Datasets:

arcee-ai
/

nuclear_patents

Dataset card Data Studio Files Files and versions Community

Split (2)

train · 33.5k rows

patent_number stringlengths 0 9	section stringclasses 4 values	raw_text stringlengths 0 954k
	abstract	The present invention concerns a hood for handling and confinement of at least two objects of slender shape, including an external enclosure and internal enclosures inside the external enclosure and at least one motor fixed above an internal enclosure and inside a barrel, the motor(s) being adapted to rotate the screw of the screw-nut mechanism of each internal enclosure and therefore the nut over a stroke A, and first and second mechanical control means, arranged in part above the cover of the external enclosure, respectively for manually guiding the internal enclosures in translation over a stroke A0 and manually pivoting the barrel in order to bring a holding member of one of the internal enclosures opposite the opening in the bottom of the external enclosure. Application to the handling and confinement of nuclear material sample holders.
	abstract	A protector bottom grid for a nuclear fuel assembly that includes three laterally staggered and horizontally oriented protrusions that extend into the fuel rod cell of a support grid below a vertically oriented spring. The three staggered protrusions extend into the cell a distance that maintains a space between the protrusions and the fuel rod. The vertically oriented spring biases the fuel rod against a dimple extending from the opposite cell wall that is at an elevation just above the spring. The protrusions below the spring trap incoming debris in the area of the fuel rod end cap and protect the fuel rod cladding from fretting.
	claims	1. A fuel assembly for a boiling water reactor, comprising:inner webs disposed in a crosswise fashion;outer webs each having an inner side and surrounding said inner webs in a form of a frame, said inner webs and said outer webs together defining and forming a number of spacers having cells for receiving fuel rods, each of said spacers containing at least one guide device disposed in a respective outer web, said guide device having a flow opening formed therein and a guide element, said guide element as viewed in a flow direction of a coolant, being disposed upstream of said flow opening, and projecting from said inner side of said respective outer web, said guide element constricting a flow channel formed by a surface of a respective fuel rod and said inner and outer webs surrounding the respective fuel rod; andinwardly projecting clamping springs each for cooperating with a respective one of the fuel rods and attached to said outer webs;said guide device disposed laterally offset from a respective inwardly projecting clamping spring at a region of said respective outer web extending between said respective inwardly projecting clamping spring and a respective inner web, said guide device disposed completely in said region. 2. The spacer according to claim 1, wherein said guide element is a flow vane integrally formed on said respective outer web and forms an acute angle α, opening in the flow direction, with a flat plane of said respective outer web. 3. The spacer according to claim 2, wherein:said flow opening has a lower edge; andsaid flow vane is a wall region of said respective outer web, said wall region having two cutouts formed therein defining said flow vane, said two cutouts extending from said lower edge opening into said outer web in an upstream direction, said two cutouts laterally delimiting said flow vane. 4. The spacer according to claim 2, wherein said flow vane is formed by a deep-drawn wall region of said respective outer web that adjoins a lower edge of said flow opening. 5. The spacer according to claim 1, wherein each of said cells adjoining one of said outer webs is assigned at least one said guide device. 6. The spacer according to claim 1, wherein said flow opening has a greater width than said guide element. 7. The spacer according to claim 1, wherein said flow opening is formed as an elongated hole and disposed obliquely to the flow direction. 8. The spacer according to claim 7, wherein said flow opening is one of two adjacently disposed flow openings that are positioned obliquely in opposite senses and enclose an acute angle β opening against the flow direction. 9. The spacer according to claim 8, wherein said two flow openings are assigned to neighboring ones of said cells. 10. The spacer according to claim 1, wherein an upper edge of said flow opening and a wall region of said respective outer web subsequent thereto in the flow direction are cambered convexly outward.
	claims	1. A charged particle device comprising a charged particle source configured to direct charged particles in the direction of a specimen under examination and an imaging device configured to convert charged particles to an image representing said specimen, wherein:said imaging device comprises a detector defining a pixel array;said detector is configured to generate electric charges for individual pixels of said pixel array such that said electric charges collectively define said image;said imaging device is configured such that a portion of said pixel array can be transitioned between a partially masked state and a substantially unmasked state;said partially masked state defines a set of masked readout pixels and a set of unmasked imaging pixels;said imaging device is programmed to operate in a frame transfer imaging mode when said pixel array is in said partially masked state and in a full frame mode when said pixel array is in said substantially unmasked state;said frame transfer imaging mode is configured such that an image captured by said set of unmasked imaging pixels is transferred to said set of masked readout pixels, said transferred image is subsequently transferred from said set of masked readout pixels to storage outside of said pixel array as a subsequent image is captured by said set of unmasked imaging pixels, and said subsequently captured image is transferred to said set of masked readout pixels after said initial image is transferred from said set of masked readout pixels to said outside storage; andsaid full frame mode is configured such that an image captured by said pixel array in said substantially unmasked state is transferred to storage outside of said pixel array as incoming charged particles are blocked from reaching said detector. 2. A charged particle device as claimed in claim 1 wherein said frame transfer imaging mode is configured such that approximately 50% of said pixel array is dedicated to said set of unmasked imaging pixels. 3. A charged particle device as claimed in claim 1 wherein said frame transfer imaging mode is configured such that substantially equal portions of said pixel array are dedicated to said respective sets of masked and unmasked pixels. 4. A charged particle device as claimed in claim 1 wherein said frame transfer imaging mode is configured such that said set of unmasked imaging pixels is positioned between respective portions of said set of masked readout pixels along a pixel plane defined by said pixel array. 5. A charged particle device as claimed in claim 1 wherein said frame transfer imaging mode is configured such that said set of unmasked imaging pixels is defined in a central portion of said pixel array and said set of masked readout pixels is defined in lateral portions of said pixel array. 6. A charged particle device as claimed in claim 1 wherein:said imaging device comprises a mechanical mask movable between a first position corresponding to said substantially unmasked state to a second position corresponding to said partially masked state; andsaid mechanical mask comprises a pair of shutters mounted to move towards and away from one another across said pixel array. 7. A charged particle device as claimed in claim 1 wherein said charged particle beam device further comprises system optics and said imaging device comprises a virtual mask implemented through the manipulation of said system optics to prevent charged particles from reaching a portion of said imaging device corresponding to said set of masked readout pixels. 8. A charged particle device as claimed in claim 1 wherein said charged particle device further comprises a user interface and a controller programmed to prompt a user to select one of a plurality of imaging modes via said user interface. 9. A charged particle device as claimed in claim 1 wherein said charged particle device further comprises a controller programmed to affect said transition of said pixel array between said partially masked state and said substantially unmasked state. 10. A charged particle device as claimed in claim 1 wherein said imaging device is configured such that said pixel array can be transitioned from said partially masked state to said substantially unmasked state without substantial interruption to operation of said device. 11. A charged particle device as claimed in claim 1 wherein said charged particle source is configured to generate a beam of charged particles comprising electrons, protons, ions, or combinations thereof. 12. A charged particle device as claimed in claim 1 wherein said charged particles comprise electrons characterized by a kinetic energy of at least about 1 keV. 13. A charged particle device as claimed in claim 1 wherein said charged particle device comprises an electron microscope. 14. A charged particle device as claimed in claim 13 wherein said electron microscope further comprises an electron energy loss spectrometer. 15. A charged particle device as claimed in claim 1 wherein said imaging device comprises said detector and one or more components selected from an energy selecting slit, a charged particle dispersing device, a charged particle lens, a charged particle deflector, a charged particle energy filter, a charged particle scintillator, and a fiber optic coupler. 16. A charged particle device as claimed in claim 1 wherein said detector comprises a CCD array, a photodiode array, or a CMOS detector. 17. A charged particle device as claimed in claim 1 wherein said image represents dimensional bounds of said specimen, physical properties of said specimen, material constituents of said specimen, or combinations thereof. 18. A charged particle device as claimed in claim 6 wherein said mechanical mask comprises a movable piston, a cam assembly, and at least one biasing spring configured such that application of pressure causes said piston to move said cam assembly so as to overcome the bias of said spring and cause the respective shutters to retract, exposing additional surface area of said pixel array. 19. A charged particle device as claimed in claim 18 wherein said cam assembly is configured such that said exposure of said additional surface area attributable to respective ones of said pair of shutters is substantially equivalent. 20. A charged particle device as claimed in claim 19 wherein said cam assembly comprises a pair of wedge-shaped cams. 21. A charged particle device as claimed in claim 6 wherein said set of masked readout pixels and said set of unmasked imaging pixels define substantially equal portions of said pixel array. 22. A charged particle device as claimed in claim 6 wherein a portion of said mechanical mask defining said masked readout pixels comprises a shielding material characterized by an atomic number of below about 30 in an amount sufficient to render the amount of hard X-rays emitted from said mask and reaching said pixel array to an insubstantial amount during operation of said device.
