rate within a volume includes one or more transmission devices that transmit one or more signals through fluid contained within the volume. The volume may be bounded, at least in part, by an outer structure and by an object at least partially contained within the outer structure. A transmission device located at a first location of the outer structure transmits a first signal to a second location of the outer structure. A second signal is transmitted through the fluid from the second location to a third location of the outer structure. The flow rate of the fluid within the volume may be determined based, at least in part, on the time of flight of both the first signal and the second signal.
045267456	claims	1. A fuel assembly for a boiling water reactor, the assembly having a vertical centerline and comprising: a lower lattice element; a plurality of vertical fuel rods supported by said lattice element; a fuel box surrounding said fuel rods, said fuel box having an exterior surface which in use of the fuel assembly faces the space which is within a boiling water reactor but is outside the fuel assembly; a sleeve-like base with a downwardly facing inlet opening for reactor coolant to be flowed over the surfaces of said fuel rods, said base supporting said fuel box and said lattice element and also having a wall with an exterior surface which in use of the fuel assembly faces the space which is within a boiling water reactor but is outside the fuel assembly; at least one vertical water channel located within said fuel box for a flow of water along but separated from said fuel rods, said vertical channel having a lower end; at least one channel positioned above said downward-facing inlet opening, said channel being extended radially relative to said center line and opening at the side surface of the fuel assembly so that in use of the fuel assembly said radially extended channel opens to the space which is within the boiling water but is outside the fuel assembly; means for hydraulically connecting said at least one radially extended channel to said lower end of said water channel to provide flow from the outside of said fuel assembly, through said radially extended channel and into said water channel, said means for hydraulically connecting comprising a hollow body, said hollow body receiving said flow only from said at least one radially extended channel, said at least one water channel receiving said flow only from said hollow body; and at least one through-hole in said wall of said base, said through-hole opening outward through the side surface of the fuel assembly and being positioned above said inlet opening but below said at least one radially extended channel, said through-hole being hydraulically connected to said inlet opening of said base, whereby in use of the fuel assembly coolant flows to said at least one water channel via said inlet opening, said at least one through-hole, the space which is within a boiling water reactor but is outside the fuel assembly, said at least one radially extended channel, and said menas for hydraulically connecting. 2. A fuel assembly according to claim 1, wherein there are a plurality of vertical water channels comprised of a plurality of vertical water tubes arranged in two rows intersecting at said vertical center line, said water tubes having lower ends; and said hollow body is a cruciform body supported by said base, said lower ends of said water tubes being supported by said hollow cruciform body and said water tubes being hydraulically connected to the interior of said hollow cruciform body, said at least one radially extended channel extending from said hollow cruciform body.
	claims	1. A jet pump beam made of improved heat-treated nickel base alloy, produced by preparing a precipitation-strengthened nickel base alloy material having a component composition containing by mass %, Ni: 50.0% to 55.0%, Cr: 17.0% to 21.0%, Nb+Ta: 4.75% to 5.50%, Mo: 2.8% to 3.3%, Ti: 0.65% to 1.15%, Al: 0.2% to 0.8%, C: 0.08% or less, Mn: 0.35% or less, Si: 0.35% or less, S: 0.015% or less, P: 0.03% or less, Cu: 0.30% or less, B: 0.001% or less, and Co: 0.05% or less, and Fe and inevitable impurities constituting a remaining part,the nickel base alloy material having been subjected to solution heat treatment at a temperature of 1010° C. to 1090° C., andthe nickel base alloy material having a γ″ phase (Ni3Nb) precipitate formed by subjecting the nickel base alloy material to age-hardening heat treatment at a temperature of 694° C. to 714° C. for 5 to 7 hours after the solution heat treatment. 2. The jet pump beam according to claim 1, produced from an improved heat-treated nickel base alloy material obtained by forming the nickel base alloy material into a product shape by machining or product processing after the solution heat treatment, and subsequently subjecting the material to the age-hardening heat treatment. 3. A method for producing a jet pump beam made of improved heat-treated nickel base alloy, comprising the steps of: preparing a precipitation-strengthened nickel base alloy material having a component composition containing by mass %, Ni: 50.0% to 55.0%, Cr: 17.0% to 21.0%, Nb+Ta: 4.75% to 5.50%, Mo: 2.8% to 3.3%, Ti: 0.65% to 1.15%, Al: 0.2% to 0.8%, C: 0.08% or less, Mn: 0.35% or less, Si: 0.35% or less, S: 0.015% or less, P: 0.03% or less, Cu: 0.30% or less, B: 0.001% or less, and Co: 0.05% or less, and Fe and inevitable impurities constituting a remaining part;forming the nickel base alloy material into a product shape by machining or cold working after subjecting the nickel base alloy material to solution heat treatment at a temperature of 1010° C. to 1090° C.; andsubjecting the nickel base alloy material formed into the product shape to age-hardening heat treatment at a temperature of 694° C. to 714° C. for 5 to 7 hours, thereby to precipitate a γ″ phase (Ni3Nb) in the nickel base alloy material. 4. The method for producing a jet pump beam according to claim 3, wherein the jet pump beam made of improved heat-treated nickel base alloy is produced by roughly forming the nickel base alloy material into the product shape by die forging after melting the nickel base alloy material, subjecting the roughly formed nickel base alloy material to the solution heat treatment at a temperature of 1010° C. to 1090° C., finishing the nickel base alloy material by the machining or the cold working, and subjecting the finished nickel base alloy material to the age-hardening heat treatment at a temperature of 694° C. to 714° C. for 5 to 7 hours, thereby to precipitate a γ″ phase (Ni3Nb) in the nickel base alloy material. 5. A jet pump beam used in a jet pump for forcibly circulating high-temperature and high-pressure water within a reactor pressure vessel of a boiling water reactor, the jet pump beam comprising:a body made of an elastic material;a vertical threaded hole provided in a center portion of the body, into which a head bolt is screwed; anda body top portion support surface for supporting a head bolt fixing device,wherein the body is made of precipitation-strengthened nickel base alloy having a component composition containing by mass %, Ni: 50.0% to 55.0%, Cr: 17.0% to 21.0%, Nb+Ta: 4.75% to 5.50%, Mo: 2.8% to 3.3%, Ti: 0.65% to 1.15%, Al: 0.2% to 0.8%, C: 0.08% or less, Mn: 0.35% or less, Si: 0.35% or less, S: 0.015% or less, P: 0.03% or less, Cu: 0.30% or less, B: 0.001% or less, and Co: 0.05% or less, and Fe and inevitable impurities constituting a remaining part, and the nickel base alloy material is subjected to solution heat treatment at a temperature of 1010° C. to 1090° C., and subjected to age-hardening heat treatment at a temperature of 694° C. to 714° C. for 5 to 7 hours after the solution heat treatment, to thereby precipitate γ″ phase (Ni3Nb) in the nickel base alloy material and provide a jet pump beam made of an improved heat-treated nickel base alloy material.
051134233	claims	1. A method for improving the coherence of a pulse of X-ray radiation comprising the steps of: longitudinally splitting said pulse into a plurality of pulses; delaying at least one of said pulses with respect to at least one other of said pulses; and combining said pulses in serial fashion to form a output beam longer than said pulse. 2. The method of claim 1, wherein delaying is accomplished by using static reflectors. 3. The method of claim 1, wherein delaying is accomplished by using a grating in conjunction with one or more static reflectors. 4. The method of claim 1, wherein combining is accomplished by using a rotating mirror.
	abstract	An apparatus with which areas near surfaces in a water environment that are contaminated by radioisotopes are decontaminated by non-thermal laser peeling without suffering re-melting, re-diffusing and re-contaminating. The apparatus includes a piping structure with which a substance to be irradiated that has been deposited on outer and inner surfaces of a nuclear reactor pressure vessel, and a nuclear reactor container tank, and the internal nuclear reactor structures all having been contaminated with radioisotopes, can be removed in the water environment. The piping structure, to secure a region in a water environment that is gas pressurized to discharge the water and filled with the gas to not interfere with laser irradiation, has a semi-hermetically closed, incomplete water seal that is half-open with a siphon provided downward, a mechanical structure that withstands water pressure in a radial direction, and an extendable bellows-like tube to enable the piping structure to tilt.
053176111	claims	1. A fuel element for a nuclear thermal engine having a fueled truncated conical shell with a base characterized by an annular unfueled lip at said base having radial passages therethrough, wherein said annular lip is integral with said truncated conical shell, and said annular lip extends axially above said base. 2. The fuel element of claim 1 wherein said radial passages are defined by walls in said annular lip which diverge radially inwardly. 3. The fuel element of claim 1 wherein said annual lip is porous with pores forming said radial passages. 4. A fuel element for a nuclear thermal engine having a fueled truncated conical shell with a base characterized by an annular unfueled lip at said base having radial passages therethrough and a plurality of angularly spaced ribs extending radially along a truncated conical surface of said shell forming channels therebetween. 5. The fuel element of claim 4 wherein said truncated conical surface is an outer surface of said truncated conical shell. 6. The fuel element of claim 5 wherein said truncated conical shell has a base and said fuel element includes an annular lip extending radially outward from said base and with radial passages therethrough. 7. A fuel element for a nuclear thermal engine having a fueled truncated conical shell characterized by a plurality of angularly spaced unfueled ribs extending radially along at least one truncated conical surface of said shell forming channels therebetween. 8. A stack of fuel elements for a nuclear thermal engine having a plurality of fuel elements, each fuel element having a truncated conical shell with a base, characterized in that an annular lip with radial passages therethrough extends radially outward from said base, said fuel elements being stacked with said lips positioned on top of one another to form said stack with frusto-conical flow passages between adjacent elements and with said radial passages communicating with said frusto-conical flow passages. 9. The stack of fuel elements of claim 8 wherein each of said fuel elements has a plurality of angularly spaced ribs extending radially along a truncated conical surface of said frusto-conical shell dividing said truncated conical flow passages into flow channels. 10. The stack of fuel elements of claim 9 wherein said truncated conical surface is an outer surface of each of said truncated conical shells. 11. The stack of fuel elements of claim 8 wherein said annular lips of said fuel elements comprise rigid rings having radial passages therethrough and compliant annular supports extending radially inward from said rigid rings and supporting said truncated conical shells at said bases thereof. 12. The stack of fuel elements of claim 11 wherein said compliant annular support defines an annular groove in which said base of the truncated conical shell seats. 13. The stack of fuel elements of claim 13 wherein said compliant annular supports have annular outer sections which are clamped between adjacent rigid rings and annular cantilevered inner sections supporting said truncated conical shells at said bases. 14. The stack of fuel elements of claim 13 wherein said cantilevered inner sections of said compliant supports extend toward said truncated conical shells generally transverse to outer surfaces of said shells and terminate in annular grooves in which said bases of said truncated conical shells are supported. 15. The stack of fuel elements of claim 13 wherein said annular cantilevered inner sections of said compliant supports extend parallel t truncated conical surfaces of said truncated conical shells with adjacent inner sections forming annular slots into which the bases of the truncated conical shells extend, and including spring means in said annular slots resiliently supporting and centering said truncated conical shells. 16. The stack of fuel elements of claim 8 wherein said annular lips comprise rings with radial passages therethrough and having internal annular grooves in which the truncated conical shells seat. 17. The stack of fuel elements of claim 16 wherein the bases of said truncated conical shells have generally radially outwardly extending flanges which seat in said grooves in said rings. 18. The stack of fuel elements of claim 8 wherein said lips are unfueled and integral with the bases of said truncated conical shells.
046506322	claims	1. In a device for magnetically confining a plasma driven by a plasma current and contained within a toroidal vacuum chamber, the device having an inner toroidal limiter on an inside wall of said vacuum chamber, an arrangement for the rapid prediction and control in real time of a major plasma disruption, the arrangement including: scanning means sensitive to infrared radiation emanating from within said vacuum chamber, said infrared radiation indicating the temperature along a vertical profile of said inner toroidal limiter, said scanning means arranged to observe said infrared radiation and to produce in response thereto an electrical scanning output signal representative of a time scan of temperature along said vertical profile; detection means for analyzing said scanning output signal to detect a first peaked temperature excursion occuring along said profile of said inner toroidal limiter, and to produce a detection output signal in response thereto, said detection output signal indicating a real time prediction of a subsequent major plasma disruption; and plasma current reduction means for reducing said plasma current driving said plasma, in response to said detection output signal and in anticipation of a subsequent major plasma disruption. 2. The arrangement of claim 1 wherein said detection means comprises means for comparing a first scanning output signal taken at an outer edge of said limiter, and means for detecting a rapid increase in said first scanning output signal relative to said second scanning output signal and for producing said detection output signal in response to said rapid increase. 3. The arrangement of claim 1 wherein said detection means comprises means for comparing a first time rate of change of a first scanning output signal taken at a midplane of said limiter to a second time rate of change of a second scanning output signal taken at an outer edge of said limiter, and means for detecting a rapid increase in said first time rate of change relative to said second time rate of change, and for producing said detection output signal in response to said rapid increase. 4. The arrangement of claim 2 wherein said scanning means comprises an array of infrared photodetectors. 5. The arrangement of claim 2 wherein said scanning means comprises an infrared camera. 6. The arrangement of claim 3 wherein said scanning means comprises an array of infrared photodetectors. 7. The arrangement of claim 3 wherein said scanning means comprises an infrared camera. 8. The arrangement of claim 7 wherein said camera operates in the wavelength range of 3-5 micrometers. 9. The arrangement of claim 7 wherein said camera operates in the wavelength range of 8 to 12 micrometers.
047956054	summary	BACKGROUND OF THE INVENTION The present invention relates to a nuclear fusion apparatus, and in particular, to the structure of a central support member for supporting a plurality of poloidal coils in a tokamak type nuclear fusion apparatus. In a conventional tokamak type nuclear fusion apparatus such as the one disclosed in Japanese Laid-Open Pat. No. 54,3695, as shown in FIG. 1, a plurality of toroidal coils 2 are radially disposed around a central support stay 1, and a plurality of poloidal coils 3 are circumferentially disposed around the central support stay 1 and spaced from each other along the axial direction thereof, i.e., direction Z in FIG. 1. Electrically insulating materials 5 are radially disposed in the central support stay 1 along the axis thereof to interrupt undesirable electric current in the circumferential direction of the central support stay 1 as described later. In the nuclear fusion apparatus described above, an electric current is made to flow through the toroidal coils 2 to confine plasma 4 within the bounds of the coils 2 and an electric current is also made to flow through the poloidal coils 3 to stabilize the position of the plasma 4. At this time, in the toroidal coils 2, an electromagnetic force caused by their respective electric currents is generated as a centripetal force in a radial direction, i.e., direction R in FIG. 1. In the central support stay 1, a voltage in the circumferential direction i.e., direction .theta. is generated by the electric current flowing through the poloidal coils 3. Accordingly, it is necessary for the central support stay 1 to withstand the centripetal force of the toroidal coils and it is necessary to interrupt the undesirable electric current through the central support stay 1 in the circumferential direction thereof. In the conventional apparatus, the central support stay 1 has a cylindrical shape to withstand the centripetal force and is divided into a plurality of segments in the circumferential direction thereof, with the electrically insulating materials 5 disposed between the segmented portions to interrupt undesirable electric current in the circumferential direction. In the conventional nuclear fusion apparatus mentioned above, it is difficult to design a central support stay 1 as a structure being circumferentially divided yet still highly rigid in the radial direction. Furthermore, in the conventional apparatus, in order to supply the electric current to the poloidal coils 3, it is necessary to dispose holes on the side surface of the central support stay 1 for passing lead wires therethrough and to carry out brazing etc. in the assembly of the central support stay 1. Furthermore, in the conventional nuclear fusion apparatus, it is not easy to dispose the poloidal coils in predetermined positions. To overcome the problems mentioned above, an object of the present invention is to provide a nuclear fusion apparatus in which a central support stay can sufficiently withstand the centripetal force of the toroidal coils and has a high electrical resistance in the circumferential direction of the central support stay, in which the structure of the central support stay is simple and yet facilitates operations for disposing lead wires in the central support stay, etc. Another object of the present invention is to provide a nuclear fusion apparatus in which the poloidal coils are easily positioned in the circumferential direction of the central support stay. Accordingly, the present invention provides a nuclear fusion apparatus comprising a central support stay assembly, two or more poloidal coils disposed in the circumferential direction of the central support stay assembly, and two or more toroidal coils radially disposed around and adjacent to the central support stay assembly. The central support stay assembly includes a central support stay portion and a radial arrangement for supporting the poloidal coils and for withstanding the centripetal force of the toroidal coils. The radial arrangement includes two or more radial portions extending radially from the central support stay portion and spaced circumferentially from each other at predetermined angles. Each of the radial portions has a radial member radially extending from the central support stay portion and a force transmission member disposed at the radial outer end of the radial member for transmitting forces from at least two toroidal coils to each radial member. The invention also provides a nuclear fusion apparatus comprising a central support stay assembly having a central support stay portion and two or more radial portions. The radial portions extend radially from the central support stay portion and are circumferentially spaced from each other at predetermined angles. Each of the radial portions includes a movable radial member which is connected to the central support stay portion and is movable in the radial direction of the central support stay assembly. The nuclear fusion apparatus also comprises two or more poloidal coils which are disposed in the circumferential direction of the central support stay assembly and which are supported by the radial portions. A groove for receiving at least one of the poloidal coils is formed at the radial outer end of each radial member. The nuclear fusion apparatus further comprises a plurality of toroidal coils which are radially disposed around the central support stay assembly and which are adjacent to the radial outer ends of the radial portions. The present invention further provides a nuclear fusion apparatus comprising a central support stay assembly having a central support stay portion and two or more of radial portions. The radial portions extend radially from the central support stay portion and are circumferentially spaced from each other at predetermined angles. The central support stay assembly is divided into two or more of support stay members in the axial direction of the central support stay assembly. Consequently, the central support stay portion and each radial portion are respectively divided into central support stay segments and radial segments. Each of the support stay members has a hole extending therethrough and the central support stay assembly further includes a bolt inserted into the hole of each of the support stay members and a nut screwed onto the bolt. The nuclear fusion apparatus also comprises two or more poloidal coils which are disposed in the circumferential direction of the central support stay assembly and are supported by the radial portions. Each of the radial segments includes an outer end surface and a step portion for receiving at least one of the poloidal coils. The step portion has a flat portion which is perpendicular to the axis of each support stay member and which is adjacent to the radial outer end surface. The nuclear fusion apparatus further comprises two or more toroidal coils which are radially disposed around the central support stay assembly and which are adjacent to the radial outer ends of the radial portions. Each of the radial segments further includes a portion for withstanding the centripetal force from each of the toroidal coils at the radial outer end of the radial segment.
	abstract	Methods and systems are provided for continuous-flow production of radioisotopes with high specific activity. Radioisotopes with high specific activity produced according to the methods described are also provided. The methods can include causing a liquid capture matrix to contact a target containing a target nuclide; irradiating the target with radiation, ionizing radiation, particles, or a combination thereof to produce the radionuclides that are ejected from the target and into the capture matrix; and causing the liquid capture matrix containing the radionuclides to flow from the target to recover the capture matrix containing the radionuclides with high specific activity. The methods are suitable for the production of a variety of radionuclides. For example, in some aspects the target nuclide is 237Np, and the radionuclide is 238Np that decays to produce 238Pu. In other aspects, the target nuclide is 98Mo, and the radionuclide is Mo that decays to produce 99mTc.
	summary
	summary
	description	The present application claims the benefit of U.S. Provisional Application No. 60/850,733, filed on Oct. 11, 2006, the entirety of which is hereby incorporated by reference. The present invention relates generally to the field of transporting and/or preparing high level radioactive waste (“HLW”) for dry storage, and specifically to apparatus and methods for transporting, removing and/or preparing HLW for dry storage from a fuel pool/pond. In the operation of nuclear reactors, the nuclear energy source is in the form of hollow zircaloy tubes filled with enriched uranium, typically referred to as fuel assemblies. When the energy in the fuel assembly has been depleted to a certain level, the assembly is removed from the nuclear reactor. At this time, fuel assemblies, also known as spent nuclear fuel, emit both considerable heat and extremely dangerous neutron and gamma photons (i.e., neutron and gamma radiation). Thus, great caution must be taken when the fuel assemblies are handled, transported, packaged and stored. After the depleted fuel assemblies are removed from the reactor, they are placed in a canister. Because water is an excellent radiation absorber, the canisters are typically submerged under water in a pool. The pool water also serves to cool the spent fuel assemblies. When fully loaded with spent nuclear fuel, a canister weighs approximately 45 tons. The canisters must then be removed from the pool because it is ideal to store spent nuclear fuel in a dry state. The canister alone, however, is not sufficient to provide adequate gamma or neutron radiation shielding. Therefore, apparatus that provide additional radiation shielding are required during transport, preparation and subsequent dry storage. The additional shielding is achieved by placing the canisters within large cylindrical containers called casks. Casks are typically designed to shield the environment from the dangerous radiation in two ways. First, shielding of gamma radiation requires large amounts of mass. Gamma rays are best absorbed by materials with a high atomic number and a high density, such as concrete, lead, and steel. The greater the density and thickness of the blocking material, the better the absorption/shielding of the gamma radiation. Second, shielding of neutron radiation requires a large mass of hydrogen-rich material. One such material is water, which can be further combined with boron for a more efficient absorption of neutron radiation. There are generally two types of casks, transfer casks and storage casks. Transfer casks are used to transport spent nuclear fuel within the nuclear facility. Storage casks are used for the long term dry state storage. Guided by the shielding principles discussed above, storage casks are designed to be large, heavy structures made of steel, lead, concrete and an environmentally suitable hydrogenous material. However, because storage casks are not typically moved, the primary focus in designing a storage cask is to provide adequate radiation shielding for the long-term storage of spent nuclear fuel. Size and weight are at best secondary considerations. As a result, the weight and size of storage casks often cause problems associated with lifting and handling. Typically, storage casks weigh approximately 150 tons and have a height greater than 15 ft. A common problem is that storage casks cannot be lifted by the cranes in typical nuclear power plants because their weight exceeds the rated capacity of the crane. Another common problem is that storage casks are too large to be placed in storage pools. Thus, in order to store spent nuclear fuel in a storage cask, a loaded canister must be removed from the storage pool, prepared in a decontamination station, and transported to the storage cask. Additional radiation shielding is required throughout all stages of the transport and preparation procedures. Removal from the storage pool and transport of the loaded canister to the storage cask is facilitated by a transfer cask. Generally, an empty canister is first placed within an open transfer cask. The transfer cask and empty canister are then submerged in the storage pool. After the fuel assemblies are removed from the nuclear reactor they are placed into the pool, within the submerged canister. While underwater, the loaded canister is fitted with a lid, thereby enclosing water and the fuel assemblies within the canister. The transfer cask, which contains the loaded canister, is then removed from the pool by a crane, or other similar piece of equipment. After being removed from the pool, the transfer cask is placed on a decontamination station to prepare the spent nuclear fuel for long-term storage in the dry state. In the decontamination station the bulk water is pumped out of the canister, thereby reducing the combined weight of the canister and transfer cask. This is called dewatering. Once dewatered, the spent nuclear fuel is further dried to an acceptable level through an appropriate drying method. Once adequately dry, the canister is back-filled with an inert gas, such as helium. The canister is then sealed and a radiation absorbing lid is secured to the transfer cask body. The transfer cask and canister are then transported to the storage cask where the canister will be transferred to the storage cask. In some instances, the transfer cask itself may be used as the storage cask. Transfer casks are designed to be lighter and smaller than storage casks because a transfer cask must be lifted and handled by the plant's crane. A transfer cask must be small enough to fit in a storage pool and light enough so that when it is loaded with a canister of spent nuclear fuel, its weight does not exceed the crane's rated weight limit. Importantly, however, a transfer cask must also perform the vital function of providing adequate radiation shielding for both neutron and gamma radiation emitted by the enclosed spent nuclear fuel. The transfer cask must also be designed to provide adequate heat transfer. Thus, in designing transfer casks and their handling procedures, the desirability of maximizing radiation shielding (which is generally achieved by increasing the mass of the cask's structure) must be balanced against the competing interest of keeping the combined weight of the transfer cask and its payload within the crane's rated weight limit. In order to achieve the necessary gamma and neutron radiation shielding properties, transfer casks are typically constructed of a combination of a gamma absorbing material (e.g., lead, steel, concrete, etc.) and a neutron absorbing material (e.g., water or another material that is rich in hydrogen). The body and lid of the cask, which are generally formed of lead, steel, concrete or a combination thereof, form the cavity in which the spent fuel is to be positioned and function as a containment boundary for all radioactive particulate matter. While the pool water sealed within the canister provides some neutron shielding, this water is eventually drained at the decontamination staging area. Therefore, many transfer casks have either a separate layer of neutron absorbing material or have an annular space filled with water that circumferentially surrounds the cavity of the transfer cask and/or the containment boundary formed by the body. Such annular spaces are typically referred to as water jackets. As stated previously, greater radiation shielding is provided by increased thickness and density of the gamma and neutron absorbing materials. However, increasing the thickness and density of the materials used to make the transfer cask results in a heavier transfer cask. Thus, the extent of radiation shielding is directly proportional to the weight of the transfer cask. The weight of a transfer cask, however, must remain below the rated lifting capacity of the crane. The load handled by the crane includes the weight of the transfer cask and the combined weight of the canister and the fuel assemblies and water (i.e. the transfer cask's payload). A transfer cask must be designed so that the total load does not exceed the rated limit of the crane. Thus, the permissible weight of the transfer cask is the rated lifting capacity of the crane minus the weight of its payload. It is important to note that when the combined weight of the transfer cask and its payload is equal to the rated lifting capacity of the crane, the radiation shielding provided by the transfer cask is at a maximum for that particular payload. This is so because the thickness of the gamma and neutron absorbing materials are at a maximum for that crane and that payload. The weight of the transfer cask's payload varies during the different stages of the transport procedure. The permissible weight of the transfer casks is calculated when the payload is at its maximum. This occurs when the transfer cask is being lifted out of the pool because it contains a loaded canister which is full of about 70 tons of water and the nuclear fuel assemblies. Upon dewatering in the decontamination station, the weight of the transfer cask drops below the rated capacity of the crane and typically remains so throughout the remaining procedures. As such, the radiation shielding provided by the transfer cask is sub-standard throughout the procedure following removal from the storage pool. However, a heavier transfer cask cannot be used throughout the entirety of the transport procedure because the combined weight of the heavier transfer cask and its payload would exceed the rated lifting capacity of the crane during the initial step of lifting the transfer cask from the storage pool. Thus, the maximum amount of radiation shielding is not provided throughout every step of the transfer and dry-storage preparation procedure. While it is possible to transfer the canister of spent nuclear fuel to a heavier transfer cask once the payload is lightened from dewatering, this would take additional time, money, effort, space and equipment. An additional transfer would also increase the amount of radiation exposure to personnel and the risk of a handling accident. A need exists for an apparatus that can provide the maximum amount of shielding throughout all stages of transferring spent nuclear fuel. A need also exists for a method of transferring a canister of spent nuclear fuel from a storage pool that provides the maximum amount of radiation shielding during all stages of the transfer procedure, even when the weight of the transfer cask's load varies. It is an object of the present invention to provide an apparatus that can provide the maximum amount of radiation shielding during all stages of an HLW transfer procedure. Another object of the present invention is to provide an apparatus for transferring HLW, the weight of which can be easily and quickly varied to maximize the amount of radiation shielding for a varied payload without substantially increasing the transfer procedure cycle lime, Yet another object of the present invention is to provide an apparatus for maximizing radiation shielding that can be placed around the transfer cask safely and efficiently subsequent to removal from the storage pool. Still another object of the present invention is to provide a method of transferring HLW that provides the maximum amount of radiation shielding during all stages of the transfer procedure, even when the weight of the payload is varied. Yet another object of the present invention is to provide a method of transferring HLW that provides adequate radiation shielding during all stages of the process even when a low capacity crane is utilized. Still another object of the present invention is to provide a method of transferring HLW that minimizes the weight of the apparatus' payload at the initial step of lifting the apparatus out of a storage pool. It is a further object of the present invention to provide an apparatus that can provide a natural thermosiphon circulation of a neutron absorbing fluid within a jacket for facilitating increased cooling of HLW. A still further object of the present invention is to provide a method of transferring HLW from a submerged state in a fuel pool to a staging area that utilizes the buoyancy of the water in the pool. These and other objects are met by the present invention, which is one aspect can be an apparatus for transporting and/or storing radioactive materials comprising: a gamma radiation absorbing body forming a cavity for receiving radioactive material; a jacket surrounding the body thereby forming a gap between the body and the jacket for holding a neutron absorbing fluid; a baffle positioned in the gap in spaced relation to both the body and the jacket so as to divide the gap into an inner region and an outer region; a passageway at or near a bottom of the gap between the inner region and the outer region that allows the neutron absorbing fluid to flow from the outer region into the inner region; and a passageway at or near a top of the gap between the inner region and the outer region that allows the neutron absorbing fluid to flow from the inner region into the outer region In another embodiment, the invention can be a jacket apparatus for providing neutron radiation shielding to a container holding radioactive materials comprising: an enclosed volume formed by a plurality of surfaces comprising an inner wall and an outer wall; a baffle positioned in the enclosed volume in spaced relation to the inner and outer walls so as to divide the enclosed volume into an inner region and an outer region; at least one passageway at or near a top end of the enclosed volume spatially connecting the inner region and the outer region; and at least one passageway at or near a bottom end of the enclosed volume spatially connecting the inner region and the outer region. In another embodiment, the invention can be a method for transporting and/or storing radioactive materials comprising: providing a container having a cavity, a water jacket surrounding the cavity and forming an annular gap filled with a neutron absorbing fluid, a baffle positioned in the annular gap so as to divide the annular gap into an inner region and an outer region, a lower passageway between the inner region and the outer region, and an upper passageway between the inner region and the outer region; positioning radioactive material having a residual heat load in the cavity; and wherein heat emanating from the radioactive materials warms the neutron absorbing fluid in the inner region so as to cause the neutron absorbing fluid to flow upward in the inner region, the warmed neutron absorbing fluid flowing through the upper passageway and into the outer region where it is cooled, the cooled neutron absorbing fluid flowing downward in the outer region and back into the inner region via the lower passageway, thereby achieving a thermosiphon fluid flow. In yet another aspect, the invention can be an apparatus for providing additional radiation shielding to a container holding radioactive materials comprising: a tubular shell extending from a first end to a second end, the tubular shell constructed of a gamma radiation absorbing material and having an inner surface that forms a cavity; a first opening in the first end of the tubular shell that provides a passageway into the cavity; a second opening in the second end of the tubular shell that provides a passageway into the cavity, the second opening being larger than the first opening; and a plurality of spacers extending from the inner surface of the shell. In still another embodiment, the invention can be an apparatus for providing additional radiation shielding to a container holding radioactive materials comprising: a tubular shell constructed of a gamma radiation absorbing material and having an inner surface that forms a cavity having an axis, the cavity having an open top end and an open bottom end; a plurality of spacers extending from the inner surface of the shell toward the axis of the cavity, the spacers extending a first height from the inner surface of the tubular shell; and one or more flange members located at or near the open top end of the cavity extending from the tubular shell toward the axis of the cavity, the flange member extending a second height from the inner surface of the shell, the second height being greater than the first height. In a further aspect, the invention can be a system for handling and/or processing radioactive materials comprising: a container having a first cavity for holding radioactive materials, the container having an outer surface and a top surface; a tubular shell having an inner surface that forms a second cavity for receiving the container, the tubular shell comprising at least one spacer extending from the inner surface of the shell toward an axis of the second cavity; the container positioned in the second cavity of the tubular shell, the at least one spacer maintaining the inside surface of the tubular shell in a spaced relationship from the outer surface of the container; and wherein the tubular structure is non-unitary and slidably removable from the container. In a yet further aspect, the invention can be a method of handling and/or processing radioactive materials comprising: a) placing a container having a first cavity containing radioactive materials in a staging area, the container having an outer surface and a top surface; b) providing a tubular shell having an inner surface that forms a second cavity for receiving the container, the second cavity having an open top end and an open bottom end, the tubular shell also comprising at least one spacer extending from the inner surface of the shell toward an axis of the second cavity; and c) positioning the tubular sleeve above the container and lowering the tubular shell so that the container slidably inserts through the open bottom end and into the second cavity, the at least one spacer maintaining the inside surface of the tubular shell in a spaced relationship from the outer surface of the container so as to form a gap between the container and the tubular shell. In still another aspect, the invention is a method of processing and/or removing radioactive materials from an underwater environment comprising: a) submerging a container having a top, a bottom, and a cavity in a body of water having a surface level, the cavity filling with water; b) positioning radioactive material within the cavity of the submerged container; c) raising the submerged container until the top of the containment apparatus is above the surface level of the body of water while a major portion of the container remains below the surface level of the body of water; and d) removing bulk water from the cavity while the top of the container remains above the surface level of the body of water and a portion of the container remains submerged. In an even further aspect, the invention can be a method of processing and/or removing high level radioactive materials from an underwater environment comprising: a) providing a container having a cavity having an open top end and closed bottom end, the container having a top; b) positioning a canister having an open top end and a closed bottom end in the cavity of the container to form a container assembly; c) submerging the container assembly in a body of water; d) positioning high level radioactive material in the canister; e) placing a lid atop the canister that substantially encloses the top end of the canister, the lid having one or more holes; f) raising the submerged container assembly until the top of the container is above a surface level of the body of water while a major portion of the container remains below the surface level of the body of water; and g) removing bulk water from the canister while the top of the container remains above the surface level of the body of water and a portion of the container remains submerged. In another aspect, the invention can be a method of removing spent nuclear fuel from an underwater environment and preparing the spent nuclear fuel for dry storage, the method comprising: a) providing a cask having both gamma radiation and neutron shielding properties, the cask having a top, a bottom and a cavity having an open top end and a closed bottom end; b) positioning a canister having an open end in the cavity; c) submerging the cask and canister into an underwater environment, the canister filling with water; d) positioning spent nuclear fuel within the canister; e) placing a lid atop the open canister thereby substantially enclosing the open end of the canister; f) raising the cask and canister until the top of the cask is above a water level of the underwater environment while a major portion of the cask remains below the water level; g) removing bulk water from the canister while a portion of the cask remains below the water level; and h) raising the entire cask above the water level of the underwater environment. Referring to FIG. 1, a transfer cask 100, according to one embodiment of the present invention, is illustrated. The transfer cask 100 is generally cylindrical in shape and vertically oriented such that its axis is in a substantially vertical orientation. The shape of the transfer cask 100, however, is not limiting of the invention and can include a multitude of other horizontal cross-sectional shapes, including without limitation square, rectangular, triangular and oval shaped transfer casks. The size, height and orientation of the transfer cask 100 also are not limiting of the invention but will be dictated by safety considerations, the desired load to be accommodated and the facility in which it is to be used. The transfer cask 100, as illustrated, is designed for use with and to accommodate a multi-purpose canister (“MPC”) in effectuating HLW transfer procedures. Preferably, the transfer cask 100 can accommodate no more than one canister, the invention is not so limited, however. An example of one suitable MPC is disclosed in U.S. Pat. No. 5,898,747 to Singh, issued Apr. 27, 1999. The invention, however, is not limited to the use of any specific canister structure. Furthermore, in some embodiments, the inventive concepts discussed herein can be incorporated into and/or utilized by transfer casks (or other containment structures) that db not utilize a canister. For example, the inventive concepts discussed herein can be incorporated into and/or implemented into containment structures, such as metal casks, that have the fuel basket built directly into the storage cavity. For exemplary purposes, the transfer cask 100, and the methods discussed herein, will be described in connection with the transport, preparation and handling of spent nuclear fuel (“SNF”). However, the invention is not so limited and can be utilized to handle, transport and/or prepare any type of HLW, including without limitation burnable poison rod assemblies (“BPRA”), thimble plug devices (“TPD”), control rod assemblies (“CRA”), axial power shaping rods (“APSR”), wet annular burnable absorbers (“WABA”), rod cluster control assemblies (“RCCA”), control element assemblies (“CEA”), water displacement guide tube plugs, orifice rod assemblies, vibration suppressor inserts and any other radioactive materials. The transfer cask 100 and its components have a top and bottom. As used herein, “bottom” refers to the end of the transfer cask 100 (or its component) that is closer to the ground than the respective end of the transfer cask 100 (or the component) that is the “top,” when the transfer cask 100 is used in the contemplated vertical orientation of FIG. 1. The terms “top” and “bottom” are not so limited, however, and the transfer cask 100 is not limited to being used in the vertical orientation of FIG. 1. Thus, for example, when the transfer cask 100 is rotated by 90 degrees from the vertical orientation of FIG. 1, the terms “top” and “bottom” refer to ends that are at the same height from the ground, but at opposite ends of the structure and or its components. The transfer cask 100 generally comprises a body 10, a bottom lid 60, a jacket 20 and a top lid 13. The body 10 forms a cavity 6 for receiving SNF. The body 10 functions as a gamma radiation absorbing structure for an SNF load that is located within the cavity 6. The jacket 20 functions to absorb the neutron radiation emanating from the SNF load located within the cavity 6. The jacket 20 circumferentially surrounds a major portion of the height of the body 10 and is adapted to receive a neutron absorbing fluid, such as water, boronated water, or another fluid that is rich in hydrogen. Both the body 10 and the jacket 20 draw the residual heat from the SNF load away from the cavity 6, and eventually removed from the transfer cask 100 via convective cooling forces on the outer surface of the transfer cask 100. As will be described in greater detail below with respect to FIGS. 3 and 4, the jacket 20 is designed to maximize heat removal from the SNF by creating a natural thermosiphon circulation of the neutron absorbing fluid within the jacket 20. The body 10 is positioned atop bottom lid 60. The bottom lid 60 acts as the floor of the cavity 6 formed by the inner surface of the body 10. The bottom lid 60 is constructed so that it adequately serves as a floor portion of the gamma radiation containment boundary, thereby preventing the gamma radiation emanating from the SNF load within the cavity 6 from escaping downward. The bottom lid 60 comprises a plurality of plates in a stacked arrangement. The plates are preferably constructed of steel, lead or another gamma radiation absorbing material. A layer/plate of neutron absorbing material can be implemented into the bottom lid 60 if desired. The bottom lid 60 is connected to the bottom of the body 10. More specifically, the bottom lid 60 is connected to the bottom surface of the bottom flange 12 of the body 10. The bottom lid 60 comprises a plurality of plates that are removable from the body 10 so as to allow transfer of the SNF load out of the bottom of the transfer cask 100 by lowering the SNF through the bottom of the cavity 6. The plates can be connected to the bottom flange 12 via bolts or other hardware. The bottom lid 60 is preferably non-unitary with respect to the body 10, thereby forming a base-to-body interface between the two. O-rings and/or other suitable seals can be implemented to hermetically seal the bottom lid 60 to the body 10. In alternate embodiments, the bottom lid 60 can be integrally formed as part of the body 10 and/or can take on a wide variety of structural detail. For example, the bottom lid 60 can be a thick forging or the like, eliminating the need for a plurality of plates. The top lid 13 is preferably a non-unitary structure with respect to the body 10 so that the top lid 13 can be repetitively secured and unsecured to the body 10 without compromising the structural integrity of the transfer cask 100 and/or the containment boundary. The top lid 13 rests atop a top edge 11 of the body 10 so as to form a lid-to-body interface therebetween. The top edge 11 of the body is formed by the upper surface of an annular ring 115. The top lid 13 is secured to the top edge 11 by extending bolls 63 through holes in the top lid 13 and threadily engaging corresponding bores in the top flange 11. The internal surfaces of the bores are preferably threaded for