	summary
	summary
039322122	description	The construction and method of operation of the invention, however, together with additional objects and advantages thereof will be best understood from the following description when read in connection with the accompanying drawings, in which: FIG. 1 is a schematic circuit diagram of the steam and water circulation system of a boiling water reactor, of the invention. FIG. 2 is a schematic circuit diagram of means for guiding the secondary condensates. FIG. 3 is a cross section through a condenser system showing the condenser having a lower condensate collecting tank, taken along line III--III of FIG. 4. FIG. 4 is a longitudinal section through the condenser assembly lower part taken along line IV--IV of FIG. 3, and FIG. 5 is a horizontal section through the condensate collecting tank, taken along line V--V of FIG. 4. FIG. 1 shows the circuit diagram for a boiling water reactor. From the boiling water reactor 100, the generated saturated steam flows first through the line 101, to the saturated steam turbine 102 and, therefrom, through lines 103, into the steam desiccator 104 and through the intermediate overheater 106 which is heated by fresh steam, via bypass line 105. From this intermediate overheater, the steam arrives through line 107, in the low pressure turbine 1. The turbines 102 and 1 are seated on a common shaft and drive the generator 108. The exhaust steam of the low pressure part of turbine 1 is guided through the line 2, into the condenser and, here, following the condensation and collection in the condensate collecting tank 4, returned through the condensate pump 109, into the cycle, in form of feed-water. The feed-water then flows in the line 110, first in the heat exchanger 111 wherein the drive steam of the steam jet vacuum pump 112 gives off its heat to the feed-water. Thereafter, the feed-water passes through the low pressure preheaters 113 and 114 which are heated via bleeder steam lines 115 and 116, with bleeder steam from the low pressure turbine 1. The feed-water pipe 117, proper, is followed by the high pressure preheaters 118 and 119 which, by bleeder steam lines 120, 121 are heated with bleeder steam from the saturated steam turbine 102. A line 122 is connected into the last high pressure preheater 119 from the intermediate overheater 106 drain or discharge. The condensates of the entire preheater path are gradually guided back through lines 123, 124 and 125, from preheater 119 to preheater 113. From the first low pressure preheater 113, the condensate is then guided through the condensate cooler 126 and the line 6, into the pressure relief chamber 5. This so-called secondary condensate is utilized now, in accordance with the invention, for the degassing of the main condensate. The pressure relief chamber 5 which is shown as a vertical standpipe absorbs all constantly collecting secondary condensates such as, for example, the condensate from the preheater 111 which is guided through line 6' to the chamber 5. These secondary condensates collect at a higher pressure than the main condensate in the condenser 3. The chamber 5 is connected, by line 127, with the condenser 3, through a throttle or choke 7, installed into the line 127. The steam being collected in the chamber 5 is thereby guided directly back into the condenser 3 and is condensed therein. Through the choke 7, a slightly higher pressure than prevails in the condenser 3, is established in the chamber 5. Thus, for example, the condenser pressure may amount to 0.04 atmospheres at 28.degree. C, while the secondary condensate in the decompressor 5 is under a pressure of 0.06 atmospheres at 35.degree.. The water is therefore approximately at the boiling point. Referring to FIG. 2, the remaining expansion of the secondary condensates and the simultaneous degassing of the main condensate is now effected in the degassing channels 15, 16, 17 and 18, which are schematically illustrated in FIG. 2 and in cross-section in FIG. 3. Depressurizing chambers 9, 10 and 10a, are connected through suitable valves and connectors 20 - 23 to channels 15 - 18. FIG. 3 shows a cross-section through the entire condenser 3 with the condensate collecting tank 4 underneath. The exhaust steam from the low-pressure turbine 1 flows through the steam dome 130 towards the piping or baffles 131 of the condenser 3. This piping 131 which is cooled is divided into individual packages whose peripheries are shown in the drawn cross-section. Also, the condenser has in its central lower portion an air cooler unit 132 where oxygen-rich condensate is precipitated. These air cooler units are disposed within a separate housing 133, open on the bottom. The degassing channels 15, 16 and 17, in which the condensate flows have on a longitudinal wall, an overflow edge 28, so that a condensate level A is created in these channels. Beneath the air cooler units 132, the guide baffles 25 and 26 are arranged so that the oxygen-rich condensate flows from these air cooler units 132, separately, toward the channels 16 and 17. The remaining condensate is caught on the right side by the baffles 24a and 24b and, on the left side, by the baffles 27 and 27b and are guided past channels 16 and 17 having the oxygen-rich condensate, into the channels 15 and 18. The condensate is then guided through baffles 24 and 27, from the external region into the channels 15 and 18. In these channels 15 to 18, are arranged horizontal feed pipes 11, 12, 13 and 14, below the overflow edge 28, so that these feed pipes are disposed completely under the resulting condensate level. These feed pipes 11 to 14 are in contact by lines 8, (FIGS. 1, 2 and 4) with the depressurizing chamber 5, and pass the secondary condensate from the chamber 5, into these feed pipes 11 to 14. These pipes have discharge openings 30 on their undersides through which the secondary condensate, as previously indicated is at higher pressure and at higher temperature than the main condensate, can flow out. The pressure of the secondary condensate in the feed pipes 11 to 14 must be high enough that this secondary condensate will reliably bubble out downwardly from the openings 30 and overcome the pressure of the thereabove situated water column of the main condensate. The pressure of the secondary condensate in the chamber 5 is so adjusted that the pressure drop through the throttle or choke 7 and the water column of the main condensate disposed above the discharge openings 30, just about maintains a balance. Following the discharge of the secondary condensate from the openings 30, it bubbles out and evaporates immediately due to the low pressure of the main condensate and thereby provides the degassing of the main condensate. The condensate running off from the overflows 28 is caught by a slightly inclined runoff sheet 31 and fed to an outlet discharge nozzle 32, by the inclined outlet 33 on the outside of the run-off sheet 31. Therefore, a mixing of the accumulating oxygen-rich condensate from the air cooler unit 132 and the remaining condensate, is effected on the run-off sheet 31. In such a boiling water plant not only the constant secondary condensates are accumulated but, also, additional secondary condensates occurring in the event of a disturbance, if, for example, the load at generator 108 is suddenly disconnected or if other interferences occur. During such a disruption, additional secondary condensates accumulate in great amounts and are collected in separated pressure relief chambers 9, 10 and 10a (FIG. 2). These secondary condensates may arrive for example, via line 218, from the steam overheating run-off 106 or may arrive via emergency run-offs 129, from the lowest preheater 113 when such a preheater stops working and is bridged. Other, not specifically illustrated, disturbances which also produce a considerable accumulation of secondary condensates are, for example, the stoppage of the steam generator for the turbine or of the decompressor for the feed pump. These secondary condensates which collect during an interference, are collected in pressure relief standpipes 9 and 10 and are expanded to the pressure of the condenser 3. In order to mix the secondary condensates occurring during a disturbance, also with the main condensate, the arrangement as shown in FIG. 4 was provided. Here, the inlet pipes 11, 12, 13 and 14 extend only approximately about three-quarters of the total length of the degassing channels 15 to 18. The read ends of these channels 15 to 18 are separated by vertical baffles 34. In this area, separated by the baffles 34, the secondary condensates collect during a disturbance from pressure relief standpipes 9 and 10, FIGS. 1 and 2, are directly fed by pipe line 19 and the inlets 23 which open into the channel 18. Due to the greater amount accumulating during a disturbance, the degassing process which occurs during operation, is no longer possible and is not required. By thus dividing the condenser into two halves, where each half is fully operable by itself, one-half of the condenser may be shut down in case of a leak in order to prevent cooling water from getting into the circuit. As will be seen in FIG. 3, the steam space above the run-off sheet 31 is divided into two halves by a partition 31a which extends transversely to the walls 36 but above the collecting space 35. At the outer ends, decay chamber 35 has condensate feed connections 44 and 45, and 46 and 47, respectively, which are connected with the outlets 32 for the condensate running off from the channels 15 to 18 via reversing lines, not shown. Because in a boiling-water reactor the condensates are weakly radioactive, it is necessary to provide additionally for these condensates a decay section in which the condensates must remain for a definite time before they are returned to circulation via the condensate pumps. Accordingly, dwelling sections are obtained by the provision of a specially designed collecting space 35 disposed underneath the run-off sheets 31 in the lower part of the condenser 4. As shown in FIG. 5, this collecting space 35 is subdivided by vertical partitions 36 into individual lanes. These partitions, however, have cutouts 37 and 38 which are staggered relative to one another, so that the condensates entering via inlets 44 to 47 are directed through the condenser collecting space 35 along winding or tortuous paths as indicated by the dashed lines 41 in FIG. 5. By providing this flow path for the condensate, a definite decay time for the radioactivity of the condensate is therefore obtained because of the longer flow time which is required. After flowing through this decay path, the condensate is fed to the main condensate pump via connections 42. By the method and the apparatus according to the present invention, degassing of the condensate is, therefore, effected directly in the condenser, and an appropriate subdivision of the cooling pipes makes it possible to degas the oxygen-rich condensate which comes from the air cooler units separately from the rest of the condensate. In addition, a true separation of degassing and storage spaces has been obtained where the storage spaces themselves, due to appropriate design, additionally serve as the decay section for the radioactivity of the condensate. The arrangement described provides a very low structural height. In operation in the turbine system as shown in FIG. 1 of the application, the main condensate of the exahust steam comes directly from turbine 1 collects in the condenser 3. In addition, the turbine system includes so-called secondary condensates which flow from the individual preheater stages or from steam-jet ejectors. Since, for the preheating of the feed water, bleeder steam is taken from higher pressure stages of the turbine as the heating medium, the condensate of the bleeder steam, after flowing from the individual preheaters, exhibits a higher pressure and higher temperature than the main condensate in the condenser 3. Both condensates consist of water. The secondary condensates coming from the preheater stages are fed through the lines 6 to the members 5. The pressure relieving standpipes 5 are connected with the main condenser 3 by means of a conduit provided with a choke 7. Since, in the standpipes 5, due to the higher pressure of the secondary condensates, a higher pressure prevails than in the condenser 3, suitable adjustment of the choke 7 can result in a lower pressure in the standpipes 5. Therefore, a depressurization of the secondary condensates takes place. In addition to the standpipes 5, there are also provided standpipes 9, 10 and 10a, wherein secondary condensates are collected which are produced in the event of a disturbance. Such a disturbance, may, for example arise when a preheater is defective and is disconnected and must be bridged by a bypass line. Condensate is produced also in this case, particularly from so-called mixing preheaters. This condensate produced in the event of a disturbance should be collected separately. The condensate flowing from condenser 3 is now collected in degassing channels 15 to 18. The secondary condensate from standpipe 5 is conducted into the individual pipes 11 to 14 through line 8. These pipes are provided with openings at their underside, through which the secondary condensate flows into the main condensate. However, since the secondary condensate exhibits a higher pressure and higher temperature than the main condensate, it will immediately vaporize upon leaving pipes 11 to 14 and rise to the surface in the form of steam bubbles. Since the secondary condensate has a higher temperature than the main condensate, the main condensate is thereby heated to the point where it practically begins to boil, so that in this way an additional deaeration of the main condensate is assured. Regarding the feeding of the various condensates, that is, the oxygen-enriched condensate from the air coolers and the remaining main condensate, it must be noted that this is not done selectively but these two condensates are produced continuously and in each case at the same location. The oxygen-enriched condensate drips onto the run-off baffles 25 and 26 and thus is fed to the channels 16 and 17, whereas the remaining main condensate falls upon the remaining surface, that is, baffles 27a and 27b as well as 27, and is conducted from there into the degassing channels 15 and 18. Conduction of the condensate from the run-off surfaces 25 and 26 onto the baffle 27 is not provided for. Rather, the condensates collect only after running over the overflow baffles 28 onto the run-off sheet 31 where they are mixed. During this running off, a further degassing takes place, to a small extent. All condensates caught by the run-off baffles 26, 27, 27a and 27b, come from the main condenser 3. Only the secondary condensates come from the line 8 and are fed therefrom through the pipes 11 - 14 from which the secondary condensates emerge.