engagement with the bolts 63. While bolts 63 are illustrated as the connection means, other suitable hardware and connection techniques can be used, including without limitation screws, a tight fit, etc. Referring now to FIGS. 1 and 3 concurrently, the body 10 comprises a first shell 15 and a second shell 16. The body 10 is constructed of gamma radiation absorbing material so as to provide the necessary containment boundary for SNF positioned in the transfer cask 100. While the shells 15, 16 are generally cylindrical in shape, other shapes can be used. For example, the horizontal cross-sectional profiles of the shells 15, 16 can be rectangular, oval, etc. The invention is not limited by the shape of the shells 15, 16. The annular ring 115 is connected to the tops of the shells 15, 16. The annular ring 115 adds structural integrity to the shells 15,16 and provides a solid structure to which the top lid 13 can be secured. The inner surface 116 of the first shell 15 forms a cavity 6 for receiving and holding a canister of SNF. As mentioned above, if desired, the cavity 6 can be adapted to accommodate SNF directly by incorporating a fuel basket assembly directly therein so as to eliminate the need for a canister. The first shell 15 and the second shell 16 are preferably made from steel because of its gamma radiation absorbing and heat conducting attributes. However, other gamma absorbing materials can be used. The second shell 16 concentrically surrounds the first shell 15 so as to form an annular gap 14 therebetween which is filled with a gamma absorbing material, thereby forming an additional layer of gamma absorbing material. The annular gap 14 can be filled with any gamma absorbing material, including without limitation concrete, lead, steel, etc. or combinations thereof. Preferably, the gamma absorbing material used in the annular gap 14 is a material, such as steel, that can adequately conduct heat radially outward away from the cavity 6 so that residual heat emanating from SNF can be removed. It also possible that the annular gap 14 comprise another shell rather than a filled gap. While the body 10 is illustrated and described as a multilayer structure, the body 10 can be constructed as a unitary structure from a single thick shell or from a combination of concrete and metal, such structural details of the body 10 are not limiting of the invention, so long as the necessary cooling and gamma radiation adsorption are provided by the body 10 for the radioactive load to be positioned in the cavity 6. The top edges of the first and second shells 15, 16 are connected to a bottom surface of the annular ring 115 via welding or other connection technique. Similarly, the bottom edges of the first and second shells 15, 16 are connected to the top surface of the bottom flange 12 of the body 10. The bottom flange 12 is a plate-like structure that contains the necessary holes and hardware for both connecting the plates of the bottom lid 16 to the body 10 and connecting the transfer cask 100 to a mating device during canister transfer operations. Referring solely to FIG. 1, the inner surface 116 of the first shell 15 forms the cavity 6 for receiving the SNF load. The cavity 6 is a cylindrical cavity having an axis that is in a substantially vertical orientation. The invention is not so limited however, and the axis could be in a substantially horizontal orientation or another orientation. The horizontal cross-sectional profile of the cavity 6 is generally circular in shape, but is dependent on the shape of the first shell 15, which is not limited to circular. The top end of the cavity 6 is open, providing access to the cavity 6 from outside of the transfer cask 100 (the top lid 13 provides closure to the top end of the cavity 6 when secured to the transfer cask 100). The bottom end of the cavity 6 is also open, and can be closed by the bottom lid 60. More specifically, the top surface 117 of the bottom lid 60 acts as a floor for the cavity 6. Two trunnions 61 are provided at the top of the body 10. The trunnions 61 provide a means by which a lifting device can engage the transfer cask 100 for lifting and transport. The trunnions 61 are preferably circumferentially spaced from one another about 180° apart and made of a material having high strength and high ductility. The invention is not limited to a trunnion, any means for attaching a lilting device can be used, including without limitation, eye hooks, protrusions, etc. Referring now to FIGS. 1 and 3 concurrently, the transfer cask 100 further comprises a jacket 20. The height of jacket 20 is less than the height of body 10. The jacket 20 is preferably tall enough to cover the height of the SNF stored in the cavity 6. The jacket 20 is formed by a shell 120 which is concentric to and surrounds the second shell 16. The shell 120 can be constructed of steel or other materials, such as metals, alloys, plastics, etc. However, it is preferred that the shell 120 be formed of a good heat conducting material, such as steel. In the illustrated embodiment, the shell 120 is formed by a plurality of panels 22. A total of eight panels 22 are used to form the shell 120. The invention, however, is not so limited and the shell 120 can be a unitary shell or consist of any number of panels 22. The shell 120 has a top edge 125 and a bottom edge 126 (best seen in FIG. 4). The jacket 20 comprises a gap/space 19 formed between the shell 120 and the second shell 16 for receiving a neutron absorbing fluid. The gap 19 is adapted to receive a neutron absorbing fluid, such as boronated water, to provide a layer of neutron shielding for the SNF load within the cavity 6. The second shell 16 acts as the inner wall of the gap 19 while the shell 120 acts as the outer wall of the gap 19. The jacket 20 further comprises bottom ring plate 55 and a top ring plate 56 which form the floor and the roof of the gap 19. The top and bottom ring plates 55, 56 are ring-like plate structures that surround the outer surface 121 of the second shell 16. While the bottom ring plate 55 is a single unitary ring-like structure, the top ring plate 56 is formed of a plurality of sections in stepped manner to accommodate the trunnions 61. Of course, either the top or bottom ring plates 55, 56 can be constructed in either manner. The jacket 20 further comprises one or more fill valves 23 located at or near the top of jacket 20. The fill valve 23 is adapted so as to be capable of being moved between an open position and a closed position. When the fill valve 23 is in a closed position, it is hermetically sealed. When the fill valve 23 is in the open position, it allows for efficient filling of the jacket 20 with a neutron absorbing fluid, such as boronated water or the like. The jacket 20 further comprises one or more drain valves (not illustrated). The drain valves are also adapted so as to have an open and a closed position. When the drain valves are in the open position, they allow for removal of the neutron absorbing fluid from the jacket 20. When the drain valves are in the closed position, they are hermetically sealed. As is best visible in FIG. 4, the bottom and top ring plates 55, 56 are respectively connected to the top and bottom edges, 125,126 of the shell 120 in a hermetic manner. Likewise, the inner edges of the bottom and top ring plates 55, 56 are connected to the outer surface 121 of the shell 16 in a hermetic manner. A proper weld will achieve these hermetic connections. The outer surface 121 of the second shell 16 acts as the inner wall of the gap 19 while the inner surface 122 of the shell 120 acts as the other wall of the gap 19. The floor of the gap 19 is formed by the top surface 123 of the bottom ring plate 55. The ceiling of the gap 19 is formed by the bottom surface 124 of the top ring plate 56. The gap 19 is a hermetically sealable space/volume capable of holding a neutron absorbing fluid without leaking. The gap 19, of course, can be other shapes beside annular. Referring now to FIGS. 2 and 3 concurrently, the jacket 20 further comprises a plurality of radial plates 21 positioned within the gap 19. The radial plates 21 are preferably made of steel or another metal or material having good heat conduction properties. Each radial plate 21 comprises a first face 27, a second face 28, an outer lateral edge 25 an inner lateral edge 26, a top edge 24 and a bottom edge 23. The outer lateral edge 25 and inner later edge 26 are vertically oriented. The outer lateral edges 25 of the radial plates 21 are connected to the inner surface 122 of the shell 120 while the inner lateral edges 26 of the radial plates 21 are connected to outer surface 121 of the second shell 16. The radial plates 21 act as fins for improved heat conduction from the body 10, through the jacket 20 and to the atmosphere surrounding the transfer cask 100. In another embodiment, the lateral edges 25, 26 of the radial plates 21 may be radially offset from one another so that a straight line does not exist through the radial plate 21 from the second shell 16 to the jacket 20. For example, the radial plates 21 can be bent so as to have a zig-zag horizontal cross-sectional profile. This prohibits neutron radiation escape through the radial plates 21. The top edge 24 of the radial plate is connected to the bottom surface 124 of the top ring plate 56. The bottom edge 24 of the radial plate 21 is connected to the top edge 123 of the bottom ring plate 55 The radial plates 21 extend radially between the second shell 16 and the shell 120 of the jacket 20, thereby dividing the gap 19 into a plurality of circumferential zones 41A-H. At least one hole 34 (visible in FIG. 4) preferably exists that forms an open passageway between each of the adjacent circumferential zones 41A-H. By providing these holes 34, neutron absorbing fluid can flow freely throughout the entirety of the gap 19 when supplied to a single circumferential zone 41 during the jacket filling procedure. In the illustrated embodiment, the holes 34 are formed by chamfered edges of the radial plates 21. However, the passageways can be provided in any manner desired, for example as a plurality of gaps between the top edge 24 of the radial plate 21 and the top ring plate 56. Referring still to FIGS. 2 and 3, the jacket 20 further comprises a plurality of baffles 40. As will be discussed in further detail below, the baffles 40 facilitate a natural thermosiphon circulation of the neutron absorbing fluid within the gap 19 of the water jacket 20 to assist in heat removal/cooling of the SNF within the cavity 6. The baffles 40 are plate-like structures positioned in the gap 19 in a substantially vertical orientation. The baffles 40 have a top edge 44, a bottom edge 43, a first lateral edge 45 and a second lateral edge 46 (best seen in FIG. 4). The baffles 40 are located between the shell 120 and the second shell 16 in spaced relation from both the shells 120, 16. A single baffle 40 is located within each circumferential zone 41A-41H. The baffles 40 are supported in the gap 19 so that a distance exists between the top and bottom edges of the baffle 40 and the top and bottom ring plates 56, 55 respectively. In other words, the height of baffle 40 is less than the height of the gap 19. The baffles 40 are supported in this floating manner by connecting the lateral edges 45, 46 of the baffles 40 to the first and second faces 27, 28 of the radial plates 21. Welding or other connection techniques could be used. Referring now to FIGS. 3 and 4 concurrently, the structure and functioning of the jacket 20 relative to the thermosiphon circulation within the gap 19 will be discussed in greater detail. The structure and functioning of the jacket 20 relative to the thermosiphon circulation will be discussed in relation to a single circumferential zone 41 with the understanding the principles and structure are applicable to all zones 41A-41H. The baffles 40 comprise a first plate 42 and a second plate 48. The first and second plates 42, 48 are connected to one another along their major surfaces. However, as will be discussed below, this connection is preferably accomplished so that intimate surface contact does not exist between the major surfaces of inner and outer plates 42, 48 of the baffle 40. The inner and outer plates 42, 48 are preferably made of stainless steel. Moreover, while the baffles 40 are illustrated as a plurality of circumferential plates 42, 48 separated by the radial plates 21, a single plate or shell can be used to act as the baffle for the entire gap 19. The baffle 40 is positioned in the gap 19 in radially spaced relation to the outer surface 121 of the second shell 16 and the inner surface 122 of the shell 120. Thus, the baffle 40 divides the gap 19 into an inner region 19A and an outer region 19B. The inner region 19A is that region of space located between the baffle 40 and the outer surface 121 of the second shell 16. The outer region 19B is that region of space located between the baffle 40 and the inner surface 122 of the shell 120. As mentioned above, the height of the baffle 40 is less than the height of the gap 19. As a result, passageways 50, 51 exist between the inner region 19A and the outer region 19B. The passageway 50 is located at or near the top of the gap 19 while the passageway 51 is located at or near the bottom of gap 19. More specifically, the passageway 50 is formed between the top edge of the baffle 40 and a bottom surface 124 of the top ring plate 56. Similarly, the passageway 51 is formed between the bottom edge of the baffle 40 and a top surface 123 of the bottom ring plate 55. The invention is not so limited and passageways 50, 51, could be formed as holes in the baffle 40 itself so long as sufficient fluid passes therethrough between the inner region 19A and the outer region 19B of the gap 19. In such an embodiment, the baffle 40 could be connected to the surface 124 and the surface 123. Holes at or near the top and bottom of baffle 40 could provide the passageways for fluid to flow between the inner and outer regions 19A, 19B. Referring solely to FIG. 4, when SNF is loaded into the cavity 6 of the transfer cask 100, the heal emanating from the SNF conducts radially outward through the body 10. As this heat exits the outer surface 121 of the second shell 16, the heat is absorbed by the neutron absorbing fluid that is located in the inner region 19A of the jacket 20. As the neutron absorbing fluid in the inner region 19A becomes heated, the warmed neutron absorbing fluid rises within the inner region 19A. As a result, cool neutron absorbing fluid from the outer region 19B is draw into the inner region 19A via the passageway 51. The healed neutron absorbing fluid that rose within the inner region 19A is likewise drawn into the outer region 19B via the passageway 50. As the heated neutron absorbing fluid comes into contact with the shell 120, the heat from the neutron absorbing fluid conducts through the shell 120 where it is removed by convective forces on the outer surface 125 of the shell 120. Thus, the neutron absorbing fluid in the outer region 19B cools. As the neutron absorbing fluid cools in the outer region 19B, it flows downward in the outer region 19B until it is adequately cooled and drawn back into the inner region 19A where the process repeats. It is in this manner in which a natural thermosiphon circulation of the neutron absorbing fluid takes place within the gap 19 of the jacket 20. This natural fluid flow is illustrated by the wavy arrows. In order to promote the thermosiphon flow, it may be preferable that the coefficient of thermal conductivity (K(B)) of the baffle 40 in the radial direction be less than the coefficient of thermal conductivity of the neutron absorbing fluid (K(F)) in the gap 19. Making K(B) less than K(F) may help ensure that the neutron absorbing fluid in the outer region 19B remains cooler than the neutron absorbing fluid in the inner region 19A, thereby maximizing the fluid circulation rate. In one embodiment, this can be achieved by making the baffle 40 of two plates 42,48 having a gap between the two. Of course, when the baffle 40 or the neutron absorbing fluid is made of a composite, then it is the effective coefficient of thermal conductivity of the baffle 40 that is preferably less than the effective coefficient of thermal conductivity of the neutron absorbing fluid. Referring now to FIG. 5, a shield 200 according to one embodiment of the present invention is illustrated. The shield 200 is a sleeve-like structure that is designed to slidably fit over a containment apparatus, such as transfer cask 100, to provide additional radiation shielding and missile protection. The shield 200 is intended to be placed over a transfer cask once it is in the staging area (i.e. removed from the fuel pond). Although the term “staging area” generally refers to an area in a facility for drying and other preparations of a cask, as used herein, staging area can be any area of a facility including an area where nothing is being preformed to the cask. Although the shield 200 is designed for use with and to accommodate the transfer cask 100, the invention is not limited to the use of any specific transfer cask. It is to be further understood that the shield 200, in and of itself, is a novel device and can constitute an embodiment of the invention independent of the components of the transfer cask 100. The shield 200 comprises a thick shell 220 and a top plate 210. The top plate 210 is a ring-like plate having a central opening 223. The top plate 210 is connected to the top edge of the thick shell 220. The thick shell 220 has an open bottom end thereby forming a bottom opening 225 of the shield 200. The central opening 223 has a smaller diameter than the bottom opening 225. The diameter of the bottom opening 225 is large enough so that the shield 200 can be slid over the top of the transfer cask 100, as will be discussed with reference to FIG. 6. The inner surface 221 of the shell 220 forms an internal cavity 211 for receiving the transfer cask 100. The cavity 211 has a diameter greater than the diameter of transfer cask 100, or the containment apparatus with which the shield 200 is to be used. The shield 200 further comprises a plurality of eye hooks 212 are welded to the top surface of the top plate 210 and are used by a crane to carry the shield 200. The invention is not limited to eye hooks, any means for attaching a transport device may be used, including trunnions and other protrusions. The shell 220 and the top plate 210 are made of a gamma absorbing material, such as steel, lead, etc. The shield 200 can be as thick as required, preferably at least 5 inches thick. In another embodiment, the shield 200 could be a multi-layer structure rather that a single layer structure. The shield 200 further comprises a plurality of spacers 230 located on the inner surface 221 of the shell 220 and the bottom surface 213 the top plate 210. The spacers 230 are generally L-shaped plates that extend radially into the cavity 211 formed by the shell 220. The spacers 230 comprises a horizontal portion 231 and a vertical portion 232. The horizontal portion 231 extends along the along the bottom surface 213 of the top plate 210 for the entire width of the top plate 210. As will be discussed below with reference to FIG 6, the horizontal portion 231 acts as a flange to support the weight of the shield 200. In an alternative embodiment, the top plate 210 could act as a flange instead of the horizontal portion 231 of the spacers 230. In such an embodiment, the top plate 210 could extend into the cavity 211 rather than connecting solely to the top edge of the shell 230. The horizontal portion 231 extends into the cavity 211 a further distance than does the vertical portion 232. Stated another way, the horizontal portion 23 of the spacer 230 extends from the inner surface 221 of the shell 220 into the cavity 211 by a first distance. The vertical portion 232 of the spacer 230 extends from the inner surface 221 of the shell 220 into the cavity 211 by a second distance. The first distance is greater than the second distance. The vertical portion 232 extends along the inner surface 221 of the shell 220 from the horizontal portion 231 to the bottom of the shield 200. The invention is not so limited, however, and the vertical portion 232 could be segmented or formed from a plurality of pins, bars, etc. Additionally, where the vertical portion 232 is segmented, the segments do not have to be vertically aligned. The spacers 230 are preferably circumferentially spaced from another by about 60° (best seen in FIG. 7), but could comprise more spacers 230 spaced closer together, etc. The spacers 230 are made of a material having high strength and ductility, sufficient so that the horizontal portion 231 is strong enough to support the full weight of the shield 200. Referring to FIG. 6, the shield 200 slidably fits around the transfer cask 100 so as to form a shield-to-transfer cask interface. The shield 200 has a height that is less than the height of the transfer cask 100. As a result, the shield 200 does not extend the full height of transfer cask 100. As will be discussed below, this allows a space to exist between the shield 200 and the ground so that air can circulate under the shield 200 and over the outer surface of the transfer cask 100 when the shield 200 is fitted over the transfer cask 100. The horizontal portion 231 of the spacers 230 acts as a flange and rests on the top surface 56 of the transfer cask 100 while the vertical portion of the spacers 230 contacts the outer surface of the wall of the transfer cask 100. Referring to FIG. 7, the spacers 230 maintain channels 240 between the inner surface of the shell 220 spaced from the outer surface of the transfer cask 100. The spacers 230 divide the gap between the shell 220 and the cask 100 into a plurality of channels 240. The channels 240 allow air to flow between the shield 200 and the transfer cask 100 so as to cool the transfer cask 100 that is heated by the SNF stored in the cavity 6. The channels 240 are not limited to linear passageways and could be formed as tortuous paths from the bottom of the shield 200 to the top of the shield 200. Referring to FIG. 8, air can enter via an opening 241 below the shield 200 and enter into the spaces 240. The air is warmed by heat emanating from the transfer cask 100 and naturally rises within the spaces 240. The warmed air exits the spaces 240 via an exit opening 242 at the top of the shield 200. The wavy arrows indicate this natural thermosiphon/chimney flow. Referring now to FIG. 9, a method of the present invention is illustrated in the form of a flowchart 900. The steps of FIG. 9 will be discussed in relation to the apparatus shown in FIGS. 1-8. In defueling a nuclear reactor and storing the spent nuclear fuel, a transfer cask 100 having cavity 6 and a neutron radiation absorbing jacket 20 surrounding the cavity 6 is provided. Thereby accomplishing step 910. An open multi purpose canister (MPC) is placed in cavity 6 of transfer cask 100, completing step 920. When the embodiment is utilizing a canister and cask, i.e., a dual containment system, the entire structure is thought of as a container having a top, a bottom, and a cavity. The transfer cask 100 with the open MPC is submerged into a fuel pond so that the top of the MPC is below a surface level of the fuel pond. The water from the fuel pond fills the open MPC, thereby completing step 930. When the nuclear fuel is depleted in the nuclear reactor, the spent nuclear fuel is removed from the reactor, lowered into the fuel pond, and placed into the MPC, thereby completing step 940. Once the MPC is fully loaded, a lid is secured to the MPC enclosing the both the spent nuclear fuel and water from the storage pond, completing step 950. A crane or other lifting device is attached to trunnions 61 of transfer cask 100. Once secured to trunnions 61, the crane lifts transfer cask 100, containing the loaded MPC, in an upright orientation toward the water level of the storage pond, completing step 960. The top surface of transfer cask 100 is lifted to be just above the water level so that water from the storage pond can no longer flow into the MPC. Preferably, the top surface of the transfer cask 100 is between 1 to 12 inches above the surface level of the body of water so that a substantial portion of the transfer cask 100 and MPC remains below the surface level of the water in the fuel pond. Additionally, it is to be understood that rather than raising the transfer cask 100 above the surface level of the fuel pond, the water in the fuel pond could be drained until the top of the MPC is above the lowered surface level of the fuel pond. Stated broadly, step 960 can be achieved by relative movement of the transfer cask 100 and the water in the fuel pond. Upon the transfer cask 100 being just above the water level, bulk water is removed from the MPC, thereby completing step 970. The weight within transfer cask 100 has now been reduced in an amount equal to the weight of bulk water removed. At this stage, the lifting device removes transfer cask 100 containing the MPC from the storage pond and places it onto a staging area, completing step 980. While in the staging area, the empty volume of the MPC is filled with water, completing step 990. A removable radiation shield/skirt 200 is then slidably placed around the transfer cask 100. The shield 200 is positioned above the transfer cask 100 by using a crane connected to the eye hooks 212. The shield 200 is lowered so that the open bottom end 225 of the shield 200 slides over the transfer cask. 100. The horizontal portion 231 of the spacer 230 contacts an upper surface of the top ring plate 56 and rests thereupon. Cool air then enters into the chamber 240 and rises within the chamber 240 until exiting at the top. This cool air acts to remove heat emitted by the spent nuclear fuel stored in transfer cask 100. Step 1000 is now complete. The lid is now welded onto the MPC and the spent nuclear fuel is prepared for long term dry-state storage. The water is drained from the MPC and the MPC is filled with an inert gas. Such filling with gas is well known in the art. Thus, step 1010 is completed. The method of the invention can comprise any combination of the steps mentioned above. All of the steps are not necessary to practice the invention.
	abstract	The invention relates to heat power engineering, in particular, to methods that use a working medium for producing useful work from heat of an external source. The method comprises interaction of the working medium with an energy source and interaction of the working medium with an additional low-temperature energy source in the form of the positron state of the Dirac's matter by means of bringing the working medium into quantum-mechanical resonance with said state. The quantum-mechanical resonance is initiated by changing at least one of the thermodynamic parameters of the working medium, while the value of spontaneous fluctuations of the variable parameter in the vicinity of the line of absolute instability in the state diagram of the working medium is predetermined, and the change step for the thermodynamic parameter is set to be lower than the predetermined value of said fluctuations.
042344492	abstract	Radioactive alkali metal is mixed with particulate silica in a rotary drum reactor in which the alkali metal is converted to the monoxide during rotation of the reactor to produce particulate silica coated with the alkali metal monoxide suitable as a feed material to make a glass for storing radioactive material. Silica particles, the majority of which pass through a 95 mesh screen or preferably through a 200 mesh screen, are employed in this process, and the preferred weight ratio of silica to alkali metal is 7 to 1 in order to produce a feed material for the final glass product having a silica to alkali metal monoxide ratio of about 5 to 1.
	description	This application is based upon and claims the benefit of priority from Japanese Patent Application No. 2017-049738 filed on Mar. 15, 2017, the entire content of which is incorporated herein by reference. The embodiments of this invention relate to a thermal-neutron reactor core and a design method for a thermal-neutron reactor core. Some very small, lightweight nuclear reactor systems use a metal hydride as a moderator. It has been demonstrated that the use of the hydride as the moderator could lead to positive temperature reactivity coefficient in the core due to a change in the neutron energy spectrum resulting from rising of core temperature. Such demonstration is disclosed in “Investigation of the TOPAZ-II space nuclear reactor moderator thermal transient”, Robert D. Rockwell, MIT, 1993, the entire contents of which are incorporated herein by reference. Meanwhile, the core of a nuclear reactor is generally required to have a negative temperature reactivity coefficient. If the core does not have the negative temperature reactivity coefficient, such operations as using control rods depending on the core temperature are needed. This means that the safety of the core is not sufficiently guaranteed. As mentioned above, the problem with the core employing the metal hydride as the moderator is that the temperature reactivity coefficient is positive. Another problem is that, when gadolinium (Gd), which is a typical burnable poison, is used to suppress excess reactivity in the core, the temperature reactivity coefficient would become far more positive. Embodiments of the present invention have been made to solve the above problems and an object thereof is to keep the temperature reactivity coefficient from being positive in a range between normal and high temperatures in a thermal-neutron reactor core. According to an aspect of the present invention, there is provided a thermal-neutron reactor core comprising: a solid moderator expanding to a lengthwise direction; a fuel in the moderator, parallel to the lengthwise direction of the moderator, the fuel containing a fissile material; a cooling tube parallel to the lengthwise direction of the moderator; and a plurality of kinds of burnable poison included in the fuel. According to an aspect of the present invention, there is provided a design method for a thermal-neutron reactor core including a solid moderator, the method comprising: deciding a specification of the thermal-neutron reactor core which includes a kind of a fuel, a size of the thermal-neutron reactor core, a composition of the moderator, and a cooling system; determining a neutron energy spectrum based on the specification; selecting a plurality of kinds of burnable poison; examining a temperature dependence of an effective neutron multiplication factor in the thermal-neutron reactor core based on a proportion of the plurality of kinds of burnable poison; deciding whether the proportion is acceptable for an operation of the thermal-neutron reactor core or is not acceptable based on the temperature dependence of the effective neutron multiplication factor, wherein the effective neutron multiplication factor should decrease as temperature rises for the operation. Hereinafter, with reference to the accompanying drawings, a thermal-neutron reactor core and a design method for a thermal-neutron reactor core will be described. The same or similar portions are represented by the same reference symbols and will not be described repeatedly. FIG. 1 is a flowchart showing the procedure of a design method for a thermal-neutron reactor core according to a present embodiment. What is shown here is the procedure of a design method of a thermal-neutron reactor core which includes burnable poison and in which an effective neutron multiplication factor monotonically decreases as operation temperature rises. First, a specification of the thermal-neutron reactor core is set (Step S01). Specifically, major specification of the thermal-neutron reactor core such as the type of nuclear fuel and its form, the type of moderator and its form, the cooling method for the thermal-neutron reactor core, the size of key parts including the diameter and height of the core are set. At this step, the nuclear fuel is supposed not to contain any burnable poison. Hereinafter, as the thermal-neutron reactor core, a small reactor core will be described as an example. FIG. 2 is a cross-sectional view showing the configuration of one-fourth of the small reactor core according to the present embodiment. The diagram shows circumferentially divided one-fourth sector of a small reactor core 1. The small reactor core 1 is in the shape of a column as a whole and includes at least one fuel 2, at least one moderator 3, reflector 4, neutron multiplication materials 5 and at least one cooling tube 6. The small reactor core 1 may have one fuel 2, or one moderator 3, or one cooling tube 6. Although not shown in the diagram, the small reactor core 1 includes a device that has a reactivity control function. The reactivity control method here may be a method by which neutron absorbing materials are inserted or removed, a method that lets voids form or vanish in the core portion, a method by which the reflector is moved, or other method. The small reactor core 1 is housed in a reactor vessel 8. A plurality of fuels 2 and a plurality of moderators 3 are disposed in layers radially and alternately. The moderator 3 at the radially center part has a columnar shape. The fuels 2 and the moderators 3 on the radially outer side of that portion have a cylindrical shape. The fuels 2 and the moderators 3 extend axially. In the radially outermost layer, the reflector 4 having a cylindrical shape is disposed. In such a way as to be adjacent to the radially inner and outer surfaces of a region where the fuels 2 are disposed, neutron multiplication materials 5 having a cylindrical region are disposed. In the layer of the fuels 2, the cooling tubes 6 are disposed with circumferential intervals therebetween. The cooling tubes 6 extend into the small reactor core 1 in such a way as to be parallel to the central axis of the small reactor core 1. In the example shown in FIG. 2, the moderators 3 are placed at the center, and, on the radially outer side thereof, there are three layers of the fuels 2 and the moderators 3. However, the configuration is not limited to this. Instead, the fuels 2 may be disposed in the central portion. The radial thickness of the moderators 3 and the fuels 2, as well as the number of layers, shall be selected according to the design. The size and number of the cooling tubes 6 may be set according to the heat distribution and other factors. The cooling tubes 6 are, for example, a heat absorbing portion of heat pipes. The tubes 6 may be pipes for coolant, such as water. A system in which such tubes pass through the core may be employed. The fuels 2 at least include fissile materials such as uranium, thorium or transuranic elements. The example described below includes uranium as a fissile material. The form of the fuels 2 includes metallic component. However, it is not limited to that. For example, the form may be oxide, nitride or carbide. The fuels 2 have a metallic cladding (not shown) which surrounds the nuclear fuel material, in order to contain fission products resulting from the reaction of fissile materials. The material of the moderators 3 is, for example, metal hydride and solid such as calcium hydride (CaH2) or zirconium hydride (ZrH2). According to the present embodiment, the material of the moderators 3 is CaH2. The neutron multiplication material 5 has the function of multiplying neutrons. Therefore, the neutron multiplication material 5 generates secondary neutrons from primary neutrons generated from the fuels 2 through nuclear reaction. For the reflector 4 and the neutron multiplication material 5, material such as beryllium (Be) may be used. In the present embodiment, Be is used for the reflector 4 and the neutron multiplication material 5. A reactor vessel 8 is hollow cylindrical and has a sealed structure, with both axial ends closed with closure parts (not shown). The cooling tubes 6 pass through one of the closure parts, or both. One feature of metal hydride is that hydrogen dissociates at high temperature. Therefore, there is an upper limit of the operable temperatures at which hydrogen of metal hydride does not dissociate. Hereinafter, the operation temperature is supposed to be between normal temperatures, or about 300K, and about 1,000K. The small reactor core 1 of the present embodiment has the above-described configuration, and the fuels 2 at least partially include a burnable poison. The factors such as type, concentration of the burnable poison are set as follows, based on the design method of the thermal-neutron reactor core of the present embodiment. Such a reactor core is the original small reactor core 1 of the present embodiment. However, hereinafter, even before the setting of burnable poison, those having the same configuration, excluding what the burnable poison is, will be also referred to as the small reactor core 1 of the present embodiment. Then, based on the specification of the small reactor core 1 that is set at the core specification setting step, the neutron energy spectrum of the small reactor core 1 depending on the temperature across the range of the operation temperature is confirmed (Step S02). FIG. 3 is a graph showing the neutron energy spectrum of the small reactor core of the present embodiment. The horizontal axis represents neutron energy (eV). The vertical axis shows neutron flux Φ (n/m2) against each value of the neutron energy. The diagram shows three neutron energy spectra, depending on the operation temperature of the small reactor core 1. The solid curved line “a” shows the case that the temperature is 300K. The dotted curved line “b” shows the case that the temperature is 600K. The broken curved line “c” shows the case that the temperature is 1,000K. The small reactor core 1 is a thermal-neutron reactor core having the moderators 3. Therefore, it is affected by the temperature of the small reactor core 1 in a thermal-neutron region that is thermally in equilibrium with the temperatures of the small reactor core 1. Meanwhile, in a region where the neutron energy is several electron volt (eV) or more, there is little change in the neutron energy spectrum with respect to changing temperature. Specifically, in the thermal neutron energy region between about 0.01 eV and about 1 eV, there are differences in the neutron energy spectrum. In the region where the neutron energy is more than about 1 eV, there are almost no differences. The neutron energy spectrum peaks at about 0.08 eV in the case of 300K, represented by the curved line “a”; at about 0.2 eV in the case of 600K, represented by the curved line “b”; and at about 0.3 eV in the case of 1,000K, represented by the curved line “c”. That is, as the temperature of the small reactor core 1 goes up, the neutron energy spectrum shifts to the higher energy side in the thermal neutron energy region. Then, the dependence of the effective neutron multiplication factor, keff, of the small reactor core 1 on the temperature of the small reactor core 1 is examined (Step S03). First, the cross sections of uranium 235 (U235), a fissile nuclide that constitutes the fuels 2, and of uranium 238 (U238), a parent substance, will be explained, before an example of calculation of the effective neutron multiplication factor, keff, of the small reactor core 1 is shown. FIG. 4 is a graph showing the shape of the spectrum of the reaction cross section of uranium 235 in comparison with the shape of the neutron energy spectrum of the small reactor core. The horizontal axis represents Neutron Energy, i.e., neutron energy (eV). As for the reaction cross section, the vertical axis represents Cross Section (barns). One barn, the unit of the cross section, is 10−24 cm2 or 10−28 m2. This diagram of cross section is excerpted from JENDL-4.0 Data, Nuclear Data Research Group, Japan Atomic Energy Agency, Feb. 2, 2016. In FIG. 4, the solid curved line, CT1, represents the entire cross section. The one-dot chain curved line, CF1, represents cross section σf of fission reaction. The two-dot chain curved line, CC1, represents cross section σc of neutron capture reaction. The broken curved line, CE1, represents cross section σe of neutron elastic scattering. These are the same as in the diagrams shown below regarding each of the cross sections. The superimposed neutron energy spectrum is the same as in FIG. 3, and should be deemed relative values even though its vertical axis is not displayed. When the temperature of the small reactor core rises from 300K to 600K, neutron flux Φ in a region represented as an energy range A significantly increases. Meanwhile, with respect to the increase in the neutron energy in energy range A, the degree of a change in fission cross section σf of U235 indicated by one-dot chain line is not significant compared to the degree of a change in neutron flux Φ with respect to the increase in neutron energy in energy range A. Therefore, fission response rate σf·Φ increases as a result. FIG. 5 is a graph showing the spectrum of the reaction cross section of uranium 238 in comparison with the neutron energy spectrum of the small reactor core. The horizontal and vertical axes are the same as those in FIG. 4. In FIG. 5, the solid curved line, CT2, represents the entire cross section. The broken curved line, CE2, represents the cross section of elastic scattering. The two-dot chain curved line, CC2, represents cross section σc of neutron capture reaction. The one-dot chain curved line, CF2, represents cross section σf of fission reaction. This diagram of cross sections is excerpted from Non-Patent Document 2. In a region greater than or equal to several dozens of eV, there are multiple peaks of large capture cross section σc. As the temperature of the small reactor core 1 rises, neutron flux Φ in an energy range B, as shown in FIG. 5, shifts to a higher energy side. Meanwhile, as the temperature of the small reactor core 1 rises, the peaks of capture cross section σc of uranium 238 spread due to Doppler effect, and the effective neutron multiplication factor keff of the small reactor core 1 is suppressed regardless of the change of spectrum in the energy range B. FIG. 6 is a graph showing the temperature dependence of the effective neutron multiplication factor when no burnable poison is added to the nuclear fuel of the small reactor core. The horizontal axis represents the temperature of the small reactor core 1. The vertical axis represents the effective neutron multiplication factor keff of the small reactor core 1. The temperature reactivity coefficient of the small reactor core 1 is a differential value of reactivity ρ with respect to temperature T of the small reactor core 1. Reactivity ρ is calculated by the following equation (1), based on the effective neutron multiplication factor keff. Reactivity ρ is in one-on-one relationship with the effective neutron multiplication factor keff.ρ=(keff−1)/keff (1) Accordingly, with respect to temperature T of the small reactor core 1, when the differential value of the effective neutron multiplication factor keff is positive, the temperature reactivity coefficient is positive, and, when differential value of the effective neutron multiplication factor keff is negative, the temperature reactivity coefficient is negative. Thus, whether the temperature reactivity coefficient is positive or negative can be evaluated. Or, instead, as described below, the temperature dependence of the effective neutron multiplication factor keff of the small reactor core 1 may be used as a basis for evaluation. As described above, in a region where the temperature of the small reactor core 1 is between about 300K and about 600K, fission reaction rate σf·Φ of U235 increases. Therefore, as the temperature of the small reactor core 1 rises, the effective neutron multiplication factor keff of the small reactor core 1 increases. Meanwhile, in a region where the temperature of the small reactor core 1 is between about 600K and about 1,000K, neutron flux Φ of energy range B increases. But due to the Doppler effect, contributions from neutron capture reaction rate σc·Φ by U238 increase. As a result, with respect to the temperature of the small reactor core 1, the effective neutron multiplication factor keff would change in such a way as to have a peak at about 600K. Therefore, in the temperature region between about 300K and about 600K, the temperature reactivity coefficient is positive. In the region between about 600K and about 1,000K, the temperature reactivity coefficient is negative. Then, based on a change in the neutron energy spectrum, a burnable poison is selected (Step S04). In this process, such factors are taken into account as whether a peak of the capture cross section in the thermal neutron region exists or not, and the trend of the changing capture cross section with respect to a change in neutron energy. As for the burnable poison, the following cases will be sequentially explained: when gadolinium (Gd), an element conventionally used, is selected; when cadmium (Cd) is selected; when europium (Eu) is selected; and when a combination of Cd and Eu is selected. FIG. 7 is a graph showing the shape of the spectrum of the reaction cross section of gadolinium 155 in comparison with the shape of the neutron energy spectrum of the small reactor core. The horizontal axis and the vertical axis are the same as those in FIGS. 4 and 5. In FIG. 7, the solid curved line, CT3, represents the entire cross section. The two-dot chain curved line, CC3, represents cross section σc of neutron capture reaction. The dotted curved line, CE3, represents cross section σe of elastic scattering. This diagram of cross sections is excerpted from Non-Patent Document 2. As shown in FIG. 7, in response to an increase in neutron energy in the energy range A, neutron capture cross section σc of Gd155 monotonically and significantly decreases. As the temperature of the small reactor core rises from 300K to 600K and to 1,000K, neutron flux Φ in the energy range A shifts to the higher energy side. As a result, as the temperature of the small reactor core 1 rises, capture reaction rate σc·Φ of Gd155, a minus-side factor of the effective neutron multiplication factor keff of the small reactor core 1, decreases. FIG. 8 is a graph showing the temperature dependence of the effective neutron multiplication factor when gadolinium is added as burnable poison to the nuclear fuel of the small reactor core. The horizontal axis represents temperature (K) of the small reactor core 1. The vertical axis represents the effective neutron multiplication factor keff of the small reactor core 1. Solid curved line A represents the temperature dependence of the effective neutron multiplication factor keff when no burnable poison is added to nuclear fuel, and is the same as the one shown in FIG. 6. Solid curved line BT shows the temperature dependence of the effective neutron multiplication factor keff when burnable poison is added to the nuclear fuel of the small reactor core 1. If gadolinium (Gd) is added as burnable poison to the nuclear fuel, the effective neutron multiplication factor keff monotonically increases in response to a rise in temperature between 300K and 1,100K. This means that the temperature reactivity coefficient is positive. Moreover, the slope of the increase in the effective neutron multiplication factor keff is greater than the slope of the increase in the effective neutron multiplication factor keff in a range of 300K to about 400K with no gadolinium (Gd) added. The value of the effective neutron multiplication factor keff when Gd is added to the nuclear fuel in the case of FIG. 8 is a result of taking into account a decrease of the effective neutron multiplication factor keff resulting from the added burnable poison on the basis of the effective neutron multiplication factor keff of the case when no burnable poison is added to the nuclear fuel. That is, since it is sufficient in confirming the trend of changes in the effective neutron multiplication factor keff that occur with changing temperature, no adjustments have been made again to the effective neutron multiplication factor keff based on core criticality conditions. In order to clarify factors behind the positive temperature reactivity, coefficient, each factor has been evaluated by looking into how the effective neutron multiplication factor keff would change with changing temperature. Broken curved line B1 shows a change in temperature with only moderator or metal hydride. In this case, as the temperature rises, the effective neutron multiplication factor keff increases. Two-dot chain curved line B2 shows a change in temperature with only multiplication material and reflector. In this case, the effective neutron multiplication factor keff remains almost constant even as the temperature changes. Dotted curved line B3 shows a change in temperature with only fuel or uranium. In this case, as the temperature rises, the effective neutron multiplication factor keff decreases. Accordingly, it is clear that an increase in the effective neutron multiplication factor keff that is associated with an rise in the temperature of the moderator, as indicated by broken line B1, is a factor that makes the temperature reactivity coefficient positive when Gd is added as burnable poison to nuclear fuel. This is because: as shown in FIG. 7, in response to an increase in neutron energy in energy range A in FIG. 7, neutron capture cross section σc of Gd155 monotonically and significantly decreases; and, as a result, neutron capture reaction rate σc·Φ, which contributes to shifting the effective neutron multiplication factor keff to the negative side, decreases. FIG. 9 is a graph showing the spectrum of reaction cross section of cadmium 113 in comparison with the neutron energy spectrum of the small reactor core. The horizontal and vertical axes are the same as in FIG. 7. In FIG. 9, the solid curved line, CT4, represents the entire cross section. The two-dot chain curved line, CC4, represents cross section σc of neutron capture reaction. The dotted curved line, CE4, represents cross section σe of elastic scattering. This diagram of cross sections is excerpted from Non-Patent Document 2. As for Cd 113, isotopically enriched products are available. As shown in FIG. 9, in the case of Cd 113, neutron capture cross section σc peaks at about 0.2 eV in the thermal-neutron energy region. Therefore, in the thermal-neutron region, until neutron energy reaches about 0.2 eV, neutron capture cross section σc increases in response to a rise in neutron energy. In the energy region over about 0.2 eV, neutron capture cross section σc decreases in response to a rise in neutron energy. As for the state of the thermal-neutron region of the neutron energy spectrum for a change in the temperature of the small reactor core, as described above along with FIG. 3, there is a peak near 0.08 eV when the temperature of the small reactor core is 300K; there is a peak near 0.2 eV when the temperature is 600K. As the temperature of the small reactor core changes from 300K to 600K, neutron capture cross section σc of Cd113 increases. However, when the temperature of the small reactor core changes from 600K to 1,000K, neutron capture cross section σc of Cd 113 decreases. As a result, as the temperature of the small reactor core changes from 300K to 600K and then to 1,000K, neutron capture reaction rate σc·Φ, which causes a drop in the effective neutron multiplication factor keff, increases and then decreases. FIG. 10 is a graph showing the temperature dependence of the effective neutron multiplication factor when concentrated cadmium 113 is used as burnable poison for nuclear fuel of the small reactor core. The horizontal axis represents temperature (K) of the small reactor core. The vertical axis represents the effective neutron multiplication factor keff of the small reactor core. As described above, as the temperature of the small reactor core changes from 300K to 600K and then to 1,000K, neutron capture reaction rate σc·Φ of Cd113, which causes a drop in the effective neutron multiplication factor keff, increases and then decreases. As a result, the effective neutron multiplication factor keff of the small reactor core changes in an opposite way as shown in FIG. 10, so that the effective neutron multiplication factor keff first decreases and then increases. In this case, there is a need to correct the situation where the temperature reactivity turns positive in a region close to a state where the temperature of the small reactor core is 1,000K. FIG. 11 is a graph showing the shape of spectrum of reaction cross section of europium 151 in comparison with the shape of the neutron energy spectrum of the small reactor core. In FIG. 11, the solid curved line, CT5, represents the entire cross section. The two-dot chain curved line, CC5, represents cross section σc of neutron capture reaction. The dotted curved line, CE5, represents cross section σe of elastic scattering. This diagram of cross sections is excerpted from Non-Patent Document 2. As for Eu, isotopically enriched products are available. As shown in FIG. 11, in the case of Eu 151, there is a resonance absorption peak of neutron capture cross section σc at about 0.4 eV of energy. From about 0.2 eV to about 0.3 eV, neutron capture cross section σc increases. FIG. 12 is a graph showing the temperature dependence of the effective neutron multiplication factor when concentrated europium 151 is used as burnable poison for nuclear fuel of the small reactor core. The horizontal axis represents the temperature of the small reactor core 1. The vertical axis represents the effective neutron multiplication factor keff of the small reactor core 1. As shown in FIG. 12, in a temperature region less than about 500K, the effective neutron multiplication factor keff increases as the temperature rises, which means the temperature reactivity coefficient is positive. In a temperature region greater than or equal to about 500K, the effective neutron multiplication factor keff decreases in response to a rise in temperature. This means that the temperature reactivity coefficient is negative. As described above, when a minute amount of Cd 113 is added, the temperature reactivity coefficient is negative in the low-temperature region. When a minute amount of Eu 151 is added, the temperature reactivity coefficient is negative in the high-temperature region. FIG. 13 is a graph showing the temperature dependence of the effective neutron multiplication factor when cadmium and europium with natural compositions are used as burnable poison for nuclear fuel of a small reactor core, for comparison with the present embodiment. The natural isotopic composition of cadmium is approximately: 28.7 percent of Cd 114, 24.1 percent of Cd 112, 12.8 percent of Cd 111, 12.5 percent of Cd 110, 12.2 percent of Cd 113, 7.5 percent of Cd 116, 1.3 percent of Cd 106 and 0.9 percent of Cd 108. The natural isotopic composition of europium is approximately: 52.2 percent of Eu 153 and 47.8 percent of Eu 151. When cadmium and europium each with natural isotopic composition are used, the effective neutron multiplication factor keff of the small reactor core 1 increases in response to a rise in temperature, as long as the temperature of the small reactor core 1 is between about 300K and about 500K. When the temperature of the small reactor core 1 is between about 500K and about 1,000K, the effective neutron multiplication factor keff of the small reactor core 1 decreases in response to the rise in temperature. That is, the temperature reactivity coefficient is positive when the temperature of the small reactor core 1 is between about 300K and about 500K. The temperature reactivity coefficient is negative when the temperature is between about 500K and 1,000K. The next step in the design method of the thermal-neutron reactor core is to change the fractions of the selected burnable poisons and survey a change in the effective neutron multiplication factor keff of the thermal-neutron reactor core (Step S05). Then, a determination is made as to whether there is an acceptable case where the temperature reactivity coefficient is negative across the operation temperature region (Step S06). A combination described below of cadmium 113 and europium 151 is acceptable. FIG. 14 is a graph showing the temperature dependence of the effective neutron multiplication factor keff when cadmium 113 and europium 151 are used as burnable poison for nuclear fuel of the small reactor core. The diagram shows the temperature dependence of the effective neutron multiplication factor keff when a predetermined amount of concentrated Cd and a predetermined amount of concentrated Eu are added to nuclear fuel. The predetermined amount of concentrated Cd means that: the degree of concentration of Cd 113 is 96 percent, and the number-density ratio of Cd to nuclear fuel is 0.0072 percent. The predetermined amount of concentrated Eu is: the degree of concentration of Eu 151 is 96 percent, and the number-density ratio of Eu to nuclear fuel is 0.02 percent. As shown in FIG. 14, when the temperature of the small reactor core rises from 300K to 400K, the effective neutron multiplication factor keff slightly decreases. Accordingly, the temperature reactivity coefficient is negative although the absolute value thereof is small. As the temperature of the small reactor core rises to as high as 1,000 k from 400K, the effective neutron multiplication factor keff remarkably decreases. In this case, the temperature reactivity coefficient is negative. In this manner, minute amounts of two types of isotopes, Cd 133 and Eu 151, are added to nuclear fuel, and the temperature reactivity coefficient of the small reactor core is therefore negative at between 300K and 1,000K. If the determination result indicates an acceptable case (Step S06, YES), a decision on the concentration of burnable poison in the acceptable case is made based on criticality and burning conditions (Step S07). The following is a result of adjustments intended to satisfy the criticality and burning conditions: the degree of concentration of Cd 113 in total Cd is 96 percent, and the number-density ratio of Cd to nuclear fuel is 0.0072 percent; and the degree of concentration of Eu 151 in total Eu is 96 percent, and the number-density ratio of Eu to nuclear fuel is 0.02 percent. The effective neutron multiplication factor keff on the vertical axis of FIG. 14, too, is a result of these adjustments. If the determination result indicates no acceptable case (Step S06, NO), a series of steps from the selection of burnable poison to the determination is carried out again. As described above, according to the present embodiment, as for the thermal-neutron reactor, the temperature reactivity coefficient can be prevented from becoming positive in a range from normal to high temperature. While the embodiment of the present invention has been described, this embodiment is presented by way of example and not intended to limit the scope of the invention. According to the embodiment, a plurality of fuels 2 and a plurality of moderators 3 are disposed in layers radially and alternately. However, the present invention is not limited to that. The groups of fuels 2 and moderators 3 each of which surrounds each of fuels 2 may be located in a lattice manner. According to the embodiment, cadmium 113 and europium 151 are used as burnable poison. However, the present invention is not limited to them. Other various combinations of isotopes having a neutron absorption effect can produce similar results to those described in the embodiment, as long as their neutron capture cross sections decrease in response to a rise in temperature within the operation temperature range. According to the embodiment, the present invention is applied to a small thermal-neutron reactor that uses metal hydride as moderator, because it is especially effective. However, the present invention is not limited to this. That is, even when other moderators are used, as long as it is a thermal-neutron reactor having moderator, a similar method to the one described in the embodiment is applied. According to the embodiment, the nuclear fuel is uranium. However, the present invention is not limited to this. The present invention may be applied to mixed oxide fuel, which is made of uranium and plutonium, for example. Furthermore, the above-described embodiments may be put to use in various different ways and, if appropriate, any of the components thereof may be omitted, replaced or altered in various different ways without departing from the spirit and scope of the invention. All the above-described embodiments and the modifications made to them are within the spirit and scope of the present invention, which is specifically defined by the appended claims, as well as their equivalents.