	description	The present application is a continuation of U.S. patent application Ser. No. 15/053,608, filed Feb. 25, 2016, issuing as U.S. Pat. No. 9,514,853. U.S. patent application Ser. No. 15/053,608 is continuation-in-part of U.S. patent application Ser. No. 14/534,391, filed Nov. 6, 2014, which is a continuation of U.S. patent application Ser. No. 13/208,915, filed Aug. 12, 2011, which in turn claims the benefit of U.S. Provisional Patent Application Ser. No. 61/373,138, filed Aug. 12, 2010. U.S. patent application Ser. No. 15/053,608 is also a continuation-in-part of U.S. patent application Ser. No. 14/394,233, filed Oct. 13, 2014, issuing as U.S. Pat. No. 9,396,824, which is a United States national stage application under 35 U.S.C. § 371 of PCT Application No. PCT/US2013/036592, filed on Apr. 15, 2013, which claims the benefit of U.S. Provisional Patent Application No. 61/624,066 filed Apr. 13, 2012. U.S. patent application Ser. No. 15/053,608 is also a continuation-in-part of U.S. patent application Ser. No. 14/395,790, filed Oct. 20, 2014, which is a U.S. national stage application under 35 U.S.C. § 371 of PCT Application No. PCT/US2013/037228, filed on Apr. 18, 2013, which claims the benefit of U.S. Provisional Patent Application 61/625,869, filed Apr. 18, 2012. U.S. patent application Ser. No. 15/053,608 is also a continuation-in-part of U.S. patent application Ser. No. 14/424,201, filed Feb. 26, 2015, issuing as U.S. Pat. No. 9,442,037, which is a United States national stage application under 35 U.S.C. § 371 of PCT Application No. PCT/US2013/057855, filed Sep. 31, 2013, which claims priority to U.S. Provisional Application Ser. No. 61/695,837, filed Aug. 31, 2012. U.S. patent application Ser. No. 15/053,608 is also a continuation-in-part of U.S. patent application Ser. No. 14/655,860, filed Jun. 25, 2015, which is a United States national stage application under 35 U.S.C. § 371 of PCT Application No. PCT/US2013/077852 filed Dec. 26, 2013, which claims priority to U.S. Provisional Application Ser. No. 61/746,094 filed Dec. 26, 2012. U.S. patent application Ser. No. 15/053,608 is also a continuation-in-part of U.S. patent application Ser. No. 14/762,874, filed Jul. 23, 2015, issuing as U.S. Pat. No. 9,466,400, which is a United States national stage application under 35 U.S.C § 371 of PCT Application No. PCT/US2014/013185, filed Jan. 27, 2014, which claims priority to U.S. provisional application No. 61/902,559 filed Jan. 25, 2013, and to U.S. provisional application No. 61/902,550, filed Nov. 11, 2013. The disclosures of the aforementioned priority applications are incorporated herein by reference in their entireties. The storage, handling, and transfer of high level waste, (hereinafter, “HLW”) such as spent nuclear fuel (hereinafter, “SNF”), requires special care and procedural safeguards. For example, in the operation of nuclear reactors, it is customary to remove fuel assemblies after their energy has been depleted down to a predetermined level. Upon removal, this spent nuclear fuel is still highly radioactive and produces considerable heat, requiring that great care be taken in its packaging, transporting, and storing. In order to protect the environment from radiation exposure, spent nuclear fuel is first placed in a canister. The loaded canister is then transported and stored in large cylindrical containers called casks. A transfer cask is used to transport spent nuclear fuel from location to location while a storage cask is used to store spent unclear fuel for a determined period of time. In a typical nuclear power plant, an open empty canister is first placed in an open transfer cask. The transfer cask and empty canister are then submerged in a pool of water. Spent nuclear fuel is loaded into the canister while the canister and transfer cask remain submerged in the pool of water. Once fully loaded with spent nuclear fuel, a lid is typically placed atop the canister while in the pool. The transfer cask and canister are then removed from the pool of water, the lid of the canister is welded thereon and a lid is installed on the transfer cask. The canister is then properly dewatered and filled with inert gas. The transfer cask (which is holding the loaded canister) is then transported to a location where a storage cask is located. The loaded canister is then transferred from the transfer cask to the storage cask for long term storage. During transfer from the transfer cask to the storage cask, it is imperative that the loaded canister is not exposed to the environment. One type of storage cask is a ventilated vertical overpack (“VVO”). A VVO is a massive structure made principally from steel and concrete and is used to store a canister loaded with spent nuclear fuel (or other HLW). VVOs stand above around and are typically cylindrical in shape and extremely heavy, weighing over 150 tons and often having a height greater than 16 feet. VVOs typically have a flat bottom, a cylindrical body having a cavity to receive a canister of spent nuclear fuel, and a removable top lid. In using a VVO to store spent nuclear fuel, a canister loaded with spent nuclear fuel is placed in the cavity of the cylindrical body of the VVO. Because the spent nuclear fuel is still producing a considerable amount of heat when it is placed in the VVO for storage, it is necessary that this heat energy base a means to escape from the VVO cavity. This heat energy is removed from the outside surface of the canister by ventilating the VVO cavity. In ventilating the VVO cavity, cool air enters the VVO chamber through bottom ventilation ducts, flows upward past the loaded canister, and exits the VVO at an elevated temperature through top ventilation ducts. The bottom and top ventilation ducts of existing VVOs are located near the bottom and top of the VVOs cylindrical body respectively. While it is necessary that the VVO cavity be vented so that heat can escape from the canister, it is also imperative that the VVO provide adequate radiation shielding and that the spent nuclear fuel not be directly exposed to the external environment. The inlet duct located near the bottom of the overpack is a particularly vulnerable source of radiation exposure to security and surveillance personnel who, in order to monitor the loaded overpacks, must place themselves in close vicinity of the ducts for short durations. Thus, a need exists for a VVO system for the storage of high level radioactive waste that has an inlet duct that reduces the likelihood of radiation exposure while providing extreme radiation blockage of both gamma and neutron radiation emanating from the high level radioactive waste. The effect of wind on the thermal performance of a ventilated system can also be a serious drawback that, to some extent, afflicts all systems in use in the industry at the present time. Storage VVO's with only two inlet or outlet ducts are especially vulnerable. While axisymmetric air inlet and outlet ducts behave extremely well in quiescent air, when the wind is blowing, the flow of air entering and leaving the system is skewed, frequently leading to a reduced heat rejection capacity. The thick top lid is one of the most expensive components of a radioactive waste canister. Such canisters may be used to store and transport non-fuel radioactive waste from nuclear generation plants such as activated reactor internals, control components, sundry non-fissile materials, and waste from operations such as resins, and in some applications vitrified nuclear waste fuel (“glass logs”) encased in an outer metal cylinder. On existing canisters, the thick top lid is needed to shield personnel from radiation who are working on the lid (e.g. welding, bolting, fluid operations, etc.). The lid must also be thicker because the lid further performs the main canister lifting connection, and therefore must have the thickness needed for structural reasons to support the weight of the entire canister when hoisted via a crane or similar equipment used to move the canister. For these reasons, the thick top lid of a waste canister adds considerably to the overall weight and expense of the canister. An improved radioactive waste canister is desired. A need also exists periodic leak testing is often required for monitoring the integrity of the inner and outer confinement boundaries on canisters holding radioactive materials. Some present leak testing processes involve removing the cask lid, which is undesirable, as doing so has the potential to increase radiation exposure to workers. Other leak testing processes and systems involve installing a continuous leak testing monitoring system that uses a compressed helium tank and pressure transducers. Such a system, however, requires periodic replacement of the transducers and replenishment of the helium gas stored in the tank. In view of the shortcomings of present leak detection processes and systems, improvements are desirable which reduce the on-site maintenance requirements, improve leak detection capabilities, and reduce potential radiation exposure to workers. A need also exists for the ability to better examine welds formed on containers that are used to store spend nuclear fuel. Finally, a need exists to better enable spent nuclear fuel to be transferred from place to place as necessary. These, and other drawbacks, are remedied by the present invention. In one embodiment, the invention can be a system for storing high level radioactive waste comprising: an overpack body extending along a vertical axis and having a cavity for storing high level radioactive waste, the cavity having an open top end and a floor; an overpack lid positioned atop the overpack body to enclose the open top end of the cavity; an air inlet vent for introducing cool air into the cavity, the air inlet vent extending from an opening in an outer surface of the overpack body to an opening in the floor, the opening in the outer surface of the overpack body extending about an entirety of a circumference of the outer surface of the overpack body, and an air outlet vent in the overpack lid for removing warmed air from the cavity. In another embodiment, the invention can be a system for storing high level radioactive waste comprising: an overpack body extending along a vertical axis and having a cavity for storing high level radioactive waste, the cavity having an open top end and a floor, the overpack body comprising an air inlet vent for introducing cool air into a bottom portion of the cavity; a plurality of plates disposed within a portion of the air inlet vent, each of the plates extending along a reference line that is tangent to a third reference circle having a center point coincident with the vertical axis; and an overpack lid positioned atop the overpack body to enclose the open top end of the cavity, the overpack lid comprising an air outlet vent for removing warmed air from the cavity. In yet another embodiment, the invention can be a system for storing high level radioactive waste comprising: an overpack body extending along a vertical axis and having a cavity for storing high level radioactive waste, the cavity having an open top end and a floor, the overpack body comprising an air inlet vent for introducing cool air into a bottom portion of the cavity; an overpack lid positioned atop the overpack body to enclose the open top end of the cavity, the overpack lid comprising an air outlet vent for removing warmed air from a top portion of the cavity; and the air inlet vent comprising a first section that extends substantially horizontally from an outer surface of the overpack body to a terminal end and a second section extending from the first section of the air inlet vent to an opening in the floor at an oblique angle relative to the vertical axis. In still another embodiment, the invention can be a radioactive waste container system comprising: a canister having an interior chamber for holding radioactive waste and an open top; a lid assembly comprising a confinement lid and a shielded lifting lid, the confinement lid being detachably mounted to the lifting lid; the confinement lid being configured for mounting on the canister and having a first thickness; the lifting lid including a lifting attachment and having a second thickness; wherein the confinement lid is independently mountable on canister from the lifting lid. In still a further embodiment, the invention can be a radioactive waste container system comprising: a canister having an interior chamber for holding radioactive waste and an open top; a lid assembly comprising a lower confinement lid and an upper shielded lifting lid, the confinement lid being detachably bolted to the lifting lid; the lifting lid including a plurality of first bolt holes having a first diameter and a plurality of second bolt holes having a second diameter, the first diameter being larger than the second diameter; the confinement lid including a plurality of third bolt holes having a third diameter, wherein each of the third bolt holes is concentrically aligned with one of the first or second bolt holes of the lifting lid, and a plurality of first mounting bolts inserted through the first bolt holes and threadably attaching the confinement lid to the canister without engaging the lifting lid. In a yet further embodiment, the invention can be a method for storing radioactive waste using a container system, the method comprising: detachably mounting a confinement lid to a shielded lifting lid, the confinement lid and shielded lifting lid collectively forming a lid assembly; placing a canister having an interior chamber for holding radioactive waste into an outer protective overpack; lifting the lid assembly using the lifting lid; placing the lid assembly on an open top of the canister; attaching the confinement lid to the canister using a first set of mounting bolts without threadably engaging the lifting lid with the bolts; detaching the lifting lid from the confinement lid; and removing the lifting lid from the canister. In another embodiment, the invention cast be a module for storing high level radioactive waste, the module comprising: an outer shell having a hermetically closed bottom end; an inner shell forming a cavity, the inner shell positioned inside the outer shell so as to form a space between the inner shell and the outer shell; at least one divider extending from a top of the inner shell to a bottom of the inner shell, the at least one divider creating a plurality of inlet passageways through the space, each inlet passageway connecting to a bottom portion of the cavity; a plurality of inlet ducts, each inlet duct connecting at least one of the inlet passageways to ambient atmosphere and each comprising an inlet duct cover affixed over a surrounding inlet wall, the inlet wall being peripherally perforated; and a removable lid positioned atop the inner shell, the lid having at least one outlet passageway connecting the cavity and the ambient atmosphere, wherein the lid and a top of the inner shell are respectively configured to form a hermetic seal at a top of the cavity. In still another embodiment, the invention can be a system for storing radioactive materials, the system comprising: a canister comprising: a first hermetically sealed vessel having a first cavity; a second hermetically sealed vessel having a second cavity, wherein the first vessel is positioned in the second cavity; an interstitial space between the first and second vessels; and a test port through the second vessel in fluidic communication with the interstitial space; a conduit having a first end fluidically coupled to the test port; and a removable seal operably coupled to a second end of the conduit. In yet another embodiment, the invention can be a method of storing radioactive materials, the method comprising: a) providing a cask having a cask body that forms a cask cavity having an open top end; b) positioning a canister loaded with the radioactive materials in the cask cavity, the canister comprising a first hermetically sealed vessel having a first cavity in which the radioactive materials are disposed and a second hermetically sealed vessel having a second cavity, wherein the first vessel is positioned in the second cavity, such that an interstitial space exists between the first and second vessels, and wherein the second vessel includes a test port that is in fluidic communication with the interstitial space; c) fluidically coupling a first end of a conduit to the test port, the conduit extending from the first end to a second end located outside of the cask; and d) securing a cask lid to the cask body to substantially enclose the open top end of the cask cavity. In another embodiment still, the invention can be a system for leak testing a canister containing radioactive materials, the system comprising: a canister comprising: a first hermetically sealed vessel having a first cavity; a second hermetically sealed vessel having a second cavity, wherein the first vessel is positioned in the second cavity; an interstitial space between the first and second vessels; and a test port through the second vessel in fluidic communication with the interstitial space; a conduit having a first end fluidically coupled to the test port; a removable seal operably coupled to a second end of the conduit, and a leak detector configured to operably couple to the second end of the conduit and to detect whether a leak exists in at least one of the first vessel and the second vessel. In a further embodiment, the invention can be a method of leak testing a storage canister for radioactive materials, the method comprising: a) positioning the canister in a cask cavity of a cask body, the canister comprising a first hermetically sealed vessel having a first cavity in which the radioactive materials are disposed and a second hermetically sealed vessel having a second cavity, the first vessel positioned in the second cavity such that an interstitial space exists between the first and second vessels, and wherein the second vessel includes a test port that is in fluidic communication with the interstitial space; b) coupling a first end of a conduit to the test port, the conduit extending from the first end to a second end located outside of the cask body; c) securing a cask lid to the cask body to substantially enclose the cask cavity; and d) operatively coupling a leak defector to the second end of the conduit to perform a leak test comprising determining whether a leak exists in at least one of the first vessel and the second vessel In a still further embodiment, the invention can be a method of leak testing a canister containing radioactive materials, the method comprising: a) coupling a first end of a conduit to a test port of the canister that is in fluid communication with an interstitial space of the canister, the conduit extending from the first end to a second end; and b) operatively coupling a leak detector to the second end; c) drawing gas from the conduit using the leak detector to establish a vacuum within the conduit and the interstitial space; and d) monitoring the drawn gas for the presence of a first indicator which is representative of a leak in a fluidic containment boundary of the canister that contains the radioactive materials. In another embodiment, the invention can be a canister for storing radioactive materials, the canister comprising: a base plate; a side wall having a bottom sealed to the base plate; and a top plate including a top surface with a top edge having a bevel and with a channel set in from the top edge, wherein a weld is formed between the beveled top edge and a top of the side wall to seal the top plate to the side wall, and wherein the base plate, side wall, and top plate form a sealed vessel. In another embodiment, the invention can be a method of forming a sealed canister, the method comprising: placing a top plate on a top opening of a side walk a bottom of the side wall being sealed to a base plate, wherein the top plate includes a top surface with a top edge having a bevel and with a channel set in from the top edge; and forming a weld between the beveled top edge and the top opening of the side wall to seal the top plate to the side wall. In another embodiment, still, the invention can be a method of storing radioactive materials, the method comprising: placing radioactive materials in a cavity formed by a side wall having a bottom sealed to a base plate; placing a top plate on a top opening of the side wall, the top plate including a top surface with a top edge having a bevel and with a channel set in from the top edge; forming a weld between the beveled top edge and the top opening of the side wall to seal the top plate to the side wall, so that the cavity is sealed; placing a first probe in the channel and a second probe opposite the first probe and adjacent the side wall, such that the weld is disposed between the two probes; activating the first and second probes to determine an integrity of a volume of the weld between the probes; and moving the first and second probes synchronously around the top plate to determine the integrity of an entire volume of the weld. In another embodiment, the invention can be an apparatus for transferring spent nuclear fuel, the apparatus comprising: a cylindrical inner shell forming a cavity configured to receive a canister containing spent nuclear fuel, the cavity configured so that an annulus is formed between a canister placed in the cavity and an inner wall of the cylindrical inner shell; an intermediate shell disposed concentrically around and spaced apart from the inner shell; an outer shell disposed concentrically around and spaced apart from the intermediate shell; a bottom flange affixed to bottoms of each of the shells; a bottom lid removably affixed to the bottom flange and including at least one first channel fluidically connecting the annulus to an exterior of the bottom lid, wherein the at least one first channel is configured to preclude a direct line of travel from within the cavity to the exterior of the bottom lid; a top flange affixed to tops of each of the shells and including at least one second channel fluidically connecting the first annulus to an exterior of the top flange, wherein the at least one second channel is configured to preclude a direct line of travel from within the cavity to the exterior of the top flange; and a top lid removably affixed to the top flange. In yet another embodiment, the invention can be an apparatus for transferring spent nuclear fuel, the apparatus comprising: a cylindrical inner shell forming a cavity configured to receive a canister containing spent nuclear fuel; an intermediate shell disposed concentrically around and spaced apart from the inner shell; an outer shell disposed concentrically around and spaced apart from the intermediate shell; a bottom flange affixed to bottoms of each of the shells; a bottom lid removably affixed to the bottom flange; a top flange affixed to tops of each of the shells, the top flange including at least two integrally formed trunnions configured to enable hoisting of the apparatus; and a top lid removably affixed to the top flange. In still another embodiment, the invention can be an apparatus for transferring spent nuclear fuel, the apparatus comprising: a cylindrical inner shell forming a cavity configured to receive a canister containing spent nuclear fuel; an intermediate shell disposed concentrically around and spaced apart from the inner shell; an outer shell disposed concentrically around and spaced apart from the intermediate shell; a bottom flange affixed to bottoms of each of the shells; a bottom lid removably affixed to the bottom flange, the bottom lid including an impact zone comprising an impact absorbing structure; a top flange affixed to tops of each of the shells; and a top lid removably affixed to the top flange. In another embodiment, the invention can be a method for transferring spent nuclear fuel from a pool, the method comprising: lifting a transfer cask from a pool, the transfer cask comprising: a cylindrical inner shell forming a cavity configured to receive a canister containing spent nuclear fuel, the cavity configured so that an annulus is formed between a canister placed in the cavity and an inner wall of the cylindrical inner shell; an intermediate shell disposed concentrically around and spaced apart from the inner shell; an outer shell disposed concentrically around and spaced apart from the intermediate shell; a bottom flange affixed to bottoms of each of the shells; a bottom lid removably affixed to the bottom flange and including at least one first channel fluidically connecting the annulus to a channel inlet at an exterior of the bottom lid, wherein the at least one first channel is configured to preclude a direct line of travel from within the cavity to the exterior of the bottom lid; a removable plug sealingly affixed to the channel inlet, a top flange affixed to tops of each of the shells and including at least one second channel fluidically connecting the first annulus to an exterior of the top flange, wherein the at least one second channel is configured to preclude a direct line of travel from within the cavity to the exterior of the top flange, and a top lid removably affixed to the top flange; removing the removable plug from the channel inlet, thereby allowing ambient air to enter the at least one first channel; draining the pool water from the canister; and moving the transfer cask to a staging area. Further areas of applicability of the present invention will become apparent from the detailed description provided hereinafter. It should be understood that the detailed description and specific examples, while indicating the preferred embodiment of the invention, are intended for purposes of illustration only and are not intended to limit the scope of the invention. All drawings are schematic and not necessarily to scale. Parts 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. The following description of the preferred embodiment(s) is merely exemplary in nature and is in no way intended to limit the invention, its application, or uses. The description of illustrative embodiments according to principles of the present invention is intended to be read in connection with the accompanying drawings, which are to be considered part of the entire written description. In the description of embodiments of the invention 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 derivatives 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 unless explicitly indicated as such. 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. Moreover, the features and benefits of the invention are illustrated by reference to the exemplified embodiments. Accordingly, the invention 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; the scope of the invention being defined by the claims appended hereto. Multiple inventive concepts are described herein and are distinguished from one another using headers in the description that follows. Specifically, FIGS. 1-8 are relevant to a first inventive concept, FIGS. 9-26 are relevant to a second inventive concept, FIGS. 27-34 are relevant to a third inventive concept, FIGS. 35-47 are relevant to a fourth inventive concept, FIGS. 48-52B are relevant to a fifth inventive concept, and FIGS. 53-59 are relevant to a sixth inventive concept. The first through sixth inventive concepts should be considered in isolation from one another. It is possible that there may be conflicting language or terms used in the description of the first through sixth inventive concepts. For example, it is possible that in the description of the first inventive concept a particular term may be used to have one meaning or definition and that in the description of the second inventive concept the same term may be used to have a different meaning or definition. In the event of such conflicting language, reference should be made to the disclosure of the relevant inventive concept being discussed. Similarly, the section of the description describing a particular inventive concept being claimed should be used to interpret claim language when necessary. With reference to FIGS. 1-8, a first inventive concept will be described. Referring to FIGS. 1-4 concurrently, a system for storing high level radioactive waste will be described in accordance with an embodiment of the present invention. The system can be considered a VVO 100. The VVO 100 is a vertical, ventilated dry spent fuel storage system that is fully compatible with 100 ton and 125 ton transfer casks for spent fuel canister operations. Of course, the VVO 100 can be modified/designed to be compatible with any size or style transfer cask. The VVO 100 is designed to accept spent fuel canisters for storage. All spent fuel canister types engineered for storage in free-standing and anchored overpack models can be stored in VVO 100. As used herein the term “canister” broadly includes any spent fuel containment apparatus, including, without limitation, multi-purpose canisters and thermally conductive casks. For example, in some areas of the world, spent fuel is transferred and stored in metal casks having a honeycomb grid-work/basket built directly into the metal cask. Such casks and similar containment apparatus qualify as canisters, as that term is used herein, and can be used in conjunction with VVO 100 as discussed below. In certain embodiments, the VVO 100 is a substantially cylindrical containment unit having a vertical axis A-A and a horizontal cross-sectional profile that is substantially circular in shape. Of course, it should be understood