047864605	abstract	In an installation for handling assemblies between the main vessel of a fast neutron nuclear reactor and an adjoining vessel, use is made of a transfer hood formed by a thick tube whose inclined axis can be moved alternatively into the prolongation of the axes of two inclined ramps opening into the main vessel and adjoining vessel respectively, by the rotation of a platform of vertical axis bearing the hood and also two flaps which close the ramps when the reactor is operating, so that an assembly-transferring pot can be guided isostatically inside the ramps and the hood, the platform also possibly comprising a spare flap.. Application to all assemblies forming a reactor core, such as fuel assemblies, lateral neutron screening assemblies and absorbing assemblies.
060552956	description	DETAILED DESCRIPTION OF THE INVENTION FIG. 1 schematically depicts one possible embodiment of a computerized x-ray imaging system 10 which employs an apparatus 38 for providing automatic collimation according to the present invention. The x-ray imaging system 10 is used for diagnostic imaging studies of the peripherals (e.g. legs, arms, neck, and head). The x-ray system 10 comprises a computerized imaging device 12 which includes an x-ray source 14 and a collimator 16. The x-ray source 14 produces an x-ray beam 18 which is collimated by the collimator 16. The collimator 16 may be one of several types of collimators generally used in x-ray studies of the peripherals. For example, the collimator 16 may be a block, a multi-leaf or a finger collimator. The collimated x-ray beam 20 passes through the area of interest 24 of the body 22 and strikes an x-ray image intensifier 26. In other embodiments of the x-ray imaging system, the collimator may be located immediately in front of the image intensifier. The image intensifier 26 processes the x-ray beam 20 so that it can be recorded by recording media 28 such as film or a CRT. The imaging device 12 is horizontally movable in the direction of arrow 30 so that x-ray images can be taken of the peripherals at a plurality of imaging stations. Leg studies may typically include 5-7 imaging stations. Horizontal movement of the imaging device 12 is provided by a stepper motor 32 which is controlled by a stepper controller 34. A contrast media injection device 36 is provided for injecting a contrast media into the patient 22 just prior to diagnostic imaging. Since x-ray diagnostic imaging is well known and commonly used, the details of these components need not be set forth herein any greater detail. The automatic collimation apparatus 38 interfaces with the collimator 16 of the system 10. The automatic collimation apparatus 38 performs a method in which an x-ray image of the peripherals is segmented into body parts and non-body parts. The method uses this information to provide the collimator 16 with appropriate collimator settings during image acquisition. The settings are used by the collimator 16 to adjust itself to cover as much of the non-body region and as little of the body region as possible, given the collimator's hardware constraints. The task of locating the body in an x-ray fluoroscopy image is difficult for a number of reasons. First, segmentation should take place without knowing which part of the body is being looked at. Second, since x-ray fluoroscopy studies use low radiation dosages, the images generally have low signal to noise ratios. Third, soft tissue boundaries often have very poor contrast. Due to the poor contrast makes, conventional edge detection algorithms fail to detect these boundaries. Fourth, existing collimation and noise make local intensity characteristics at a pixel inadequate for determining if it belongs to the body. Finally, segmentation and automatic collimation need to be done at image acquisition time. This places tight constraints on the complexity of image processing operators which can be used. The method performed by the automatic collimation apparatus 38 of the present invention successfully overcomes these difficulties. When the method is implemented as software, it operates robustly and efficiently on noisy, low contrast, possibly pre-collimated x-ray fluoroscopy images. In comparison with manual segmentation of body parts, the method has very high (>95%) sensitivity and specificity. In one illustrative example of the method efficiently implemented as software, the method runs in less than 500 milliseconds or better per station, on a common 200 MHz Pentium Pro PC running Windows NT 4.0. With the use of parallelism and hardware acceleration, the running time can be further improved. FIG. 2 is a flow chart depicting the steps of the automatic collimation method of the present invention. The method segments the body regions in an x-ray image of the peripherals from the background (existing collimation, direct exposure). The information about where the body is in an image is used to provide settings for the collimator 16 of the x-ray imaging system 10 shown in FIG. 1. In step A of the block diagram of FIG. 2, the X-ray source of the x-ray imaging system is adjusted to provide an appropriate low-dosage fluoroscopy, and the collimator is adjusted to predetermined default settings for collimation. This step is repeated at each station. In step B, one of the incoming images at each station is preprocessed by downsizing and smoothing the image. These preprocessing steps are well known in the art and popularly used, thus further elaboration is not required here. Step C involves detecting soft-tissue boundaries using directional curvatures of intensity profiles. Finding the soft-tissue boundaries accurately is critical because the subsequent steps of global feature extraction and classification rely on this information. Prior art edge detection methods often fail to detect soft-tissue boundaries in x-ray images because the boundaries often have very low contrast and unusually defined intensity distributions. Soft-tissue boundaries are detected in the present invention by determining negative curvature points along line profiles of intensity in multiple chosen scanning directions. These provide a reliable indicator of low-contrast boundaries, such as soft-tissue. Even for very low contrast or hazy soft-tissue boundaries, well defined points of negative curvature exist on the line profiles of intensity. This is illustrated in FIG. 3A-3C. FIG. 3A depicts a peripheral x-ray image with low-contrast soft tissue boundaries near the ankle. An intensity profile along a horizontal line passing through the region of interest in FIG. 3A has points of negative curvature corresponding to these soft-tissue boundaries as shown in FIG, 3B. From a histogram equalized image of the line profile curvatures, very low-contrast boundaries are clearly preserved in the negative curvature image of FIG. 3C. Curvatures of line profiles of intensity are computed as follows with the following formulas: EQU Idiff(i,j)=I(i,j)-I(i,j-w))/wdenom(i,j)=w*sqrt(I+Idiff(i,j).sup.2) EQU hcurv(i,j)=(atan(Idiff(ij+w))-atan(Idiff(i,j)))(denom(i,j+w)+(denom(i,j)) I(i,j) is defined as the normalized intensity at pixel (i,j), w defines a parameter related to the window width and is dependent on the input image size, and atan is the are tangent function. The above computation is for horizontal curvatures of line profiles of intensity, however, similar computations can be performed for vertical or other directional curvatures of line profiles of intensity. In the present invention, all positive curvature values are ignored and therefore, removed by zeroing out the positive curvature values. It has been found that soft tissue boundaries are better captured by considering only negative curvatures. The curvature values shown in FIG. 3C are a combination of negative horizontal and vertical curvatures. The combined curvature value at a pixel is the minimum of the horizontal and vertical curvatures. Although the negative curvature image in FIG. 3C reliably indicates all soft-tissue boundaries, it still contains numerous spurious boundaries. Curvature is a second-order statistic and is therefore, noisy. In order for negative curvatures to be useful, the spurious boundary pixels must be reliably removed while preserving all the important boundaries of the soft-tissues. Accordingly, in step D of the block diagram of FIG. 2, the noise in the negative curvature image is adaptively removed using magnitude and alignment information. Magnitude information alone is not used because it may eliminate very faint soft-tissue boundaries which are desirable to locate. Noise is adaptively removed from the negative curvature image by strengthening well aligned curvature pixels and then finding an adaptive threshold based on the cumulative histogram of curvature values in the image. FIGS. 4A-4C depict adaptive noise removal from negative curvature images. FIG. 4A depicts an intensity image (X-ray image) and FIG. 4B depicts a corresponding noisy line profile negative curvature image. FIG. 4C depicts the negative curvature image of FIG. 4B after adaptively removing the noise. The images of FIGS. 4B and 4C are histogram equalized to show detail. Step E of the method depicted in FIG. 2 involves dividing the image into "regions". This is accomplished by providing the boundaries of significant regions in the image with a one-pixel representation of the cleaned up negative curvatures. The one-pixel representation is obtained by finding the local extrema in horizontal and vertical directions, combining them and performing simple noise removal using conventional connected components analysis techniques. The region information is for subsequently extracting global feature values which in turn, are used for classification as explained further on in greater detail. FIG. 5B shows the region boundaries for the image shown in FIG. 5A. In step F of the block diagram of FIG. 2, appropriate features such as range of intensity values, size, etc., are computed along horizontal and vertical lines within each region created in step E. The method of the invention preferably uses three features for segmentation. These features are homogeneity, representative intensity, and station number. Homogeneity is the minimum amount of intensity variation along chosen directions, per pixel, inside a region. Representative intensity is the median intensity in a region. Station number is number of the current station with respect to the full-leg study. Station numbers start at 0 at the pelvic region. Due to running time constraints, it is necessary that these features be simple. Additional features such as, the size of a region, the location of a region with respect to the station and with respect to the full-leg study, the variance of intensities in the region, etc., have been tried. However, supervised decision tree methods used for classification in the present invention, advantageously indicate which features are the most useful for discrimination. These and traditional feature selection methods have shown that no improvement in the classification results are obtained by adding more features to the above set of three. Step G of the method involves inter-region and intra-region propagation of the features. The features are first computed along scan lines in chosen directions in the image and then efficiently propagated over entire regions. Well known adaptive smoothing techniques are used herein for feature value propagation. FIGS. 6A-6C show the homogeneity and representative intensity feature values for a typical image after propagation. In step H of the block diagram of FIG. 2, each pixel in the image is classified as a body part or a non-body part, based on its feature values, using a decision tree. This involves constructing a set of rules which enable body and non-body regions to be determined on the basis of the feature values. The rules are constructed using supervised learning and are therefore, referred to herein as automatically learned classifiers. Automatically learned classifiers advantageously improve automatically over time, as new data comes in. Binary decision trees are used as the specific classifiers in the present invention. Binary decision trees are easy to understand and analyze and make classification very fast because if/else statements are used. Manually found and hard-coded rules typically used in prior art classification, are not used in the present invention because they may not generalize well, and the rules may need to be reconstructed when new data arrives. A predetermined number of data points (pixels in the image) are randomly selected as a training set. The training set is then used in a conventional decision tree method or algorithm to automatically construct the binary decision tree. The preferred decision tree method used is a conventional classification and regression tree (CART) method. This method is described by Breiman et al., Classification and Regression Trees, Chapman & Hall Publishers, 1984 (Software available from Salford Systems, Inc.). Since the CART method is well known, it need not be set forth in any great detail herein. However, some of the more important points of this method will now be described. The CART method takes as input a collection of labeled training instances, each instance having some attributes and a class label and produces a hierarchical decision tree as output. In the present invention, the instances are individual pixels, the attributes are the features computed above and the class labels are body part (1) and non-body part (0). CART then is used to construct binary decision trees from the data. At each stage, CART analyzes the training set to determine the test ("attribute?value.fwdarw.") that best discriminates between the classes, based on a feature evaluation criterion. The training set is then split into two subsets based on the best test. Tree growing continues recursively until no more nodes can be created. Once a full tree is constructed, CART prunes back the tree to remove noise-fitting nodes and/or marginally useful nodes, based on a portion of the training set that is reserved for this purpose. The preferred binary decision tree used in the present invention is relatively small and has only 160 terminal nodes. Having a small tree is important because it shows that the chosen features are appropriate for the classification task, and that the tree has a high probability of classifying unseen data correctly. If for example, the selected number of data points includes 95,000 data points, a tree with 95,000 terminal nodes can be theoretically built. A 160-node tree, which has high accuracy on hundreds of thousands of unseen data points, indicates that the features used are appropriate for classification. In using the binary tree for classification, the feature vectors at each pixel are individually "dropped down" the decision tree until a terminal node is reached. The label at the node is then assigned to the pixel. It should be understood that although the CART method is preferred, other decision tree methods or algorithms may be used if desired. In step I of the block diagram of FIG. 2, the classification result is post-processed to remove noise. This involves smoothing labels over regions and performing a connected components analysis. Such image processing operators are well known in the art and need no further description. In step J, a setting for the collimator is automatically determined from the classification result of step I. The collimator setting is selected to cover as much of the non-body part region as is possible while leaving uncovered as much of the body part as is possible. The collimator settings are automatically tailored to the constraints of the particular collimator used, for example, by taking into account the number of leafs, degrees of freedom, etc. In step k, the imaging system records the automatically computed collimator setting parameters which are subsequently used by the collimator. In step 1, the x-ray source is moved to the next station and the method is repeated until all the stations are processed and the collimation parameters are recorded by the imaging system. FIG. 7 depicts a graphical user interface which may be displayed by the automatic collimation apparatus. The images depicted are for a full leg study. It should be noted that other peripheral studies can be used as well. The interface displays multiple station input images 40 for the full leg, the input image of a single station 42, the segmentation result for the one station 44, and results for the full leg 46. It is understood that the above-described embodiments illustrate only a few of the many possible specific embodiments which can represent applications of the principles of the invention. For example, a human override option may be provided for allowing the physician or operator to override the automatically selected collimator setting if desired. This and other numerous modifications and changes can be made by those skilled in the art without departing from the spirit and scope of the invention.