that the invention is not limited to cylinders having circular horizontal cross sectional profiles but may also include containers having cross-sectional profiles that are, for example, rectangular, ovoid or other polygon forms. While the VVO 100 is particularly useful for use in conjunction with storing and/or transporting SNF assemblies, the invention is in no way limited by the type of waste to be stored. The VVO cask 100 can be used to transport and/or store almost any type of HLW. However, the VVO 100 is particularly suited for the transport, storage and/or cooling of radioactive materials that have a high residual heat load and that produce neutron and gamma radiation, such as SNF. This is because the VVO 100 is designed to both provide extreme radiation blockage of gamma and neutron radiation and facilitate a convective/no force cooling of any canister contained therein. The VVO 100 of the present invention generally comprises an overpack body 110 for storing high level radioactive waste and a removable overpack lid 120 that is positioned atop the overpack body 110. The overpack body 110 extends along the vertical axis A-A. The overpack lid 120 generally comprises a primary lid 121 and a secondary lid 122. The primary lid 121 is secured to the overpack body 110 by bolts 123 that restrain separation of the primary lid 121 of the overpack lid 120 from the overpack body 110 in case of a tip over situation. Moreover, the secondary lid 122 is secured to the primary lid 121 by bolts 124. The overpack lid 120 is a steel/concrete structure that is equipped with an axisymmetric air outlet vent or passageway 145 for the ventilation/removal of air as will be discussed in more derail below. An annular opening 157 is formed in an outer sidewall surface 178 of the overpack lid 120 that forms a passageway from the air outlet vent 145 to the external environment. More specifically, the annular opening 157 is a 360° opening in the outer sidewall surface 178 of the overpack lid 120. The overpack lid 120 has a quick connect/disconnect joint to minimize human activity for its installation or removal. In certain embodiments, the overpack lid 120 may weigh in excess of 15 tons. The VVO 100 further comprises shock absorber or crush tubes 102 in its top region. The shock absorber tubes 102 are arranged at suitable angular spacings to serve as a sacrificial crush material if, for any reason, the VVO 100 were to tip over. The shock absorber tubes 102 also facilitate guiding and positioning of a canister within a cavity 111 of the VVO 100 in a substantially concentric disposition with respect to the VVO 100. Referring to FIGS. 1, 4 and 6 concurrently, the overpack body 110 comprises a cylindrical wall 112, a bottom enclosure plate 130 and the overpack lid 120 described above. The cylindrical wall 112 has an inner shell 113, an intermediate shell 114 and an outer shell 115. In the exemplified embodiment, each of the inner, intermediate and outer shells 113, 114, 115 are formed of one-inch thick steel. Of course, the invention is not to be so limited and in other embodiments the inner, intermediate and outer shells 113, 114, 115 can be formed of metals other than steel and can be greater or less than one-inch in thickness. The inner shell 113 has an inner surface 116 that defines an internal cavity 111 for containing a hermetically sealed canister that contains high level radioactive waste (FIG. 5). The inner surface 116 of the inner shell 113 also forms the inner wall surface of the overpack body 110. Furthermore, the outer shell 115 has an outer surface 117. The outer surface 117 of the outer shell 115 also forms the outer sidewall surface of the overpack body 110. In the exemplified embodiment, the inner, intermediate and outer shells 313, 114, 115 are concentric shells that are rendered into a monolithic weldment by a plurality of connector plates 105a, 105b. The inner shell 113 is spaced from the intermediate shell 114 by connector plates 105a and the intermediate shell 114 is spaced from the outer shell 115 by connector plates 105b. Of course, in certain other embodiments the connector plates 105a, 105b can be altogether omitted. The space between the inner shell 113 and the intermediate shell 114 is intended for placement of a neutron shielding material. For example, in certain embodiments the neutron radiation shielding material is a hydrogen-rich material, such as, for example, Holtite, water or any other material that is rich in hydrogen and a Boron-10 isotope. In certain embodiments, there is approximately seven inches of Holtite filling the space between the inner and intermediate shells 113, 114. Thus, the space between the inner and intermediate shells 113, 114 serves to prevent neutron radiation from passing through the VVO 100 and into the external environment. An axially intermediate portion of the space between the intermediate shell 114 and the outer shell 115 is filled with a heavy shielding concrete to capture and prevent the escape of both gamma and neutron radiation. The density of the concrete is preferably maximized to increase the radiation absorption characteristics of the VVO 100. In certain embodiments, there is approximately twenty-eight inches of concrete filling the intermediate portion of the space between the intermediate and outer shells 114, 115. In some embodiments, steel plates are placed within the concrete to serve as a supplemental radiation curtain. There are no lateral penetrations in the multi-shell weldment that may provide a streaming path for the radiation issuing from the high level radioactive waste. The top and bottom portions of the space between the intermediate and outer shells 114, 115 (both above and below the concrete) are top and bottom forgings 128, 129 in the form of thick annular rings made of a metal material, such as steel. The top forging 128 comprises machine threaded holes 126 that are sized and configured to receive the bolts 123 of the primary lid 121 therein during attachment of the overpack lid 120 to the overpack body 110. As noted above, the inner surface 116 of the inner shell 113 defines the cavity 111. In the exemplified embodiment, the cavity 111 is cylindrical in shape. However, the cavity 111 is not particularly limited to any specific size, shape, and/or depth, and the cavity 111 can be designed to receive and store almost any shape of canister. In certain embodiments, the cavity 111 is sized and shaped so that it can accommodate a canister of spent nuclear fuel or other HLW. More specifically, the cavity 111 has a horizontal cross-section that can accommodate no more than one canister. Even more specifically, it is desirable that the size and shape of the cavity 111 be designed so that, when a spent fuel canister is positioned in the cavity 111 for storage, a small clearance exists between outer side walls of the canister and the inner surface 116 of the inner shell 113, as will be discussed in more detail below with reference to FIG. 5. Referring to FIGS. 4 and 5 concurrently, the present invention will be further described. The cavity 111 comprises a floor 152 and an open top end 151 that is enclosed by the overpack lid 120 as has been described herein above. A plurality of support blocks 153 are disposed on the floor 152 of the cavity 111 to support a canister 200 contained within the cavity 111 above the floor 152. In the exemplified embodiment, four support blocks 153 are illustrated (see FIG. 6). However, more or less than four support blocks 153 can be used in alternate embodiments. Each of the support blocks 153 is a low profile lug that is welded to the inner surface 116 of the inner shell 113 and/or to the floor 152. In the exemplified embodiment, the canister 200 is a hermetically sealed canister for containing the high level radioactive waste. When the canister 200 is positioned within the cavity 111, it rests atop the support blocks 153 so that a space 154 exists between a bottom 202 of the canister 200 and the floor 152. The space 154 is a bottom plenum that serves as the recipient of ventilation air flowing up from an inlet vent as will be described below. Furthermore, when the canister 200 is positioned within the cavity 111, an annular gap 155 exists between the inner surface 116 of the inner shell 113 (i.e., the inner wall surface of the overpack body 110) and an outer surface 201 of the canister 200. The annular gap 155 is an uninterrupted and continuous gap that circumferentially surrounds the canister 200. In other words, the canister 200 is concentrically spaced apart from the inner shell 113, thereby creating the annular gap 155. As described in more detail below, the annular gap 155 forms an annular air flow passageway between an annular air inlet passageway 142 and the air outlet vent 145. The VVO 100 is configured to achieve a cyclical thermosiphon flow of gas (i.e., air) within the cavity 111 when spent nuclear fuel emanating heat (i.e., the canister 200) is contained therein. In other words, the VVO 100 achieves a ventilated flow by virtue of a chimney effect. Such cyclical thermosiphon flow of the gas further enhances the transmission of heat to the environment external to the VVO 100. The thermosiphon flow of gas is achieved as a result of an air inlet vent 140 that introduces cool air into the bottom of the cavity 111 of the overpack body 110 from the external environment and an air outlet vent 145 for removing warmed air from the cavity 111. Thus, as a result of thermosiphon flow, cool external air can enter into the space 154 of the cavity 111 between the bottom 202 of the canister 200 and the floor 152 via the air inlet vent 140, flow upward through the cavity 111 within the annular gap 155 between the canister 200 and the inner surface 116 of the inner shell 113, and flow back out into the external environment as warmed air via the air outlet vent 145. The newly entered air will warm due to proximity to the extremely hot canister 200, which will cause the natural thermosiphon flow process to take place whereby the heated air will continually flow upwardly as fresh cool air continues to enter into the cavity 111 via the air inlet vent 140. Thus, the air inlet vent 140 provides a passageway that facilitates cool air entering the cavity 111 from the external environment and the air outlet vent 145 provides a passageway that facilitates warm air exiting the cavity back to the external environment. In the exemplified embodiment, the air outlet vent 145 is formed into the overpack lid 120. The air outlet vent 145 provides an annular passageway from a top portion of the cavity 111 to the external environment when the overpack lid 120 is positioned atop the overpack body 110 thereby enclosing the top end 151 of the cavity 111. Specifically, the air outlet vent 145 has a vertical section 174 that extends from the cavity 111 upwardly into the overpack lid 120 in the vertical direction (i.e., the direction of the vertical axis A-A) and a horizontal section 175 that extends from the vertical section 174 to the annular opening 157 in the horizontal direction (i.e., the direction transverse to the vertical axis A-A). More specifically, the vertical section 174 of the air outlet vent 145 extends from an annular opening 176 in a bottom surface 177 of the overpack lid 120 and the horizontal section 175 extends from the vertical section 174 to the annular opening 157 in the outer sidewall surface 178 of the overpack lid 120. As described above, the annular opening 157 is a circumferential opening that extends around the entirety of the overpack lid 120 in a continuous and uninterrupted manner and circumferentially surrounds the vertical axis A-A. The overpack body 110 additionally comprises a bottom block 160 disposed within the cylindrical wall 112, and more specifically within the inner shell 113 of the cylindrical wall 112, and a base structure at a bottom end 179 of the cylindrical wall 112. The base structure comprises a base plate 161 and an annular plate 162. The air inlet vent 140 is formed directly into the bottom block 160, which is a thick sandwich of steel and concrete. The bottom block 160 is positioned below the floor 152 of the cavity 111. More specifically, the bottom block 160 extends between the floor 152 of the cavity 111 and the base plate 161, which forms the bottom end of the VVO 100. The bottom block 160 has a columnar portion 163 and a horizontal portion 164. The annular plate 162 is a donut-shaped plate having a central hole 181. The annular plate 162 is axially spaced from the base plate 161, thereby creating a space or gap in between the annular plate 162 and the base plate 161. Moreover, the annular plate 162 extends from the outer surface 117 of the overpack body 110 inwardly towards the vertical axis A-A a radial distance that is less than the radius of the overpack body 110. More specifically, the annular plate 162 extends from the outer surface 117 of the overpack body 110 to the columnar portion 163 of the bottom block 160. Thought of another way, the columnar portion 163 of the bottom block 160 extends through the central hole 181 of the annular plate 162 and rests atop the base plate 161. Referring to FIGS. 1, 4, 6 and 8 concurrently, the air inlet vent 140 will be described in more detail. In the exemplified embodiment, the air inlet vent 140 is formed into the bottom closure plate 130 and extends into the bottom block 160 and comprises an annular air inlet plenum 141 and an annular air inlet passageway 142. The annular air inlet plenum 141 is formed in the space/gap between the annular plate 162 and the base plate 161. Thus, the annular air inlet plenum 141 is substantially horizontal and extends radially inward from the outer surface 117 of the overpack body 110. More specifically, the annular air inlet plenum 141 extends horizontally from the outer surface 117 of the overpack body 110 at an axial height below the floor 152 of the cavity 111. An opening 143 is formed in the outer surface 117 of the overpack body 110 that forms a passageway from the external environment to the annular air inlet plenum 141 to enable cool air to enter into the annular air inlet plenum 141 from the external environment as has been described above. The opening 143 circumferentially surrounds the vertical axis A-A around the entirety of the outer surface 117 of the overpack body 110 in an uninterrupted and continuous manner. In other words, the opening 143 is a substantially 360° opening in the outer surface 117 of the overpack body 110. The annular air inlet passageway 142 extends upward from a top surface 144 of the annular air inlet plenum 141 to the floor 152 of the cavity 111. More specifically, the annular air inlet passageway 142 extends upwardly from an opening 147 in the top surface 144 of the annular air inlet plenum 141 to an opening 