	description	Aspects of the invention relate to a lens system for multiple-beam charged particle applications, such as inspection system applications, testing system applications, lithography system applications and the like. In particular, aspects of the invention relate to a lens system for a plurality of charged particle beams, especially to a lens system comprising a plurality of lens openings for the respective charged particle beams, and hence to a lens system for multi-beam applications. Further aspects of the invention relate to a multiple charged particle beam device, and to a method for operating a charged particle beam device. Charged particle beam apparatuses are used in a plurality of industrial fields. Testing of semiconductor devices during manufacturing, exposure systems for lithography, detecting devices and inspection systems are some examples of these fields. In general, there is a high demand for structuring and inspecting specimens within the micrometer or nanometer scale. On such a small scale, process control, inspection or structuring is often done with charged particle beams, e.g. electron beams, which are generated and focused in charged particle beam devices, such as electron microscopes, electron beam pattern generators or charged particle inspection systems. Charged particle beams offer superior spatial resolution compared to e.g. photon beams due to their short wavelengths. However, for a given beam diameter, the charged particle beam current limits the throughput of charged particle beam systems. Since further miniaturization of e.g. structures to be imaged is necessary, the charged particle beam diameter has to be decreased. As a result, the beam current for individual beams, and thus the throughput, is decreased. In order to increase the total charged particle beam current, thus increasing the throughput, a plurality of charged particle beams can be used. In this manner, the throughput can be increased proportional to the number of columns in a multi-column system. One option for obtaining a plurality of charged particle beams may be combining several single beam columns with each other. However, some components, especially magnetic lenses, cannot be miniaturized sufficiently, since the magnetic field cannot be arbitrarily increased. Thus, the columns have to be spaced such that the distance between electron beams is 100 mm to 200 mm. To overcome this problem, U.S. Pat. No. 3,715,580 utilizes a magnetic lens with a circular excitation coil providing two holes, each for a single electron beam. Thereby, the continuous rotational symmetry of previous lenses is abandoned since the hole (optical axis) for each electron beam has different distances from the position of the excitation coil. This lack of symmetry of the magnetic focusing field results in additional aberrations, and thus reduces the obtainable resolution. Further, U.S. Pat. No. 7,576,917 describes a multi-axis lens with identical individual sub-units. The multi-axis lens allows close packing of lenses in a one dimensional array, but there remains a desire to reduce the spacing even further. Especially with the multi-axis lens, the spacing to a neighboring second array remains large. Since there is a strong desire for improving resolution, for simplifying manufacturing and for minimizing aberrations in such systems, it is an object of the present invention to further improve state of the art devices. In view of the above, a lens system according to independent claim 1, a multiple charged particle beam device according to claim 14, and a method according to independent claim 15 are provided. Further advantages, features, aspects and details are apparent from the dependent claims, the description and drawings. According to one embodiment, a lens system for a plurality of charged particle beams comprises: A lens body with a first pole piece, a second pole piece and a plurality of lens openings for the respective charged particle beams; a common excitation coil arranged around the plurality of lens openings for providing a respective first magnetic flux to the lens openings; and a compensation coil. The compensation coil is arranged between the lens openings for providing a respective second magnetic flux to at least some of the lens openings so as to compensate for an asymmetry of the first magnetic flux. According to a further embodiment, a method for operating a charged particle beam device comprises: Generating a plurality of charged particle beams; guiding each of the charged particle beams through a respective one of a plurality of lens openings of a lens body; generating a current, in a first direction, in a common excitation coil arranged around the plurality of lens openings, thereby providing a respective first magnetic flux to the lens openings; and generating a current, in a second direction opposite to the first direction, in a compensation coil arranged between the lens openings, thereby providing a respective second magnetic flux to at least some of the lens openings and compensating for an asymmetry of the first magnetic flux. The lens system described herein thus allows for close packing of multiple charged particle beams such as electron beams and hence close packing of multiple charged particle beam columns. Accordingly, the lens system allows for the design of Multi-Column Electron Beam Systems with high throughput. The design especially allows for closely-packed electron beams arranged in two dimensions. A two-dimensional arrangement is advantageous for many applications, e.g. if the sample to be scanned and inspected is relatively small. The lens system described herein also allows for a relatively symmetrical focusing field and, hence, for reduced aberrations. Thus, charged particle beams with small spot size and a correspondingly high resolution can be achieved. Embodiments are also directed at apparatuses for carrying out the disclosed methods and including apparatus parts for performing each described method step. These method steps may be performed by way of hardware components, a computer programmed with the appropriate software, by any combination of the two or in any other manner. Furthermore, embodiments according to the invention are also directed at methods by which the described apparatus operates. It includes method steps for carrying out all functions of the apparatus. Reference will now be made in detail to the various embodiments of the invention, one or more examples of which are illustrated in the figures. Within the following description of the drawings, the same reference numbers refer to same components. Generally, only the differences with respect to individual embodiments are described. Each example is provided by way of explanation of the invention and is not meant as a limitation of the invention. For example, features illustrated or described as part of one embodiment can be used on or in conjunction with other embodiments to yield yet a further embodiment. It is intended that the present invention includes such modifications and variations. Without limiting the scope of protection of the present application, in the following the charged particle multi-beam device will exemplarily be referred to as an electron multi-beam device. Thereby, an electron beam device with a plurality of electron beams might especially be an electron beam inspection system. The present invention can still be applied for apparatuses using other sources of charged particles, e.g. ions, for inspection, testing and lithography applications and, in the case of detection devices, other secondary charged particles to obtain a specimen image or the like. With reference to FIGS. 7 and 8, a lens system according to an illustrative example useful for understanding the invention will be described. As seen in the top view of FIG. 7, the lens system has a lens body with four lens openings or lens bores 416 arranged in a 2×2-array, for four electron beam columns. Further, a common excitation coil 420 is arranged around the lens openings 416. Further, as seen in the cross-sectional side view of FIG. 8, the lens body has a first pole piece 412, a second pole piece 414, and respective gaps 418 between the upper and lower pole pieces 412, 414 surrounding the lens openings 416. The pole pieces are made of permalloy, μ-metal or any other magnetic conductive material. If a current is applied to the excitation coil 420, as indicated by the current arrow 421 of FIG. 7, a magnetic flux field B (in short: magnetic field) will be applied to the lens openings 416, as indicated by the magnetic flux lines 422a, 422b, for focusing the electron beam. The lens system of FIGS. 7 and 8 allows for a closely-packed two-dimensional array of lenses. This design, however, has the drawback that the magnetic flux field B (lines 422a, 422b) created by the excitation coil 420 is not rotationally symmetrical and therefore creates aberrations of the electron beam. The asymmetry can be understood in terms of the free energy stored in the magnetic field, U=(H·B)/2. Here, B is the free magnetic flux field due to the excitation coil current, and H=B/(μr·μ0), μr being the magnetic permittivity of the material and μ0 being a constant. The field B takes a spatial configuration that minimizes the free energy U. In the ideal case of infinite magnetic permittivity μr inside the pole pieces 412, 414, the magnetic flux inside the pole pieces 412, 414 would not contribute to the free energy because H=0. Instead, only the portion of the magnetic field traversing the gap between the upper pole piece 412 and the lower pole piece 414 would contribute to the free energy. Hence, assuming that the gap is rotationally symmetrical about each of the lens openings 416, the resulting magnetic flux field B would also be perfectly rotationally symmetrical. In the case of a perfectly rotationally symmetrical flux field, an electron beam traveling on the optical axis (symmetry axis) of the lens would be influenced by the field without introducing astigmatism. However, due to the limited permittivity μr of the magnetic material and due to saturation effects, the magnetic flux inside the pole pieces 412, 414 and their contribution to the free energy U will not be negligible. Hence, the magnetic flux field will be non-symmetrical, with portions close to the coil (flux lines 422a), obtainable at less free energy at a given field strength, having a stronger magnetic field than portions far away from the coil (flux lines 422b). Thus, the magnetic flux field will be stronger at portions of the lens openings 416 near the periphery of the lens body (closer to the coil 420), and weaker at portions of the lens openings 416 near the center of the lens body (farther away from the coil 420). As a result, the individual lens portions will generally have inhomogeneous strengths and an asymmetry that will result in astigmatism and similar unwanted effects of the electron beam. Due to these asymmetries, a magnetic field gradient inside the lens opening creates a dipole effect resulting in parasitic beam deflection. Also, a quadrupole magnetic field component is created (with the poles oriented along the diagonals of FIG. 7), inducing strong astigmatism. Also, higher order multipoles will be created, e.g. a strong hexapole, which deteriorate spot size in high current systems with large bundle diameters and therefore cannot be neglected. For reducing these asymmetries of the magnetic flux field, in an alternative illustrative setup useful for understanding the invention each lens opening could be provided with an individual excitation coil surrounding the respective lens opening (as described with reference to FIG. 3 further below). However, this would result in a large distance between the lens openings, caused by the arrangement of the excitation coils between them. In the following, embodiments of the invention will be described. These embodiments reduce the magnetic flux field asymmetries of the configuration of FIGS. 7 and 8 and their detrimental effects, and additionally allow for a compact setup. FIGS. 1 and 2a show a lens system 1 according to an embodiment of the invention. As seen in the top view of FIG. 1, the lens system 1 has a lens body with four lens openings or lens bores 16 arranged in a 2×2-array, for four electron beam columns. Each of the lens openings 16 is circularly shaped and has a center, and defines an optical axis through the center. Further, a common excitation coil 20 is arranged around the lens openings 16. Further, as seen in the cross-sectional side view of FIG. 2a, taken along plane S1 or S2, the lens body 10 has a first pole piece 12, a second pole piece 14, and respective gaps 18 between the upper and lower pole pieces 12, 14 surrounding the lens openings 16. The lens body 10 (the pole pieces 12, 14) are made of a magnetically conductive material, e.g. a material comprising permalloy or μ-metal. The lens body 10 thus provides a magnetically conductive circuit which confines a magnetic field essentially to the gap region 18 between upper pole piece 12 and lower pole piece 14. This magnetic field then causes an electron beam travelling axially through the lens opening 16 to be focused, as described above with reference to FIGS. 7 and 8. Further possible variations of the lens are described below. Further, the lens system 1 has a compensation coil 30. The compensation coil 30 is arranged between the lens openings 16. Herein, an arrangement between the lens openings is to be understood as follows: At least a part of the compensation coil lies in an area between the lens openings. This area between the lens openings 16 is depicted as area A in FIG. 1. More generally, the area A between the lens openings is defined as the area (polygonal area) between centers of the lens openings, and excluding the area of the lens openings themselves. The compensation coil 30 has the shape of a rectangle with rounded edges, but may have any other circular or non-circular shape. As a general aspect independent of the shown embodiment, the compensation coil 30 has a coil axis parallel to the coil axis of the excitation coil 20. As a further general aspect, no lens openings 16 are inside the compensation coil 30, i.e. all lens openings 16 are outside of the compensation coil 30. Instead, as a further general aspect, the compensation coil is arranged around a magnetic stub 19 (see FIG. 2a). The magnetic stub 19 provides an essentially gapless connection between the upper pole piece 12 and the lower pole piece 14, so that magnetic flux can extend between the upper pole piece 12 and the lower pole piece 14 with low magnetic resistance. As a further general aspect, the interior of the excitation coil 20 and/or the compensation coil 30 has a convex shape. During operation, a current is applied to the excitation coil 20, as indicated by the current arrow 21 of FIG. 1, a magnetic flux field B (first magnetic flux 22) will be applied to the lens openings 16, as illustrated by the magnetic flux lines 22 of FIG. 2b, for focusing the electron beam in the same manner as described with reference to FIGS. 7 and 8. The flux lines 22 of FIG. 2b are a simplified illustration of the more detailed magnetic flux lines 422a, 422b of FIG. 8. Further, a current is applied to the compensation coil 30, as indicated by the current arrow 31 of FIG. 1. This compensation coil current 31 is in a direction opposite to the direction of the excitation coil current 21. The compensation coil current 31 circles around the central stub of magnetic material 19 connecting the upper pole piece 12 and the lower pole piece 14. As can be seen in FIG. 2b, the compensation coil current 31 generates a compensation magnetic flux field 32 (also referred to as second magnetic flux or flux field) for the lens openings 16. The flux lines of the compensation magnetic flux field 32 are closed via the gaps 18 of the lens openings 16 and via the magnetic stub 19. This inner stub of magnetic material 19, together with the compensation coil 30, thus has an important function. If excited appropriately, it creates the same magnetic potential difference in the central part of the multi-bore lens body as the outer coil in the outer part. Hence the radial potential drop across the individual lenses can be compensated for, or reduced. Here, to compensate for an asymmetry is understood to mean that the asymmetry is substantially reduced if not fully eliminated. In particular, the compensation coil allows for eliminating a dipole component of the lens fields by adjusting the compensation current appropriately. Also, higher order multipoles responsible for quadrupole and hexapole astigmatism can be reduced substantially. An appropriate excitation of the compensation coil 30 may be, for example, an excitation to the same number of Ampturns as the excitation coil 20. The compensation flux field 32 compensates for, at least partially, the asymmetry of the first flux field 22 as follows: The first flux field 22 is stronger at portions of the lens openings 16 near the periphery of the lens body 10 (closer to the coil 20), and weaker at portions of the lens openings 16 near the center of the lens body 10 (farther away from the coil 20). In contrast, the compensation flux field 32 has the opposite field distribution, and is weaker at portions of the lens openings 16 near the periphery of the lens body 10 (farther away from the coil 30), and stronger at portions of the lens openings 16 near the center of the lens body 10 (closer to the coil 30). When superimposed, the inhomogeneities—e.g. higher-order magnetic multipole moments—of the total magnetic field (sum of fields 22 and 32) cancel out at least partially. As a result, the total magnetic field has, overall, less inhomogeneities than the field 22, especially a lower dipole moment in a radial direction of coil 20. In other words, an asymmetry of the first magnetic flux (flux field) 22 is compensated for, i.e. reduced, by the second magnetic flux (flux field) 32. Hence, the total field in the gap region 18 interacting with an electron beam traveling through the lens opening 16 is more symmetrical than the field of the comparative example shown in FIGS. 7 and 8. Therefore, the lens allows for reduced beam aberrations and higher obtainable resolution. The lens system of FIG. 1 has some symmetries, notably mirror symmetries with respect to two planes S1 and S2 perpendicular to the drawing plane of FIG. 1. These symmetries reduce some of the higher-order magnetic multipole fields and therefore contribute to a more uniform and aberration-free focusing field. In particular, each lens opening 16 has a magnetic lens field which is symmetrical to plane S1 or S2. Further, as described above, for each lens opening 16, the dipole moment perpendicular to S1 or S2 can also be eliminated by tuning the compensation current appropriately. Thereby, as a general aspect, during operation, a lens field for each of the lens openings 16 has at least one plane of symmetry and can have at least two planes of symmetry, the at least one plane of symmetry containing the respective optical axis. With reference to FIG. 3, the asymmetry compensating effect of the compensation coil 30 can be understood by starting from a further comparative example: In the comparative example, each of the lens openings 16 is provided with an individual coil 50 surrounding the respective lens opening 16 and carrying currents 51. This setup thus produces a highly desirable magnetic flux field, but at a cost: The setup of FIG. 3 requires a large distance between the lens openings 16 due to the excitation coil portions arranged between them, mainly in regions 52. The setup of FIG. 1 produces a magnetic flux that is very similar to the flux produced by the example of FIG. 3, and at the same time disposes of the excitation coil portions arranged between the lens openings, thereby allowing for a more compact setup. Namely, it is crucial to realize that these excitation coil portions in region 52 can be omitted without significantly affecting the magnetic field: in the regions 52, currents from neighboring coils 50 flow in opposite directions, as indicated by the arrows on the respective coils 50 inside regions 52. These neighboring currents largely cancel out and are therefore ineffective for the resulting magnetic flux field. In the setup of FIG. 1, a current, and hence magnetic flux, is obtained which corresponds to the current of FIG. 3, and in which the ineffective and space-consuming portions of the coils 50 in regions 52 are omitted. Namely, the current 21 of coil 20 shown in FIG. 1 corresponds to the current of the outer portions 55 of the coils 50 shown in FIG. 3 (the coil portions towards the outside of regions 52). Further, the current 31 of coil 30 shown in FIG. 1, being arranged between the lens openings 16, corresponds to the current of the inner portions 54 of the coils 50 shown in FIG. 3 (the coil portions arranged inside the regions 52, i.e. between the lens openings). Thus, by comparing FIG. 3 with FIG. 1, it can be understood that as long as the compensation coil 30 is arranged between the lens openings 16, its magnetic flux will compensate for an asymmetry of the magnetic flux of the excitation coil 20. As a general aspect, the compensation coil 30 is arranged such that more than half, or even more than ⅔, or even more than 90% of its effective area is between the lens openings (in area A in FIG. 1), so that a corresponding portion of the compensation flux is generated between the lens openings. Thus, the currents of FIG. 1 correspond to the currents of FIG. 3, with the ineffective parts in regions 52 removed, and the coils re-connected in a more advantageous manner, allowing the current to flow only in the relevant parts. By removing the ineffective parts from the coils and, at the same time, providing a current which improves the homogeneity of the lens fields, the column pitch is reduced and the uniformity of the multi-bore lens field is increased. Further, the wavy current path one would obtain in this manner from FIG. 3 is rectified to a convex shape. Thereby, one gains even more valuable space and can move the lenses closer to each other. As a general aspect, the excitation coil 20 and/or the compensation coil 30 are convex-shaped. Now referring to FIG. 4, a lens system 101 according to a further embodiment of the invention will be described. The lens system 101 comprises lens openings 116, 117, an excitation coil 120 carrying an excitation current 121, and a compensation coil 130 carrying a compensation current 131, and arranged at least partially between the lens openings 116, 117. The lens system 101 of FIG. 4 generally corresponds to the lens system 1 of FIGS. 1 to 2b, with the modifications obvious from the drawings. For example, the compensation coil 130 is also arranged around a magnetic stub, corresponding to stub 19 of FIG. 2a and providing an essentially gapless connection between a first pole piece and a second pole piece. In the following, only the differences with respect to the lens system 1 will be described. In the lens system 101 of FIG. 4, the lens openings 116, 117 are arranged as a two-dimensional array of two rows 102, 103, and five columns. Further, the lens openings 117 of the outermost columns (at the longitudinal ends of the array) are provided as dummy openings. These dummy openings 117 are not used to focus an electron beam. The remaining lens openings 116 arranged in the three inner columns, also referred to as active lens openings 116, are used for focusing charged particle beams. Further, the compensation coil 130 is arranged between the lens openings 116, 117 (i.e. at least partially in the area A between the lens openings 116, 117), and in particular is arranged between the active lens openings 116. Hereby, in contrast to the arrangement of FIG. 1, a small part of the compensation coil 130 is also arranged outside the area A. Due to this part outside the area A, homogeneity of the magnetic flux is increased, and the part does not, in any case, have any detrimental influence on the magnetic flux relevant for the active lens openings 116. Nevertheless, more than 80% of the effective area of the compensation coil 130 is inside the area A, and more than half of the effective area is in the area between the active lens openings 116. With reference to FIG. 5, a lens system 201 according to a further embodiment of the invention will be described. The lens system 201 is very similar to the lens system 101 of FIG. 4, with reference numbers 2xx corresponding to the reference numbers 1xx of FIG. 4. As the main difference with respect to the lens system 101 of FIG. 4, the lens system 201 has three rows 202, 203, 204 of lens openings 216, 217, with two compensation coils 230a, 230b arranged between. For all further details, the description of FIG. 4 applies to FIG. 5 as well. FIG. 5 illustrates that any number of rows and columns can be used. Also, the lens openings 116, 117 may also be arranged in any other manner, e.g. in a hexagonal manner. Also, more than one dummy lens opening 117 or no dummy lens openings may be provided at each side. In a further variation (not shown) of the embodiments of FIGS. 4 and 5, shielding plates may be provided at both ends of the linear lens array. These shielding plates can be combined with the dummy lens openings 117, 217 of FIGS. 4 and 5, respectively, or can be used independently of or instead of the dummy lens openings. The shielding plates have two effects. On the one hand, the influence of the loop of the excitation currents at the end of the linear lens array is shielded. On the other hand, a magnetic neighborhood (periphery) can be provided as if the linear lens array were infinitely long. The structure of the shielding plate and their effects are described in FIGS. 7a and 7b of U.S. Pat. No. 7,576,917 and the description thereof, which are hereby incorporated in the present application. The aspects presented above, namely to provide dummy lens openings and/or to provide shielding plates, can be used independently for all kinds of lens systems. An advantage of the lens systems 101 and 201 of FIGS. 4 and 5 is that these lens systems bring a plurality of rows of 3 or more columns into close proximity, while at the same time providing the benefit of dipole compensation and astigmatism minimization as described above, due to the addition of the compensation coil. The design especially allows for closely-packed electron beams arranged in two dimensions. A two-dimensional arrangement is advantageous for many applications, e.g. if the sample to be scanned and inspected is relatively small, as is the case in wafer mask defect inspection. In this case the mask area is on the order of 100 mm×100 mm and should be scanned simultaneously by at least 4 columns, hence the column pitch in both directions should be about 50 mm. As a general aspect, centers of neighboring lens openings are spaced from one another by less than 100 mm, or less than 75 mm, or even 50 mm or less. Using a miniature compensation coil, a spacing as low as 40 mm can be achieved. With reference to FIG. 6, a further lens system 301 is described. The lens system 301 is very similar to the lens system 1 of FIG. 1, with reference numbers xx of FIG. 1 corresponding to the reference numbers 3xx of FIG. 6. In the following, only differences with respect to FIG. 1 will be described. In the lens system 301 of FIG. 6, nine lens openings 316 are provided. The lens openings 316 are arranged in a 3×3 array. Further, each of the four compensation coils 330a to 330d are arranged between the lens openings. More precisely, each of the coils is arranged between four neighboring lens openings 116, the neighboring lens openings forming a 2×2 sub-array. For example, each of the compensation coils 330a to 330d is arranged around a respective magnetic stub, corresponding to stub 19 of FIG. 2a. The arrangement of the compensation coils 330a to 330d shown in FIG. 6 is particularly advantageous, as can be understood by reasoning analogously to the reasoning described above with reference to FIG. 3. Now, possible further variations of the embodiments will be described. The lens has so far been described as a purely magnetic lens with a gap 18 between the upper pole piece 12 and the lower pole piece 14 (see e.g. FIG. 2a). The pole pieces 12, 14 and the gap 18 can be shaped in any suitable manner, and according to different shapes of lens systems, other arrangements of pole pieces can be realized. Radial gap lenses, for example, have an inner and an outer pole piece. The upper and lower pole piece are distinguished from one another by the gap 18, and while the lens body is usually made from the pole pieces as separate components and then assembled, the lens body, in principle, can also be formed as an integral one-piece component. As a general aspect, a gap in the region of the lens openings separates the first pole piece from the second pole piece. As a further general aspect, the first pole piece and/or the second pole piece are provided as a (respective) single body of magnetic material surrounding the plurality of lens openings. Thereby, magnetic flux is allowed to pass from one of the lens openings to the others with minimal magnetic resistance. Also, the lens can be provided as an electrostatic-magnetic compound lens, with an electrostatic lens provided within the lens opening 16 of FIG. 2a. The electrostatic lens comprises two electrodes arranged symmetrically with respect to the optical axis. The two electrodes are used as an electrostatic immersion lens, whereby the imaging properties can be improved. Such a compound lens is described in FIG. 9 of U.S. Pat. No. 7,576,917 and the description thereof, which are hereby incorporated in the present application. Also, an extra adjustment coil can be arranged around the respective lens openings. Such an extra adjustment corresponds to the coils 50 shown in FIG. 3. However, in contrast to FIG. 3, these adjustment coils are provided in addition to the excitation coil and the compensation coil and therefore need to provide only a weak adjustment field. For this reason, the adjustment coils can be provided with minimal spatial requirements. Also, the excitation coil and the compensation coil can be connected in series so that the (same) current supplied to them is guided in mutually opposite directions, and the number of turns of the compensation coil is adjusted such as to provide a magnetic flux compensating for an asymmetry of the first magnetic flux. The lens system described herein allows for producing a compact multiple charged particle beam device, having multiple charged particle beams with low beam spacing, possibly in two dimensions. Such a beam device has a charged particle beam source for generating a plurality of charged particle beams, and a charged particle beam column. The charged particle beam column comprises, besides elements typically used in beam columns, the lens system as described herein. While the foregoing is directed to embodiments of the invention, other and further embodiments of the invention may be devised without departing from the basic scope thereof, and the scope thereof is determined by the claims that follow.
060699302	abstract	Modified passive containment cooling systems for cooling a reactor core of a boiling water nuclear reactor are described. The passive containment cooling system (PCCS), in one form, includes a vent line coupled to the vacuum breaker. The vent line includes a first end, a second end, and a passage extending between the first and second ends for transporting noncondensibles between the first and second ends. The first end is coupled to PCCS condenser, and the second end is submerged in a suppression pool. A branch extends from an intermediate portion of the vent line and is coupled to the vacuum breaker. The branch includes a first end, a second end, and a passage extending between the first and second ends. The first end of the branch is coupled to the intermediate portion of the vent line so that the branch passage is in communication with the vent line passage. The second end of the branch is coupled to the vacuum breaker so that the branch slopes substantially downwardly from its second end to its first end.
	claims	1. A method of reducing corrosion of a material constituting a nuclear reactor structure comprising:applying a substance capable of generating an excitation current and a noble metal to a surface of a material constituting a nuclear reactor structure in advance; andcontrolling a concentration of oxidizing chemical species and a concentration of reducing chemical species in a nuclear reactor water so that a molar ratio of H2/O2 is less than a value of 2 in which a catalytic reaction to recombine the oxidizing chemical species with the reducing chemical species is not accelerated by the noble metal,wherein the concentration of oxidizing chemical species and the concentration of reducing chemical species in the nuclear reactor water are controlled so that the molar ratio of H2/O2 is less than a value of 2 by injecting hydrogen from a feed water system or a condensate system, and a hydrogen concentration in water in the feed water system or the condensate system is not less than 0.07 ppm and is less than 0.16 ppm. 2. The method of reducing corrosion of a material constituting a nuclear reactor structure according to claim 1, wherein the substance generating the excitation current is at least one of compounds selected from TiO2, ZrO2, ZnO, WO3, PbO, BaTiO3, Bi2O3, SrTiO3, Fe2O3, FeTiO3, KTaO3, MnTiO3, SnO2, and Nb2O5. 3. The method of reducing corrosion of a material constituting a nuclear reactor structure according to claim 1, wherein the noble metal is at least one of elements selected from Pt, Pd, Ir, Rh, Os, and Ru. 4. The method of reducing corrosion of a material constituting a nuclear reactor structure according to claim 1, wherein the substance generating an excitation current and the noble metal is applied by at least one of methods selected from chemical injection, flame spray coating, spraying, plating, and vapor deposition. 5. The method of reducing corrosion of a material constituting a nuclear reactor structure according to claim 4, wherein the noble metal is applied before the application or the adhesion of the substance generating the excitation current. 6. The method of reducing corrosion of a material constituting a nuclear reactor structure according to claim 4, wherein the substance generating the excitation current and the noble metal are applied at the same time or a substance generating the excitation current is applied before the application of the noble metal. 7. The method of reducing corrosion of a material constituting a nuclear reactor structure according to claim 1, wherein an electrochemical corrosion potential of the nuclear reactor structural material composed of a stainless steel is controlled to be −0.23 V vs. SHE or less. 8. The method of reducing corrosion of a material constituting a nuclear reactor structure according to claim 1, wherein an electrochemical corrosion potential of the nuclear reactor structural material composed of a nickel-base alloy is controlled to be −0.1 V vs. SHE or less.
	description	1. Field of the Invention The present invention relates to a particle-beam treatment system in which, in the case where, during particle-beam irradiation, multi-layer conformal irradiation is performed while setting of the shape (leaf position) of a multileaf collimator in an irradiation head is changed, the shape of the multileaf collimator is detected by a leaf-position detection mechanism, and more particularly to a particle-beam treatment system in which, during particle-beam irradiation, the shape of the multileaf collimator can be monitored. 2. Description of the Related Art In a particle-beam treatment system that performs multi-layer conformal irradiation, the dose to be administered to a patient and the distribution thereof are spatially divided and then delivered, so that the dose delivery is made optimal for the shape of a target. The dose distribution divided and delivered in this manner depends on the setting of the irradiation system, such as setting of the shape of the multileaf collimator and the like, and the setting condition of the patient position. In the case where, during particle-beam irradiation, the shape of the multileaf collimator or the patient position is changed from the shape or the setting position that have been decided in a treatment plan, the dose to be administered and the dose distribution differ from those in the treatment plan; therefore, it is required to immediately stop the particle-beam irradiation. For this reason, the monitoring (ascertainment) of the shape (leaf position) of the multileaf collimator and the patient position is an important function for delivering to the patient the dose distribution that has been prescribed in a treatment plan; thus, redundancy and multiplicity are required in the foregoing monitoring. In the case where, in a particle-beam treatment, a static irradiation, which has been practiced since the time before the advent of the multi-layer conformal irradiation, was performed, the shape of the multileaf collimator was ascertained immediately prior to the particle-beam irradiation, based on a light-irradiation field, formed by a light localizer, and X-ray radiographing; then, it could be secured, through detection by use of a detector incorporated in the multileaf collimator, that the forgoing shapes did not change during irradiation. In addition, the conventional monitoring and ascertainment of a patient position were visually performed by shooting a marker written on the surface of the patient body and the projected image of a laser pointer, through a video camera mounted on the ceiling or the sidewall of the treatment room. FIG. 8 is a conventional system block diagram illustrating a method of monitoring and ascertaining the shape of a multileaf collimator and a patient position in the case where a static irradiation, which has been practiced since the time before the advent of the multi-layer conformal irradiation, is performed. FIG. 9 is a configuration diagram illustrating the structure and the system of a typical multileaf collimator. In the conventional static particle-beam treatment, the ascertainment of the shape (leaf position) of a multileaf collimator has been performed by observing, immediately prior to the irradiation, the light-irradiation field formed by a light localizer 11 and the image, of a digital radiograph (DR) 19, which is radiographed, with a X-ray source 13 movably provided on the beam axis, in addition to automatic comparison performed by a leaf-position detection mechanism (e.g., the position is detected by use of an encoder) incorporated in the multileaf collimator. In some cases, a particle-beam flatness monitor is further utilized. In addition, the X-ray source 13 moves on a monitor drive stand 51 that is provided in such a way as to be separated from and to be on a multileaf collimator 14; thus, the X-ray source 13 can be disposed on the beam axis. Explanations therefor will be made with reference to FIGS. 8 and 9. In FIG. 8, reference characters 1, 2, 2a, 2b, 3, 4, 5, and 6 denote an irradiation head, a patient, a diseased site of the patient, a patient-position marker, a particle beam, a dose monitor, wobbler magnets, and a scatterer, respectively; reference characters 7, 8, 9, 10, 11, 12, 13, 14, and 14a denote a ridge filter, a range shifter, an irradiation-system control computer, an irradiation-head control device, alight localizer, a mirror, an X-ray source, a multileaf collimator, and a multileaf-collimator control device, respectively; reference characters 15, 16, 17a, 18, 19, 20, 20a, and 51 denote a patient-monitoring video camera, a video-camera controller, an image monitor, a treatment table, a DR, a laser pointer, a laser beam, and a monitor drive stand, respectively. In FIG. 9, reference characters 14a, 14b, 21, 22, 23, 24, 25, 26, 27, and 28 denote a multileaf collimator control device, a multileaf collimator head unit, a shape of a multileaf collimator, a collimator leaf, a leaf drive mechanism, a mechanical stopper, a leaf-position detector, a leaf drive unit, a signal processing circuit, and a collimator manipulation unit, respectively. The particle beam 3 accelerated by a particle-beam accelerator is led by a beam transport system to the irradiation head 1, limited by the multileaf collimator 14 to a necessary irradiation region, and then irradiated onto the patient 2. The ascertainment of the shape (leaf position) 21 of the multileaf collimator has been performed by, immediately prior to the irradiation, disposing the light localizer 11 and the mirror 12 at the upstream side of the multileaf collimator 14, thereby visually ascertaining the multileaf-collimator shape which is projected onto a plane perpendicular to the traveling direction of the particle beam, in addition to performing automatic comparison and ascertainment of the respective output information items of the position-detection mechanisms 25 for the corresponding collimator leaves and the original setting information in the treatment plan; furthermore, the shape 21, of the multileaf collimator, which is X-rayed with the X-ray digital radiograph (DR) 19 with the X-ray source 13 movably arranged on the beam line, has been ascertained. Additionally, in the conventional monitoring of a patient position, the patient-position marker 2b written on the surface of the patient body and the light marker (e.g., a cross line), which is obtained by projecting, onto the surface of the patient body, the laser beam 20a from the laser pointer 20 provided on the sidewall or the ceiling of the treatment room, have been shot by the video camera 15 that is also disposed on the sidewall or the ceiling of the treatment room and ascertained on the image monitor 17a. In addition, the technical literatures for this field include the following documents: Patent Document 1: Japanese Patent Application Laid-Open No. 1989-274741 Patent Document 2: Japanese Patent Application Laid-Open No. 1990-182273 Patent Document 3: Japanese Patent Application Laid-Open No. 1994-246015 Patent Document 4: U.S. Pat. No. 4,882,741 (corresponding Japanese publication: Japanese Patent Application Laid-Open No. 1989-146564) Patent Document 5: GB Patent No. 2,211,710A (corresponding Japanese publication: Japanese Patent Application Laid-Open No. 1989-146564) Non-patent Document 1: PHYSICS Annual Report 2001-2002, 4. Improvement of the HIMAC Treatment System with the Layer-Stacking Conformal Irradiation Method, Nobuyuki Kanematsu, et al. In a conventional static irradiation method utilizing a particle-beam treatment system, as illustrated in FIG. 8, the monitoring and the ascertainment of the shape of a multileaf collimator has been performed by visually ascertaining, immediately prior to irradiation, the light-irradiation field formed by the light localizer 11 and the mirror 12 that are disposed at the upstream side of the multileaf collimator and observing the image, of the digital radiograph (DR) 19, which is radiographed with the X-ray source 13 disposed on the beam axis, in addition to automatic comparison performed by leaf-position detection mechanisms incorporated in the multileaf collimator; however, because, the ascertainment work in a treatment room is involved, the foregoing methods, except for the method of ascertainment performed by the leaf-position detection mechanisms incorporated in the multileaf collimator, cannot be applied to a dynamic irradiation method such as the multi-layer conformal irradiation in which the setting for a multileaf collimator is changed during irradiation. Moreover, because the space above the monitor drive stand 51 is limited, it is difficult to provide additional reinforcement. Still moreover, the conventional monitoring of a patient position has been performed by shooting the marker 2b written on the surface of the patient body and the image of the laser pointer 20, through the video camera 15 mounted on the ceiling or the sidewall of the treatment room, thereby carrying out visual ascertainment on the image monitor 17a; however, in some cases, the monitoring subject has not securely been captured, depending on the irradiation arrangement. It is possible in principle to make the leaf-position detector 25 multiple and redundantly ascertain the shape 21 of the multileaf collimator, in order to ascertain the setting condition during particle-beam irradiation; however, the multileaf collimator has a great number of drive elements and the space where the newly added leaf-position detectors and signal transmission paths are mounted in the multileaf collimator head unit is limited, whereby many difficult issues exist. Furthermore, in contrast to the conventional static irradiation method, in the multi-layer conformal irradiation, a dose is administered to a treatment target that is divided into a plurality of irradiation units; therefore, a change in the patient position during particle-beam irradiation forms high-dose and low-dose regions in the dose distribution. The foregoing issue cannot be addressed with the setting margin, for the target, which is set in accordance with the conventional static irradiation method and in consideration of the change of the body position; thus, nothing but improvement in the fixing method for the fixing device and the like and more stringent monitoring of body-position change can serve as measures for the foregoing issue. In the conventional monitoring through the video camera 15 disposed on the sidewall or the ceiling of a treatment room, a dead angle or the like, which may be caused depending on the irradiation arrangement, may make it difficult to securely monitor the body position that is subject to irradiation. The present invention has been implemented in order to solve the foregoing problems; the objective of the present invention is to obtain a particle-beam treatment system in which monitoring of the shape of a multileaf collimator can be performed even during particle-beam irradiation, and even in the case where, during particle-beam irradiation, the shape of the multileaf collimator is changed, redundant monitoring can be performed, in addition to monitoring of the shape of the multileaf collimator, through leaf-position detection mechanisms. A particle-beam treatment system, according to the present invention, in which, in the case where, during particle-beam irradiation, multi-layer conformal irradiation is performed while setting of the shape of a multileaf collimator in an irradiation head is changed, the shape of the multileaf collimator is detected by a leaf-position detection mechanism, is provided with an optical shape-monitoring unit mounted attachably and detachably in the snout portion at the downstream side of the multileaf collimator, the optical shape-monitoring unit having a shape-monitoring mirror, opposing the multileaf collimator, for monitoring the shape of the multileaf collimator, a video camera for shooting the multileaf-collimator shape reflected by the shape-monitoring mirror, and an image monitor for displaying an image of the video camera that shoots the shape of the multileaf collimator; the particle-beam treatment system enables the shape of the multileaf collimator to be monitored during particle-beam irradiation. According to a particle-beam treatment system of the present invention, monitoring of the shape of a multileaf collimator can be performed even during particle-beam irradiation, and even in the case where, during particle-beam irradiation, the shape of the multileaf collimator is changed, redundant monitoring can be performed, in addition to monitoring, through a leaf-position detection mechanism, of the shape of the multileaf collimator. A particle-beam treatment system, according to the present invention, in which, in the case where, during particle-beam irradiation, multi-layer conformal irradiation is performed while setting of the shape of a multileaf collimator in an irradiation head is changed, the shape of the multileaf collimator is detected by leaf-position detection mechanisms, is provided with an optical shape-monitoring unit mounted attachably and detachably in the snout portion at the downstream side of the multileaf collimator, the optical shape-monitoring unit having a shape-monitoring mirror, opposing the multileaf collimator, for monitoring the shape of the multileaf collimator, a video camera for shooting the multileaf-collimator shape reflected by the shape-monitoring mirror, and a comparison means for comparing an image of the video camera with multileaf-collimator-shape information in a treatment plan and determining whether or not the comparison result is appropriate; the particle-beam treatment system performs particle-beam irradiation or particle-beam cutoff processing, depending on whether or not the comparison result is appropriate. According to a particle-beam treatment system of the present invention, an image of the video camera that shoots the multileaf-collimator shape reflected by the shape-monitoring mirror and multileaf-collimator-shape information in a treatment plan are compared, and particle-beam irradiation or particle-beam cutoff processing is performed depending on whether or not the comparison result is appropriate; therefore, inappropriate particle-beam irradiation can be avoided. The foregoing and other objects, features, aspects and advantages of the present invention will become more apparent from the following detailed description of the present invention when taken in conjunction with the accompanying drawings. FIG. 1 is a system block diagram illustrating a particle-beam treatment system according to Embodiment 1 of the present invention. FIG. 1 illustrates principle constituent elements in a configuration in which an optical shape-monitoring unit, which includes a multileaf-collimator-shape monitoring mirror utilized in the multi-layer conformal irradiation, is mounted in an irradiation head. FIG. 2 is a system block diagram illustrating a multileaf collimator head unit and the control system therefor according to Embodiment 1. FIG. 3 is a flowchart representing a collimator-shape monitoring flow according to Embodiment 1. Meanwhile, the structure and the leaf-position detection mechanism of the typical multileaf collimator explained with reference to FIG. 9 can directly be applied to Embodiment 1; in the present specification, the same reference characters in the figures denote identical or corresponding parts; therefore, explanations therefor may be omitted. In FIG. 1, reference characters 1, 2, 2a, 2b, 3, 4, 5, and 6 denote an irradiation head of a particle-beam treatment system, a patient, a diseased site of the patient, a patient-position marker, a particle beam, a dose monitor, wobbler magnets, and a scatterer such as Pb, respectively; reference characters 7, 8, 9, 10, 13, and 14 denote a ridge filter such as Al, a range shifter such as an acrylate resin, an irradiation-system control computer, an irradiation-head control device, an X-ray source, and a multileaf collimator, respectively; reference characters 15a, 16, 17a, 17b, 31, 32a, 33, and 34 denote a video camera, a video-camera controller, an image monitor, a keyboard, an optical shape-monitoring unit, a shape-monitoring mirror, a video signal processing circuit, and an image processing computer, respectively. In FIG. 2, reference characters 14a, 22, 31a, 31b, 31c, 35a, 36, 37 and 38 denote a multileaf collimator control device, collimator leaves, a shape-monitoring-mirror mounting stand, a compensation filter mounting stand, a patient collimator mounting stand, collimator-shape information in a treatment plan, an image comparison means for a multileaf-collimator shape, an irradiation OK signal, and an irradiation stop signal or an irradiation prohibition signal, respectively. In FIG. 3, reference characters 35a, 39, 40, 41, and 42 denote collimator-shape information in a treatment plan, a collimator-shape raw image, image processing, an image corresponding to a direct view after the image processing, and a set of comparison-image data pieces. Next, the operation of the particle-beam treatment system will be explained. In FIG. 1, the particle beam 3 accelerated by a particle-beam accelerator in the particle-beam treatment system enters the dose monitor 4 in irradiation head 1, by way of a beam transport system; in the dose monitor 4, the irradiation dose is counted. The wobbler magnets 5 and the scatterer 6 form the particle beam 3 whose irradiation field is enlarged. After exiting from the scatterer 6, the particle beam 3 passes through the ridge filter 7; the Bragg peak is enlarged in the depth direction and a homogeneous dose region is formed; then the range is adjusted by the range shifter 8. In the multi-layer conformal irradiation, a spatial dose delivery is administered in such a way as to be divided in the depth direction; upon the initial irradiation, the wobbler magnets 5, the range shifter 8, and the multileaf collimator 14 (the multileaf-collimator shape) are set in accordance with the dose delivery at the deepest portion, and then the particle beam 3 is irradiated onto the diseased site 2a. After the irradiation to the deepest portion ends, the range shifter 8 automatically adjusts the range to extend up to the position that is shallower by a depth corresponding to the peak width than the deepest portion, and the settings for the wobbler magnets 5 and the multileaf collimator 14 are also changed; then, irradiation is carried out. Thereafter, similarly, the range shifter 8 adjusts the range and the settings for the wobbler magnets 5 and the multileaf collimator 14 are changed, so that a dose optimized as a whole for the shape of the diseased site 2a is delivered. In the multi-layer conformal irradiation for a particle-beam treatment, in order to perform a high-accuracy particle-beam treatment as described above, it is necessary to ascertain and monitor, in each of the irradiation steps, the setting for the shape of the multileaf collimator. Accordingly, by mounting the detachable and attachable optical shape-monitoring unit 31 and the video camera 15a in the snout portion at the downstream side of the irradiation head 1, particularly in a case situated at the downstream side of the multileaf collimator and disposing in the optical shape-monitoring unit 31 the shape-monitoring mirror 32a, for monitoring the shape of the multileaf collimator 14, slanted on the beam axis, the video camera 15a shoots the reflected image of the multileaf collimator. On this occasion, the downstream-side shape of the multileaf collimator appears in the shape-monitoring mirror 32a. The image distortion (aspect ratio and the like), caused by the arrangement of the shooting system, e.g., the shape-monitoring mirror 32a, of the multileaf-collimator-shape raw image 39 shot by the video camera is corrected by the video signal processing circuit 33, so that a multileaf-collimator-shape direct-view-corresponding image 41, which is equivalent to an image as directly viewed in the beam axis, is generated. The image processing computer 34 extracts the outline of the set shape of the multileaf collimator from the direct-view-corresponding image 41, by use of image discrimination processing such as the binarization method (1 and 0 or white and black) and displays the outline on the image monitor 17a. Furthermore, the image processing computer 34 performs the comparison 36 between the multileaf-collimator-shape information 35a in a treatment plan and the direct-view-corresponding image 41 and then outputs the irradiation OK signal 37 or the irradiation stop signal 38, thereby making the irradiation-system control computer 9 interlock-control the emission condition of the particle-beam irradiation, so that erroneous dose administration due to erroneous setting of the multileaf collimator is avoided. As described above, by introducing the optical image shooting through the optical shape-monitoring unit 31 and the image comparison, not only can the monitoring of setting for the leaf position be performed through the leaf-position detector 25 (refer to FIG. 9) incorporated in the multileaf collimator 14, but also can the redundant, multiple ascertainment and monitoring of the multileaf-collimator shape 21 be performed even during irradiation. In the multi-layer conformal irradiation for particle-beam irradiation, by mounting the attachable and detachable optical shape-monitoring unit 31 in the snout portion at the downstream side of the multileaf collimator, the ascertainment and the monitoring of the multileaf-collimator shape is performed. In order to suppress the range loss of the particle beam and the increase in the scattered components that are caused through mounting of the optical shape-monitoring unit 31, the shape-monitoring mirror 32a is formed by depositing aluminum on a polyimide film and disposed slanted toward the side closer to the plane perpendicular to the beam axis. The image distortion due to the shooting system, e.g., slanted disposal of the shape-monitoring mirror, is corrected by the video signal processing circuit 33. In addition, when a conventional static particle-beam irradiation is performed in which the multileaf collimator does not operate during irradiation, by removing the optical shape-monitoring unit 31, the radiation damage to the shape-monitoring mirror 32a or the video camera 