146 in the floor 152. The annular air inlet passageway 142 is wholly formed within the bottom block 160. The opening 147 in the top surface 144 of the annular air inlet plenum 141 is proximate an end of the annular air inlet plenum opposite the opening 143 in the outer surface 117 of the overpack body 110. The opening 146 in the floor 152 is an annular opening that extends 360° around the floor 152. The annular air inlet plenum 141 circumferentially surrounds the vertical axis A-A. In the exemplified embodiment, the annular air inlet passageway 142 also circumferentially surrounds the vertical axis A-A and has an inverted truncated cone shape. Thus, the annular air inlet passageway 142 extends upward from the air inlet plenum 141 to the opening 146 in the floor 152 of the cavity 111 at an oblique angle relative to the vertical axis A-A. Thought of another way, the annular inlet passageway 142 extends from the air inlet plenum 141 at a first end 183 to the floor 152 at a second end 184. The first end 183 is located a first radial distance R1 from the vertical axis A-A and the second end 184 is located a second radial distance R2 from the vertical axis A-A. The second radial distance R2 is greater than the first radial distance R1. Of course, the invention is not to be so limited and in certain other embodiments the annular air inlet passageway 142 can take on other shapes as desired. Referring to FIGS. 1, 4, 7 and 8 concurrently, the annular air inlet plenum 141 will be further described. The annular air inlet plenum 141 comprises a plurality of plates 148 therein. Each of the plates 148 extends from a first end 149 to a second end 159. The first ends 149 of the plates 148 are proximate the outer surface 117 of the overpack body 110 and the second ends 159 of the plates 148 are proximate the columnar portion 163 of the bottom block 160. A line connecting the first ends 149 of the plates 148 forms a first reference circle 171 having a diameter D1 and a line connecting the second ends 159 of the plates 148 forms a second reference circle 172 having a diameter D2, wherein the first diameter D1 is greater than the second diameter D2. Each of the plates 148 in the annular air inlet plenum 141 extend along a reference line 169 that is tangent to a third reference circle 170. Although the reference line 169 is only illustrated with regard to two of the plates 148, it should be understood that each of the plates has a reference line that is tangent to the third reference circle 170. The circumference of the third reference circle 170 is formed by an outer surface 165 of the columnar portion 163 of the bottom block 160. The third reference circle 170 has a center point that is coincident with the vertical axis A-A. In the exemplified embodiment, the plates 148 are thin steel plates that facilitate transferring the weight of the VVO 100 to the base plate 161 and also provide a means to scatter and absorb any errant gamma radiation that may attempt to exit the air inlet plenum. Furthermore, in the exemplified embodiment sixty plates 148 are illustrated. However, the invention is not to be so limited and in certain other embodiments more or less than sixty plates 148 may be disposed within the annular air inlet plenum 141. Due to the axisymmetric configuration of the air inlet plenum 141, the annular air inlet vent 140 is configured so that aerodynamic performance of the air inlet vent 140 is independent of an angular direction of a horizontal component of an air-stream applied to the outer surface 117 of the overpack body 101. Similarly, due to the axisymmetric configuration of the air outlet vent 145, the air outlet vent 145 is configured so that the aerodynamic performance of the air outlet vent 145 is independent of an angular direction of a horizontal component of an air-stream applied to the outer surface 117 of the overpack body 110. With reference to FIGS. 9-26, a second inventive concept will be described. The present invention provides a separate, reusable shielded lifting lid for waste canister lid bolting and lifting. Accordingly, the lifting lid is bolted and not welded to the canister. The canister loading is dry in an overpack such as a metal cylindrical jacket holding the radioactive waste inside. Canisters typically have thick (e.g. 10 inch) steel lids on each canister to protect the operator from radiation during canister closure operations. The thick lids are heavy and expensive, and further not reusable as they remain attached to the canister for longer-term storage. Advantageously, the present invention allows use of a significantly thinner main closure confinement lid (e.g. about 3 to 5-inch thick in exemplary embodiments) for radionuclides containment. After radioactive waste contents are placed in the canister, the confinement lid is installed and held in place by gravity alone in some embodiments. The confinement lid thickness, however, has generally poor radiation shielding value. Accordingly, the confinement lid is installed using a thicker and reusable shielded lifting lid which serves as an upper over-lid to the lower confinement lid. The two-part lid system combination of the confinement lid and shielded lifting lid provide the thickness required to shield the operator from the radioactive canister contents during the canister closure bolting operations. In use, the shielded lifting lid in one exemplary and non-limiting embodiment has holes that match the bolt spacing to allow the operator to install the confinement lid bolts in a radiation shielded environment. After the lifting lid bolts are installed, the operator hooks up the lifting rigging to the shielded lifting lid and moves away from the canister to a more distal and remote location. The shielded lifting lid may then be removed from the top of the canister, preferably with the confinement lid remaining in place, and a heavy overpack lid is installed for longer term storage and radiation shielding. Using this method, the waste canister and overpack advantageously are shorter, lighter, better shielded, and less expensive to fabricate. FIGS. 9 and 10 depict a radioactive canister system according to the present disclosure including a waste canister 1100 having a generally cylindrical body defining an interior chamber 1101 and comprised of a top 1102, bottom 1104, and cylindrical sidewall 1106 extending therebetween. Top 1102 is open for insertion of radioactive waste and bottom 1104 is preferably closed in one embodiment. A main closure confinement lid 1200 is shown attached to top 1102 of canister 1100 by a plurality of fasteners such as mounting bolts 1154 which may be circumferentially spaced apart around the top of the canister, as further described herein. In one embodiment, canister 1100 may be a non-fuel radioactive waste canister (NWC). Referring to FIG. 10, canister 1100 has an interior configured to store the size and shape of radioactive waste to be deposited in the canister. In one embodiment the canister may include a basket insert 1120 configured for holding a plurality of metal waste cylinders 1121 (see. e.g. FIG. 14) each containing radioactive waste materials. Basket insert 1120 includes a pair of vertically spaced apart top and bottom plates 1122, 1124 which are connected via a plurality of tie rods 1126. Top plate 1122 and bottom plate 1124 include a plurality of horizontally spaced apart circular openings 1123 each having a diameter which is configured and dimensioned to receive waste cylinders 1121 therethrough, as shown in FIG. 14. Referring to FIGS. 10 and 11, the top portion of tie rods 1126 may be threaded for attachment to top plate 1122 by a threaded nut 1125. Top plate 1122 may be spaced by a vertical distance below the top 1102 of canister 1100. Bottom plate 1124 may be elevated by a vertical distance above the bottom 1104 of canister 1100 by a plurality of vertical tubular sleeves 1128 having a bottom end resting on bottom 1104 of the canister 1100 and a top end attached to bottom plate 1124 as better shown in FIG. 12. In one embodiment, sleeves have an inside diameter sized to receive the bottom end portion of tie rods 1126 which are slidably received in the sleeves. This provides for vertical adjustment in the height of the basket insert 1120 to accommodate the height of waste cylinders 1121 to be stored inside canister 1100. Bottom plate 1124 remains fixed and stationary in position. The top plate 1122 with attached tie rods 1126, however, is movable upwards and downwards with respect to the canister and bottom plate 1124 to reach a desired position depending on the height of waste cylinders 1121. In some embodiments, the top plate 1122 may be thereafter be fixed in the desired position after vertical adjustments are made by securing the top plate to the interior of the canister sidewall 1106 such as by welding or other suitable means. Accordingly, adjustable basket insert 1120 may accommodate a variety of waste cylinder heights. Basket insert 1120 (i.e. top plate, bottom plate, tie rods, etc.) may be made of any suitable material, including without limitation a corrosion resistant metal such as stainless steel in one embodiment. FIG. 13 shows canister 1100 loaded into an outer overpack 1130 for transport and storage of radioactive waste. The overpack provides protection during transport and storage of the waste by encapsulating the waste canister in an outer protective jacket. Overpack 1130 has an open top 1132, and is configured and dimensioned to completely receive canister 1100 through the top 1102. Overpack 1130 has an open interior defining an interior surface 1133 and an exterior surface 1135 (see also FIG. 17). Overpack 1130 is generally cylindrical in shape further including a cylindrical sidewall 1134 and flat closed bottom 1136 (see FIG. 23) configured for resting on a flat surface such as concrete slab. Preferably, in one embodiment, overpack 1130 has a greater height than canister 1100 so that the canister is recessed below the open top 1132 of the overpack when fully inserted therein. Overpack 1130 may be made of any suitable material or combination of materials (see, e.g. FIG. 17) which may include neutron absorbing materials such as without limitation concrete, lead, or boron. An example of a suitable overpack for use with canister 1100 may be a HI-SAFE™ transport overpack as used in vertical non-fuel waste storage systems available from Holtec International of Marlton, N.J. The sidewalls 1134 forming the spaced apart cylindrical walls that define an annular space between the inner and outer surfaces 1133 and 1135 respectively may be formed of a corrosion resistant metal also selected for strength to protect the inner canister 1100, such as stainless steel as one non-limiting example. The neutron absorbing material may be disposed between the inner and outer surfaces 1133 and 1135. In some embodiments, overpack 1130 may also include Metamic® for radiation shielding which is a discontinuously reinforced aluminum/boron carbide metal matrix composite material also available from Holtec International. Referring to FIGS. 10-11 and 13, the top of the canister 1100 may include a peripheral contamination boundary seal which cooperates with the confinement lid 1200 to prevent leakage of radiation from the canister, particularly at the lid bolting locations. In particular, the boundary seal shields the mounting blocks 1150 to prevent radiation streaming. In one embodiment, the boundary seal may be configured as an annular shielding flange 1140 that extends circumferentially around the upper peripheral edge of the top 1102 of the canister. Confinement lid 1200 rests on the shielding flange when bolted to the canister 1100. Shielding flange 1140 may be horizontally flat and extend inwards in a direction perpendicular to and from sidewall 1106 towards the vertical axial centerline CL of the canister 1100. In one embodiment, shielding flange 1140 is attached to the uppermost top edge of the sidewall 1106 as shown. Shielding flange 1140 may have an at least partially scalloped configuration in top plan view in some embodiments as shown to accommodate insertion of waste cylinders 1121 into the canister. According, the scallops 1142 if provided are preferably concentrically aligned with the circular openings 1123 in basket insert 1120 in top plan view. This minimizes the required diameter of the canister 1100 for holding the waste cylinders 1121. In other possible embodiments, however, shielding flange 1140 may have an uninterrupted shape forming a continuous ring in top plan view. At the lid bolting locations, shielding flange 1140 is configured to cover a with a plurality of mounting blocks 1150 which are circumferentially spaced around the interior of canister 1100 disposed adjacent to sidewall 1106 to provide a radiation-shielded bolting system for attaching confinement lid 1200 and shielded lifting lid 1300 to the canister. Shielding flange 1140 may be formed of any suitable material including metals which are corrosion resistant such as stainless steel. With continuing reference to FIGS. 10-11 and 13, mounting blocks 1150 may have a generally arcuate and curved shape in top plan view which complements the inside radius of curvature of the sidewall 1106 to which mounting blocks 1150 may be attached. Mounting blocks 1150 may be rigidly/fixedly attached to the canister sidewall 1106 by a suitably strong mechanical connection capable of supporting at least the entire dead weight of canister 1100 and basket insert 1120 for lifting and loading the canister into overpack 1130. Accordingly, in one preferred embodiment, mounting blocks 1150 are welded to at least sidewall 1106 of the canister body for strength. In some embodiments, the mounting blocks 1150 may be abutted against but are not fixedly connected to the underside of radiation shielding flange 1140 so that lifting loads are not transferred to the flange directly but rather bypass the flange to the mounting blocks 1150 via the bolting provided. Any suitable number of mounting blocks 1150 may be provided; the number and circumferential spacing being dependent on the magnitude of the structural load imparted to the blocks dependent on whether the canister 1100 will be lifted in an empty condition or in a fully loaded condition with filled waste cylinders 1121 positioned in the canister. It is well within the ambit of those skilled in the an to determine an appropriate number and circumferential spacing of the mounting blocks 1150. In one embodiment, the mounting blocks 1150 are each configured for both lifting canister 1100 and attaching both the lower confinement lid 1200 and upper lifting lid 1300. As best shown in FIGS. 11 and 17, mounting blocks 1150 each include a plurality of threaded mounting sockets 1152 for forming a threaded connection with complementary threaded mounting bolts 1154 and 1156 used for attaching confinement lid 1200 and shielded lifting lid 1300 respectively to the canister 1100. In one non-limiting example, three threaded mounting sockets 1152 may be provided in each mounting block. However, other suitable numbers of mounting sockets may be used. In certain embodiments, the mounting sockets 1152 extend only partially into the mounting blocks 1150 as shown. Radiation shielding flange 1140 includes mating holes 1144 which are each concentrically aligned with the threaded mounting sockets 1152 of the mounting block to provide access for mounting bolts 1154, 1156 to the mounting sockets in the block. Because shielding flange 1140 in some embodiments in not intended to be a load-bearing member relied upon for lifting the canister, holes 1144 may not be threaded so that the weight of the canister is transferred, through the flange via the mounting bolts 1156 to the shielded lifting lid 1300. In one embodiment, mounting bolts 1154 and/or 1156 may be threaded bolts having an integral or separate washer disposed adjacent to the head, as best shown in FIG. 19. Mounting bolts 1154 are used for attaching the lower confinement lid 1200 to canister 1100 via mounting blocks 1150. In one embodiment, mounting bolts 1154 are not used for lifting the canister 1100 but rather for lid securement. By contrast, mounting bolts 1156 serve a dual purpose and may be used for both attaching the lower shielded lifting lid 1300 to canister 1100 and supporting the weight of the canister during lifting operations via mounting blocks 1150 engaged by bolts 1156. In one preferred embodiment, mounting bolts 1156 may have a longer shank than mounting bolts 1154 as shown. This arrangement ensures that the depth of threaded engagement between the threaded mounting sockets 1152 of the mounting blocks 1150 and mounting bolt 1156 is sufficient for lifting the canister 1100, as further explained herein. The confinement lid 1200 is generally circular in shape (top plan view) and shown in FIGS. 8, 17, and 19. Confinement lid 1200 includes a plurality of bolt holes 1202 spaced circumferentially around the peripheral side 1204 of the lid as best shown in FIG. 9 (including at locations where mounting bolts 1154 are shown installed). Bolt holes 1202 penetrate top surface 1206 of the confinement lid, and in one embodiment are not threaded. The bolt holes 1202 may be arranged in groups corresponding to the location and arrangement of the mounting blocks 1150 inside the canister 1100. The bolt holes 1202 have a diameter sized to at least pass the shank of mounting bolts 1154 and 1156 through the holes to threadably engage the mounting blocks 1150. Accordingly, some of the bolt holes 1202 are configured to receive the shanks of the confinement lid mounting bolts 1154 and others are configured to receive the shank of shielded lifting lid mounting bolts 1156. In cases where the mounting bolts 1154 and 1156 have shanks of the same diameter, the bolt holes 1202 may all have the same diameter. Where the shanks of bolts 1154 and 1156 are different in diameter, the holes 1202 may have correspondingly different diameters for each bolt. The confinement lid 1200 may have a uniform thickness front peripheral side 1204 to peripheral side 1204 as best shown in FIG. 17 in one embodiment. In other embodiments, the thickness may vary at different locations on the lid 1200. Confinement lid 1200 may be made of any suitable material, preferably an appropriate metal for the application. In an exemplary embodiment, without limitation, the confinement lid 1200 for example may be made of stainless steel for corrosion resistance. The upper shielded lifting lid 1300 is not intended to remain on canister 1100 for longer term waste storage. Instead, in some embodiments the lifting lid 1300 is configured and structured for transporting and initially lifting the canister 1100 into position in the cylindrical overpack 1130 prior to loading the waste cylinders 1121 after which the lifting lid is removed, and then after the waste cylinders are loaded in the canister, the lifting lid is replaced on the canister to shield the operator for bolting the lower confinement lid 1200 in place after which the lifting lid is removed again. It will be appreciated that this scenario for using the shielded lifting lid 1300 may be varied in other embodiments. Referring to FIGS. 15-20, shielded lifting lid 1300 is generally circular in shape (top plan view) and includes a plurality of bolt holes 1302 spaced circumferentially around the peripheral side 1304 of the lid as best shown in FIG. 9. In one embodiment, holes 1302 are not threaded. The bolt holes 1302 may be arranged in clustered groups or sets corresponding to the location and arrangement of the mounting blocks 1150 inside the canister 1100. The bolt holes 1302 have a diameter sized to at least pass the shank of mounting bolts 1154 and 1156 through the holes to threadably engage the mounting blocks 1150. Accordingly, some of the bolt holes 1302 are configured to receive the shanks of the confinement lid mounting bolts 1154 and others are configured to receive the shank of shielded lifting lid mounting bolts 1156. In cases where the mounting bolts 1154 and 1156 have shanks of the same diameter, the bolt holes 1302 may all have the same diameter. Where the shanks of bolts 1154 and 1156 are different in diameter, the holes 1302 may have correspondingly different diameters for each bolt. According to another aspect of the invention, bolt holes 1302 have different diameters in one embodiment even if the mounting bolts 1154, 1156 are used have the same shank diameter. The confinement lid mounting bolts 1154 need not engage the upper shielded lifting lid because bolts 1154 are only required to secure the lower confinement lid to canister 1100. Accordingly, in the embodiment shown in FIG. 19, the bolt holts 1302 for the confinement lid mounting bolts 1154 may have a larger diameter than the bolt holes 1302 for the lifting lid mounting bolts 1156. In this arrangement, the bolt holes 1302 for the confinement lid mounting bolts 1154 are sized with a diameter large enough to allow the shank and entire head of bolts 1154 to pass through the bolt holes so that the head and integral washer directly engage the top surface 1200 of the confinement lid 1200 (see, e.g. FIG. 9). When completely installed the heads of the mounting bolts 1154 are recessed below the top surface of the lifting lid 1300 as shown. By contrast, since the mounting bolts 1156 for the lifting lid 1300 also serve a lifting function for the canister 1100, the bolt holes 1302 have a diameter sized so that the leads of bolts 1156 do not pass through the bolt holes and instead engage the top surface 1306 of the lifting lid (thereby projecting above the top surface and remaining exposed as shown in FIG. 19). In this manner, the bolts 1156 transfer the dead load and weight of the canister 1100 from the mounting blocks 1150 directly to the shielded lifting lid 1300 without involvement of the confinement lid 1200. Accordingly, to accommodate the foregoing arrangement, the lifting lid mounting bolts 1156 preferably have a longer shank than the confinement lid mounting bolts 1154 in this embodiment. As shown in FIGS. 17 and 18, several spaced apart clusters comprised of three bolt holes 1302 may be provided in the non-limiting embodiment shown which are spaced circumferentially around and proximate to the peripheral side 1304 of the shielded lifting lid 1300. Each cluster of bolt holes 1302 is spaced apart by an arcuate distance from adjacent clusters of holes 1302. The clusters of bolts holes 1302 are each vertically aligned with a corresponding mounting block 1150 (see also FIG. 11). In this embodiment, the center hole 1302 has a smaller diameter for the lifting lid mounting bolt 1156 than the two adjacent outer holes 1302 have larger diameters for the confinement lid mounting bolts 1154. Other suitable arrangements of holes 1302 may be provided. The bolt holes 1202 in the confinement lid 1200 may also arranged in clusters of three to mate with the bolt holes 1302 of the lifting lid 1300. All three of the bolt holes 1202 in each cluster in the confinement lid, however, may have the same diameter. Advantageously, having two different size bolt holes 1302 for the confinement lid mounting bolts 1154 and the lifting lid mounting bolts 1156 reduces possible installation error and ensures that the operator will not confuse which holes are intended for each. This plays a role in deploying the two-part lid system when the confinement lid 1200 and its respective bolts 1154 are eventually left in place after bolting the confinement lid to the canister 1100 and the lifting lid mounting bolts 1156 are removed by the operator, as further described herein. The shielded lifting lid 1300 may have a non-uniform thickness from peripheral side 1304 to peripheral side 1304 as best shown in FIG. 17. Accordingly, in one possible embodiment as shown, the peripheral portion of lifting lid 1300 may include an outer annular step or shoulder 1308 having a smaller thickness than the inner central portion 1314 of the lid. The shoulder 1308 is configured to complement and abuttingly engage a corresponding top annular rim 1138 of the overpack 1130 such that portions of the lifting lid 1300 adjacent to peripheral side 1304 overlap the top of the rim to prevent radiation streaming as shown. Rim 1138 therefore defines an annulus for receiving shoulder 1308. Accordingly, as shown in FIG. 17, shielded lifting lid 1300 has a larger diameter than confinement lid 1200 to account for the overlap with the annular rim 1138 of the overpack 1130. The central portion 1314 of the lifting lid 1300 preferably has a thickness and a diameter sized to allow at least partial insertion of the central portion into the overpack 1130 such that the outwards facing annular sides of the central portion abuts the interior surface 1133 of the overpack as shown. This arrangement further prevents radiation streaming from the canister 1100 when the lifting lid 1300 is in place on the canister. Because shielded lifting lid 1300 serves a structural purpose for lifting the canister 1100, the lifting lid preferably has a thickness which is greater than the confinement lid 1200. In one embodiment, the lifting lid has a thickness which is at least twice the thickness of the confinement lid. Shielded lifting lid 1300 may be made of any suitable material, preferably an appropriate metal for the application. In exemplary embodiments, without limitation, the lilting lid 300 for example may be made of carbon steel or stainless steel. Referring to FIGS. 15 and 16, the lower confinement lid 1200 is detachably mounted to upper shielded lifting lid 1300 so that the lid assembly 1200/1300 may be lifted and moved as a single unit as shown with the lifting lid supporting the confinement lid when not attached to the canister 1100. When needed during the canister closure operations, the lifting lid 1300 may be uncoupled from the confinement lid 1200. In one embodiment, a plurality of circumferentially spaced fasteners such as threaded assembly bolts 1131 may be provided to attach lifting lid 1300 to confinement lid 1200. Assembly bolts 1131 which are inserted through the lifting lid 1300 and engage complementary threaded sockets 1208 (shown in FIG. 9) formed in the confinement lid (such arrangement and operation being apparent to those skilled in the art without further elaboration). A suitable number of assembly bolts 1131 are provided to support the lower confinement lid 1200 front the upper shielded lifting lid 1300 during hoisting. Accordingly, confinement lid 1200 may be considered to be fully supported by the lifting lid 1300 during lifting of the lid assembly 1200/1300. As shown in FIGS. 15 and 16, shielded lifting lid 1300 includes a lifting attachment such as lifting lugs 1402 and pin 1404 for grappling and hoisting the lid. Other suitable lifting attachments configured for grappling such as for example lifting bails may be used. An exemplary method for storing radioactive waste using the present container system with two-part lid assembly 1200/1300 (confinement lid 1200, lifting lid 1300) according to the present disclosure will now be described. As a preliminary step, the lower confinement lid 1200 is detachably mounted to the upper shielded lifting lid 1300 using assembly bolts 1131 to collectively form the lid assembly 1200/1300, shown in FIG. 15. Referring to FIGS. 9 and 10, the method begins with a canister 1100 first being provided with an empty basket insert 1120 disposed inside the canister as shown. Next, the empty canister 1100 is lifted and placed into the overpack 1130 as shown in FIG. 13. In one embodiment, this step may be performed by bolting the lid assembly 1200/1300 to canister 1100 using the mounting bolts 1156 to threadably engage the mounting blocks 1150, and grappling and attaching a hoist 1400 to the upper lifting lid 1300 using lifting lugs 1402 and pin 1404 as shown in FIG. 15. The hoist 1400 may be part of the lifting equipment such as a crane or other suitable equipment operable to raise and lower the canister. After positioning the basket insert 1120 into the canister 1100, the mounting bolts 1156 may be removed to disconnect the canister from the lid assembly. The lid assembly 1200/1300 may then be lifted by the hoist and removed (see FIG. 13). Next, one or preferably more lid alignment pins 1406 may be threaded into some of the threaded sockets 1152 of the mounting block to eventually help properly align the lid assembly 1200/1300 with the canister (see FIG. 13). In one non-limiting example, three alignment pins 1406 are used spaced apart on the canister. The alignment pins 1406 are preferably installed locally by an operator prior to loading the radioactively “hot” waste cylinders 1121 into the canister. Following installation of the alignment pins 1406, the waste cylinders 1121 are loaded into the canister 1100, and more specifically positioned in then respective locations provided in basket insert 1120 as shown in FIG. 14. Loading of the waste cylinders is performed remotely (i.e. at a distance) by an operator using suitable equipment to protect the operator from radiation. After loading the waste cylinders 1121, the lid assembly 1200/1300 is remotely hoisted by the operator over and vertically positioned above the top 1102 of the canister 1100, as shown in FIG. 15. Using the lid alignment pins 1406, the operator vertically aligns holes 1302 in shielded lifting lid (with holes 1202 in confinement lid being concentrically aligned with holes 1302) with corresponding pins 1406 to properly orient the lid rotationally with respect to the canister. When the pins 1400 and their corresponding holes have been axially