15a can be reduced. In order to suppress the dose distribution in the irradiation field defined by the multileaf collimator from being deteriorated due to the increase in a drift distance, the gradient θ, from the traveling direction of the particle beam, of the shape-monitoring mirror 32a opposing the multileaf collimator is made to be close to 90°, rather than 45°. As a result, the space, in the traveling direction of the particle beam, occupied by the optical shape-monitoring unit can be reduced. By correcting the aspect-ratio distortion in an image of the video camera that shoots the multileaf-collimator shape reflected by the shape-monitoring mirror, the video-camera image is displayed on the image monitor, as an image equivalent to the image of the multileaf-collimator shape as directly viewed in the beam-axis direction. After the image distortion and the aspect ratio of the reflected image are corrected through image processing, the image of the multileaf-collimator shape shot by the video camera is compared with the setting, for the multileaf-collimator shape, which has been planed in a treatment-plan apparatus; in the case where the result of the comparison is inappropriate, the irradiation is interrupted; in the case where the result of the comparison is appropriate, the irradiation is carried on or made stand-by. The foregoing operation enables, even during irradiation, redundant and high-reliability ascertainment and monitoring of the shape of a multileaf collimator, without interrupting particle-beam irradiation; in consequence, a particle-beam treatment system that reduces the probability of erroneous irradiation and enables high-accuracy particle-beam treatment can be configured. In Embodiment 1, the shape of a multileaf collimator is shot by means of the shape-monitoring mirror 32a, and the video camera 15a; however, the monitoring of a patient position, which is symmetric with the multileaf collimator 14 with respect to the mirror plane, can be performed with a similar shooting system. In other words, by making the monitoring mirror a two-side mirror and utilizing the respective sides as the shape-monitoring mirror 32a and a patient-position monitoring mirror 32b, the multileaf-collimator shape and the patient position can be monitored and ascertained concurrently or in a time-division fashion. FIG. 4 is a system block diagram illustrating a multileaf collimator head unit and the control system therefor according to Embodiment 2. FIG. 5 is a flowchart representing a patient-position monitoring flow according to Embodiment 2; In Embodiment 2, the steps in the flowchart, in FIG. 3, representing a collimator-shape monitoring flow are performed concurrently or in a time-division fashion. In addition, in the case the steps are concurrently performed, two systems of required apparatuses may be utilized. In Embodiment 2, the one side of a monitoring mirror serves as the shape-monitoring mirror 32a and the other side serves as the patient-position monitoring mirror 32b; an optical shape-monitoring unit is included in the optical shape-monitoring unit 31 mounted attachably and detachably in the snout portion at the downstream side of the multileaf collimator. FIGS. 4 and 5 will mainly be explained. Reference characters 32a and 15a denote a shape-monitoring mirror and a video camera therefor, respectively, and reference characters 32b and 15b denote a patient-position monitoring mirror and a video camera therefor, respectively. Reference numeral 35 denotes reference data pieces, for the comparison, which are multileaf-collimator-shape information in a treatment plan and patient-position information. In FIG. 5, reference character 35b denotes reference data, for the comparison, which is the patient-position information. Reference numerals 42, 43, 44 and 45 denote a set of image data pieces for the comparison, a patient-position raw image, a patient-position marker, and a patient-position direct-view-corresponding image, respectively. In FIG. 4, the shape-monitoring mirror 32a and the patient-position monitoring mirror 32b are formed, as a two-side mirror, by depositing aluminum on a single polyimide film. Next, the operation of the particle-beam treatment system will be explained. The image reflected by the shape-monitoring mirror 32a that monitors the multileaf-collimator shape and the image reflected by the patient-position monitoring mirror 32b that opposes a patient and monitors the patient position are shot by the video cameras 15a and 15b, respectively; the distortions in the foregoing reflected images are corrected by the video signal processing circuit 33. In other words, the patient-position raw image 43, shot by the video camera 15b for monitoring the patient position, which is obtained by shooting, in the beam-irradiation direction, a patient as an irradiation subject receives image processing 40 in the video signal processing circuit 33 and corrected into the patient-position direct-view-corresponding image 45. The corrected patient-position direct-view-corresponding image 45 is displayed on the image monitor 17a, along with the corrected multileaf-collimator-shape direct-view-corresponding image 41. On the image monitor 17a, for example, by dividing the display screen into two portions, the patient-position direct-view-corresponding image 45 and the multileaf-collimator-shape direct-view-corresponding image 41 are displayed. The image processing computer 34 records, in the patient-position direct-view-corresponding image 45 prior to irradiation, image information on an interested region, including characteristic points, e.g., a patient-position marker 44, as the patient-position information (reference data) 35b in a treatment plan, for the image comparison unit (comparison means) 36. Thereafter, a set of data pieces 42 for the comparison between the patient-position direct-view-corresponding image 45 shot during particle-beam irradiation and the patient-position information (reference data) 35b in a treatment plan is utilized for the image comparison 36 according to the subtraction method or the like; the irradiation OK signal 37 or the irradiation stop signal 38 is outputted so as to make the irradiation-system control computer 9 interlock-control the emission condition of the particle-beam irradiation, so that erroneous dose administration due to a change in the patient position is avoided. The combination of the operation according to the flow in FIG. 5 and the operation according to the flow in FIG. 3 enables the monitoring and ascertainment of the set shape of the multileaf collimator and the patient position; therefore, during the multi-layer conformal irradiation, erroneous irradiation due to erroneous setting of a multileaf collimator and a change in a patient position can be avoided, whereby a high-reliability particle-beam treatment can be performed. Moreover, no dead angle occurs for the patient-monitoring image shot by the optical patient-position monitoring unit; therefore, by setting an appropriate monitoring marker on the surface of a patient body, a high-accuracy patient-monitoring means can be provided. The foregoing operation can provide, even during irradiation, a redundant and high-reliability means for accurately ascertaining and monitoring a patient position, without interrupting particle-beam irradiation; in consequence, a particle-beam treatment system that reduces the probability of erroneous irradiation and enables high-accuracy particle-beam treatment can be configured. In order to suppress the dose distribution in the irradiation field defined by the multileaf collimator from being deteriorated due to the increase in a drift distance, the gradient θ, from the traveling direction of the particle beam, of the patient-position monitoring mirror 32b opposing the patient is made to be close to 90°, rather than 45°. As a result, the space, in the traveling direction of the particle beam, occupied by the optical patient-position monitoring unit can be reduced. Moreover, by correcting through image processing the aspect-ratio distortion in an image of the video camera that shoots the patient position reflected by the patient-position monitoring mirror 32b, the video-camera image may preferably be displayed on the image monitor, as an image equivalent to the image of the patient-position as directly viewed in the beam-axis direction. Still moreover, by providing the image processing computer 34 that extracts, by use of the binarization method (1 and 0 or white or black), the respective outlines or characteristic points from the signal for the image of the video camera that shoots the multileaf-collimator shape and the signal for the image of the video camera that shoots the patient position, the multileaf-collimator shape and the patient position may be enabled to be monitored based on the extracted outlines or characteristic points of the monitoring subject. FIG. 6 is a system block diagram illustrating a multileaf collimator head unit and the control system therefor according to Embodiment 3. In the case where a compensation filter (for compensating the particle-beam distribution) is mounted in the compensation filter mounting stand 31b of the optical shape-monitoring unit 31, or in the case where a patient collimator is mounted in the patient collimator mounting stand 31c, it is impossible or difficult, with Embodiment 2, to monitor and ascertain the patient position. In order to solve the foregoing problem, a patient-position monitoring mirror mounting stand 31d on which the patient-position monitoring mirror 32 is disposed is mounted on the front end of the optical shape-monitoring unit 31 so as to monitor and ascertain the patient position. In this case, the patient-position monitoring mirror 32c and the patient-position monitoring mirror mounting stand 31d configure an optical patient-position monitoring unit; i.e., the optical patient-position monitoring unit is mounted at the downstream side of the multileaf collimator. Additionally, the mounting position of the video camera 15b is moved to a position where the video camera 15b can shoot the patient-position image reflected by the patient-position monitoring mirror 32c. As a result, even in the case where the compensation filter or the patient collimator is mounted, the patient position as well as the multileaf-collimator shape can be ascertained; therefore, during the multi-layer conformal irradiation, erroneous irradiation due to erroneous setting of the multileaf-collimator shape or a change in a patient position can be avoided, whereby a high-reliability particle-beam treatment can be performed. FIG. 7 is a system block diagram illustrating a multileaf collimator head unit and the control system therefor according to Embodiment 4. In the multi-layer conformal irradiation, the image processing computer 34 performs the comparison 36 between the multileaf-collimator-shape information or the patient-position information that is image-corrected by the video signal processing circuit 33 and reference data 35 that is the multileaf-collimator-shape information or the patient-position information in the treatment plan, respectively. In the case where, after the start of particle-beam irradiation, the comparison result is inappropriate, the irradiation stop signal 38 is outputted, and then the particle-beam irradiation is immediately cut off. The case that is relatively likely to occur and in which, due to an inappropriate comparison result, particle-beam irradiation is cut off is exemplified by a case where the patient position changes. In this situation, in the case where, after the cutoff, the multi-layer conformal irradiation is resumed, high-dose and low-dose regions are formed unless the patient position prior to the cutoff is reproduced. Accordingly, patient-position information 29a upon the start of irradiation is stored in an irradiation-condition storage medium 30 so as to be able to be referred to, as reference information, when the normality of setting for the patient position is ascertained upon the resumption of irradiation. In addition, irradiation-head device setting information 29a upon the start of irradiation is also stored in the irradiation-condition storage medium 30. In this case, irradiation-head devices to be set include the wobbler magnets 5, the range shifter 8, and the multileaf collimator 14; concurrently with counting through the dose monitor 4, the irradiation-system control computer 9 stores the setting conditions of the devices in the irradiation-condition storage medium 30. After the reproducibility of the patient position is ascertained through referring to the foregoing information pieces and performing re-positioning by means of X-rays, the device conditions are set based on the irradiation-head device setting information 29b that has been stored during the interruption of irradiation and then the irradiation is resumed. Additionally, in order to comprehend the effects of erroneous setting for the devices of the irradiation head and a change in the patient position, during the interruption of irradiation, the patient-position information and the irradiation-head device setting information 29b during the cutoff of irradiation are stored in the irradiation-condition storage medium 30. As described above, by securely storing the irradiation-head device setting information and the patient-position information during the multi-layer conformal irradiation, the conditions of the irradiation-head device setting and the patient position during the interruption of irradiation can be comprehended; therefore, in the case where the resumption of irradiation is possible, by resuming irradiation at the time of the resumption of irradiation, the planned irradiation can be compensated, whereby high-reliability particle-beam treatment can be performed. While the presently preferred embodiments of the present invention have been shown and described. It is to be understood that these disclosures are for the purpose of illustration and that various changes and modifications may be made without departing from the scope of the invention as set forth in the appended claims.
	claims	1. A differential evacuation system comprising:a light generation chamber that generates EUV (Extreme Ultra Violet) light;an illumination optical chamber in which optical processing is performed by using the EUV light;a light converging mirror arranged in the light generation chamber and converging the EUV light thus generated;a chamber connecting passage that connects together the light generation chamber and the illumination optical chamber to allow the EUV light to pass therethrough;an optical element arranged in the illumination optical chamber and reflecting the EUV light that has passed through the chamber connecting passage;a flow path constricting portion provided in the chamber connecting passage and obstructing a gas to pass therethrough; said flow path constriction portion with a smallest inner diameter and being increased in inner diameter in a conical tube shape at portions thereof that are at opposite sides, respectively, of said flow path constriction portion; andone or a plurality of vacuum pumps disposed in a position closer to the light generation chamber than the flow path constricting portion and evacuating the chamber connecting passage. 2. The differential evacuation system of claim 1, further comprising one or a plurality of pipes connected to a position of the chamber connecting passage that is closer to the light generation chamber and the flow path constricting portion,wherein the one or the plurality of vacuum pumps are connected to the one or the plurality of pipes. 3. The differential evacuation system of claim 1, further comprising a gas introducing device that feeds a gas onto a surface of the light converging mirror. 4. The differential evacuation system of claim 1,wherein the gas fed onto the surface of the light converging mirror contains He, Ne, or Ar. 5. The differential evacuation system of claim 1,wherein the gas fed onto the surface of the light converging mirror contains hydrogen. 6. The differential evacuation system of claim 1, further comprising a magnetic field generation device that generates a magnetic field around the light converging mirror. 7. The differential evacuation system of claim 1,wherein the light generation chamber is provided with a laser transmitting window for introducing laser light into the light generation chamber. 8. A differential evacuation system comprising:a light generation chamber that generates light;an illumination optical chamber in which optical processing is performed by using the light generated in said light generation chamber; anda chamber connecting passage serving as a light passage that connects together said light generation chamber and illumination optical chamber to guide the light generated in the light generation chamber into the illumination optical chamber;said chamber connecting passage having a flow path constricting portion with a smallest inner diameter and being increased in inner diameter in a conical tube shape at portions thereof that are at opposite sides, respectively, of said flow path constricting portion;wherein one or a plurality of vacuum pumps are attached to each of the portions that are at opposite sides, respectively, of said flow path constricting portion. 9. The differential evacuation system of claim 8, wherein the portions of said chamber connecting passage that are at opposite sides, respectively, of the flow path constricting portion are provided with enlarged-diameter parts, respectively, said one or plurality of vacuum pumps being attached to each of the enlarged-diameter parts. 10. The differential evacuation system of claim 9, wherein said plurality of vacuum pumps are attached to a side of each of said enlarged-diameter parts. 11. The differential evacuation system of claim 9, wherein said plurality of vacuum pumps are attached to an outer peripheral surface of each of said enlarged-diameter parts. 12. The differential evacuation system of claim 8, wherein one or a plurality of pipes are connected to each of the portions of said chamber connecting passage that are at opposite sides, respectively, of the flow path constricting portion, said vacuum pumps being attached to the pipes, respectively. 13. A differential evacuation system comprising:a light generation chamber that generates light;an illumination optical chamber in which optical processing is performed by using the light generated in said light generation chamber; anda chamber connecting passage serving as a light passage that connects together said light generation chamber and illumination optical chamber to guide the light generated in the light generation chamber into the illumination optical chamber;said chamber connecting passage having a flow path constricting portion with a smallest inner diameter and being increased in inner diameter from the flow path constricting portion toward at least one of the light generation chamber and the illumination optical chamber;wherein an enlarged-diameter part is provided at an intermediate position of the chamber connecting passage, and one or a plurality of vacuum pumps are attached to the enlarged-diameter part. 14. The differential evacuation system of claim 13, wherein said plurality of vacuum pumps are attached to a side of said enlarged-diameter part. 15. The differential evacuation system of claim 13, wherein said plurality of vacuum pumps are attached to an outer peripheral surface of said enlarged-diameter part.
	abstract	A device is provided for mitigating vibration in a component of a nuclear reactor by removing vibration energy. To reduce vibration in the component, a device operatively connected to the component and including a magnet may be actuated within a conductive cylinder. This actuation may generate one or more eddy currents providing a damping function for removing vibration energy from the component, so as to alter vibration characteristics of the component.
056489951	claims	1. A method of manufacturing a tube for constituting at least an outer part of a sheath of a nuclear fuel rod or a guide tube of a nuclear fuel assembly, comprising the steps of: (a) forming a bar of an alloy of zirconium containing 50 ppm to 250 ppm iron, 0.8% to 1.3% by weight niobium, less than 1600 ppm oxygen, less than 200 ppm carbon, and less than 120 ppm silicon; (b) heating the bar to between 1000.degree. C. and 1200.degree. C. and quenching the bar in water; (c) extruding a blank from said bar after heating said bar to the range 600.degree. C. to 800.degree. C.; (d) cold rolling said blank in at least four passes to obtain a tube, with intermediate heat treatments in the range 560.degree. C. to 620.degree. C.; and (e) performing a final heat treatment in the range 560.degree. C. to 620.degree. C., 2. A method according to claim 1, wherein step (d) includes four or five cold rolling passes starting from the extruded blank. 3. A method according to claim 1, further comprising a step of carrying out a heat treatment at a temperature in the range 560.degree. C. to 620.degree. C. after step (c). 4. A method according to claim 1, wherein said intermediate heat treatments are performed for a period of two hours to four hours, at a set temperature in the range 565.degree. C. to 605.degree. C. 5. A method according to claim 1, wherein the final heat treatment is performed for a period of two to four hours at a temperature lying in the range 565.degree. C. to 605.degree. C. 6. A method according to claim 5, wherein the final heat treatment is a temperature of about 580.degree. C. 7. A method according to claim 1, wherein the iron content is about 150 ppm. 8. A method according to claim 1, wherein the oxygen content lies in the range 1000 ppm to 1600 ppm. 9. A sheathing tube for a fuel assembly of a PWR made of a zirconium-based alloy in fully recrystallized condition, having 50 ppm to 250 ppm iron, 0.8% to 1.3% by weight niobium, 1000 ppm to 1600 ppm oxygen, less than 200 ppm carbon, less than 120 ppm silicon, the balance being zirconium and unavoidable impurities.
	claims	1. A method for distributing a process variable using statistically-correct spatial interpolation, the method comprising:configuring a circuit simulation tool to perform computerized modeling of a chip to be fabricated using the process variable in a process step;forming, using the circuit simulation tool, an array of equilateral triangles in a planar coordinate frame on a region of the chip;assigning, using the circuit simulation tool, a numerical value of the process variable at each vertex of the array of equilateral triangles;defining, using the circuit simulation tool, a plurality of test points at respective spatial locations in the planar coordinate frame that are bounded by the array of equilateral triangles;distributing, using the circuit simulation tool, a numerical value of the process variable at each of the test points by spatial interpolation from one or more of the numerical values of the process variable assigned at each vertex of the array of equilateral triangles; andadjusting, using the circuit simulation tool, the numerical value of the process variable distributed at each of the test points with a respective correction factor. 2. The method of claim 1 wherein forming, using the circuit simulation tool, the array of equilateral triangles further comprises:defining, using the circuit simulation tool, a local spatial correlation that equals zero at a minimum distance between test points; andassigning, using the circuit simulation tool, a length to each side of the equilateral triangles to be equal to the minimum distance for the local spatial correlation. 3. The method of claim 1 wherein assigning, using the circuit simulation tool, the numerical value of the process variable at each vertex of the equilateral triangles further comprises:providing, using the circuit simulation tool, a global distribution function for the process variable; andspecifying, using the circuit simulation tool, the numerical value at each vertex from the global distribution function. 4. The method of claim 3 wherein the global distribution function has an associated standard deviation, and each respective correction factor is selected to reduce a difference between a local standard deviation of the numerical values of the process variable at the test points and the standard deviation of the global distribution function. 5. The method of claim 1 wherein the numerical value at each vertex of the array of equilateral triangles is assigned from a global distribution function having a mean, and further comprising:randomly selecting, using the circuit simulation tool, a numerical value of the process variable chosen from the global distribution function with the mean set equal to zero;determining, using the circuit simulation tool, a respective multiplicative factor based upon the spatial location of each of the test points relative to respective spatial locations in the planar coordinate frame for one or more of the vertices of the array of equilateral triangles; andmultiplying, using the circuit simulation tool, the respective multiplicative factor and the randomly selected numerical value of the process variable to determine the respective correction factor used to adjust the numerical value of the process variable distributed at each of the test points. 6. The method of claim 1 further comprising:calculating, using the circuit simulation tool, the correction factor for the numerical value of the process variable at each of the test points with a mathematical algorithm. 7. The method of claim 1 wherein adjusting, using the circuit simulation tool, the numerical value of the process variable at each of the test points with the respective correction factor further comprises:adding, using the circuit simulation tool, a numerical offset as the respective correction factor to the numerical value of the process variable distributed by the spatial interpolation at each of the test points. 8. The method of claim 1 wherein the process variable is a height or a thickness for metallization in a wiring level of a multi-level interconnect of the chip.

Subsets and Splits

No community queries yet

The top public SQL queries from the community will appear here once available.