aligned, the operator lowers lid assembly 1200/1300 onto the canister 1100 as shown in FIG. 16 (see pins 1406 extending through holes 1302). The operator will now be shielded from radiation emitted from the canister so that the confinement lid 1200 may be bolted in place locally. Next, the lid alignment pins 1406 and assembly bolts 1131 which hold the lower confinement lid 1200 to upper shielded lifting lid 1300 may be removed (see, e.g. FIG. 18). All of the confinement lid mounting bolts 1154 may then be installed to mount the confinement lid 1200 to the canister 1100 using the mounting blocks 1150. The mounting bolts 1154 are threaded through bolt holes 1302 until the heads of the bolts engage the top surface 1206 of the confinement lid 1200 and the bolts are tightened to the required torque (see FIGS. 19 and 20). Prior to removing the shielded lifting lid 1300, a set of overpack lid alignment pins 1408 may next be installed in threaded sockets 1510 of the overpack 1130. With the confinement lid 1200 now fully fastened to canister 1100, the shielded lifting lid 1300 may next be removed via the hoist remotely by an operator as shown in FIG. 23. In the following steps, the overpack lid 1500 is installed on overpack 1130 following closure of canister 1100 described above. FIG. 23 shows the shielded lifting lid 1300 being removed and the overpack lid 1500 staged for installation. FIG. 21 shows overpack lid 1500 in greater detail. Overpack lid 1500 is circular in shape (top plan view) and includes a plurality of mounting holes 1502, top surface 1504, peripheral sides 1506, and a lifting bail 1508 attached towards the center of the lid for engagement by a hoist. Overpack lid 1500 serves a structural role of protecting the canister 1100 disposed inside the overpack 1130, and in some embodiments supporting the weight of the overpack when mounted thereto for transport and lifting. Accordingly, overpack lid 1500 may have a thickness greater than the thickness of the confinement lid 1200. Referring now to FIGS. 23 and 24, the overpack lid 1500 is grappled and lifted via the attached hoist 1400 by crane or other equipment, vertically aligned with overpack 1130 using the alignment pins 1408 in a manner similar to alignment pins 1406, and lowered onto the overpack. Alignment pins 1408 are then removed and mounting bolts 1512 are then installed in the threaded sockets 1510 of the overpack 1130 to complete installation and securement of the overpack lid 1500, as shown in FIG. 25. Optionally, the lifting bail 1508 may be removed. FIG. 26 shows the overpack 1130 with overpack lid 1500 fully installed and canister 1100 disposed inside loaded with waste cylinders 1121. Protective caps 1514 may be installed over mounting bolts 1512. An operator is shown in FIG. 26 to provide perspective on the size of overpack 1130 in one non-limiting embodiment, which may be about 6 or more feet in diameter and about 6 or more feet in height. Any suitable size overpack may be used. As noted herein, the shielded lifting lid 1300 is reusable. Accordingly, in some embodiments, the exemplary method described above may further comprise a step of detachably mounting a second different confinement lid 1200 to the shielded lifting lid 1300; the second confinement lid and shielded lifting lid collectively forming a second lid assembly. It will be appreciated that the two-part lid assembly 1200/1300 may also be used in applications where the confinement lid 1200 is intended to be welded to the canister 1100 for closure rather than by bolting. With reference to FIGS. 27-34, a third inventive concept will be described. FIG. 27A illustrates a high level waste (“HLW”) storage container 2010, encased in surrounding concrete 2011, as it would be in an installation. FIG. 28 illustrates the storage container 2010 in a sectional view, still with the surrounding concrete 2011. While the HLW storage container 2010 will be described in terms of being used to store a canister of spent nuclear fuel, it will be appreciated by those skilled in the art that the systems and methods described herein can be used to store any and all kinds of HLW. The HLW storage container 2010 is designed to be a vertical, ventilated dry system for storing HLW such as spent fuel. The HLW storage container 2010 is fully compatible with 100 ton and 125 ton transfer casks for HLW transfer procedures, such as spent fuel canister transfer operations. All spent fuel canister types engineered for storage in free-standing, below grade, and/or anchored overpack models can be stored in the HLW storage container 2010. As used in this section the term “canister” broadly includes any spent fuel containment apparatus, including, without limitation, multi-purpose canisters and thermally conductive casks. For example, in some areas of the world, spent fuel is transferred and stored in metal casks having a honeycomb grid-work/basket built directly into the metal cask. Such casks and similar containment apparatus qualify as canisters, as that term is used herein, and can be used in conjunction with the HLW storage container 2010 as discussed below. The HLW storage container 2010 can be modified/designed to be compatible with any size style of transfer cask. The HLW storage container 2010 can also be designed to accept spent fuel canisters for storage at an Independent Spent Fuel Storage Installations (“ISFSI”). ISFSIs employing the HLW storage container 2010 can be designed to accommodate any number of the HLW storage container 2010 and can be expanded to add additional HLW storage containers 2010 as the need arises. In ISFSIs utilizing a plurality of the HLW storage container 2010, each HLW storage container 2010 functions completely independent form any other HLW storage container 2010 at the ISFSI. The HLW storage container 2010 has a body 2020 and a lid 2030. The lid 2030 rests atop and is removable/detachable from the body 2020. Although an HLW storage container can be adapted for use as an above grade storage system, by incorporating design features found in U.S. Pat. No. 7,933,374, this HLW storage container 2010, as shown, is designed for use as a below grade storage system. Referring to FIG. 28, the body 2020 includes an outer shell 2021 and an inner shell 2022. The outer shell 2021 surrounds the inner shell 2022, forming a space 2023 therebetween. The outer shell 2021 and the inner shell 2022 are generally cylindrical in shape and concentric with one another. As a result, the space 2023 is an annular space. While the shape of the inner and outer shells 2022, 2021 is cylindrical in the illustrated embodiment, the shells can take on any shape, including without limitation rectangular, conical, hexagonal, or irregularly shaped. In some embodiments, the inner and outer shells 2022, 2021 will not be concentrically oriented. The space 2023 formed between the inner shell 2022 and the outer shell 2021 acts as a passageway for cool air. The exact width of the space 2023 for any HLW storage container 2010 is determined on a case-by-case design basis, considering such factors as the heat load of the HLW to be stored, the temperature of the cool ambient air, and the desired fluid flow dynamics. In some embodiments, the width of the space 2023 will be in the range of 1 to 6 inches. While the width of space 2023 can vary circumferentially, it may be desirable to design the HLW storage container 2010 so that the width of the space 2023 is generally constant in order to effectuate symmetric cooling of the HLW container and even fluid flow of the incoming air. As discussed in greater detail below, the space 2023 may be divided up into a plurality of passageways. The inner shell 2022 and the outer shell 2021 are secured atop a floor plate 2050. The floor plate 2050 is hermetically sealed to the outer shell 2021, and it may take on any desired shape. A plurality of spacers 2051 are secured atop the floor plate 2050 within the space 2023. The spacers 2051 support a pedestal 2052, which in turn supports a canister. When a canister holding HLW is loaded into the cavity 2024 for storage, the bottom surface of the canister rests atop the pedestal 2052, forming an inlet air plenum between the underside of the pedestal 2052 and the floor of cavity 2024. This inlet air plenum contributes to the fluid flow and proper cooling of the canister. Preferably, the outer shell 2021 is seal joined to the floor plate 2050 at all points of contact, thereby hermetically sealing the HLW storage container 2010 to the ingress of fluids through these junctures. In the case of weldable metals, this seal joining may comprise welding or the use of gaskets. Most, preferably, the outer shell 2021 is integrally welded to the floor plate 2050. An upper flange 2077 is provided around the top of the outer shell 2021 to stiffen the outer shell 2021 so that it does not buckle or substantially deform under loading conditions. The upper flange 2077 can be integrally welded to the top of the outer shell 2021. The inner shell 2022 is laterally and rotationally restrained in the horizontal plane at its bottom by support legs 2027 which straddle lower ribs 2053. The lower ribs 2053 are preferably equispaced about the bottom of the cavity 2024. The inner shell 2022 is preferably not welded or otherwise permanently secured to the bottom plate 2050 or outer shell 2021 so as to permit convenient removal for decommissioning, and if required, for maintenance. The inner shell 2022, the outer shell 2021, the floor plate 2050, and the upper flange 2077 are preferably constructed of a metal, such as a thick low carbon steel, but can be made of other materials, such as stainless steel, aluminum, aluminum-alloys, plastics, and the like. Suitable low carbon steels include, without limitation, ASTM A516, Gr. 70, A515 Gr. 70 or equal. The desired thickness of the inner and outer shells 2022, 2021 is matter of design choice and will determined on a case-by-case basis. The inner shell 2022 forms a cavity 2024. The size and shape of the cavity 2024 is also a matter of design choice. However, it is preferred that the inner shell 2022 be designed so that the cavity 2024 is sized and shaped so that it can accommodate a canister of spent nuclear fuel or other HLW. While not necessary, it is preferred that the horizontal cross-sectional size and shape of the cavity 2024 be designed in generally correspond to the horizontal cross-sectional size and shape of the canister-type that is to be used in conjunction with a particular HLW storage container. More specifically, it is desirable that the size and shape of the cavity 2024 be designed so that when a canister containing HLW is positioned in the cavity 2024 for storage (as illustrated in FIG. 30A), a small clearance exists between the outer side walls of the canister and the side walls of the cavity 2024. Designing the cavity 2024 so that a small clearance is formed between the side walls of the stored canister and the side walls of the cavity 2024 limits the degree the canister can move within the cavity during a catastrophic event, thereby minimizing damage to the canister and the cavity walls and prohibiting the canister from tipping over within the cavity. This small clearance also facilitates flow of the heated air during HLW cooling. The exact size of the clearance can be controlled/designed to achieve the desired fluid flow dynamics and heat transfer capabilities for any given situation. In some embodiments, for example, the clearance may be 1 to 3 inches. A small clearance also reduces radiation streaming. The inner shell 2022 is also equipped with multiple sets of equispaced longitudinal ribs 2054, 2055, in addition to the lower ribs 2053 discussed above. One set of ribs 2054 are preferable disposed at an elevation that is near the top of a canister of HLW placed in the cavity 2024. This set of ribs 2054 may be shorter in length in companion to the height of the cavity 2024 and a canister. Another set of ribs 2055 are set below the first set of ribs 2054. This second set of ribs 2055 is more elongated than the first set of ribs 2054, and these ribs 2055 extend to, or nearly to, the bottom of the cavity 2024. These ribs 2053, 2054, 2055 serve as guides for a canister of HLW is it is lowered down into the cavity 2024, helping to assure that the canister properly rests atop the pedestal 2052. The ribs also serve to limit the canister's lateral movement during an earthquake or other catastrophic event to a fraction of an inch. A plurality of openings 2025 are provided in the inner shell 2022 at or near its bottom between the support legs 2027. Each opening 2025 provides a passageway between the annular space 2023 and the bottom of the cavity 2024. The openings 2025 provide passageways by which fluids, such as air, can pass from the annular space 2023 into the cavity 2024. The openings 2025 are used to facilitate the inlet of cooler ambient air into the cavity 2024 for cooling a stored HLW having a heat load. As illustrated, eight openings 2025 are equispaced about the bottom of the inner shell 2022. However, any number of openings 2025 can be included, and they may have any spacing desired. The exact number and spacing will be determined on a case-by-case basis and will be dictated by such considerations as the heat load of the HLW, desired fluid flow dynamics, etc. Moreover, while the openings 2025 are illustrated as being located in the side wall of the inner shell 2022, the openings can be provided in the floor plate in certain modified embodiments of the HLW storage container. The openings 2025 in the inner shell 2022 are sufficiently tall to ensure that if water enters the cavity 2024, the bottom region of a canister resting on the pedestal 2052 would be submerged for several inches before the water level reaches the top edge of the openings 2025. This design feature helps ensure thermal performance of the system under accidental flooding of the cavity 2024. With reference to FIG. 29, a layer of insulation 2026 is provided around the outside surface of the inner shell 2022 within the annular space 2023. The insulation 2026 is provided to minimize heating of the incoming cooling air in the space 2023 before it enters the cavity 2024. The insulation 2026 helps ensure that the heated air rising around a canister situated in the cavity 2024 causes minimal pre-heating of the downdraft cool air in the annular space 2023. The insulation 2026 is preferably chosen so that it is water and radiation resistant and undegradable by accidental wetting. Suitable forms of insulation include, without limitation, blankets of alumina-silica fire clay (Kaowool Blanket), oxides of alumina and silica (Kaowool S Blanket), alumina-silica-zirconia fiber (Cerablanket), and alumina-silica-chromia (Cerachrome Blanket). The desired thickness of the layer of insulation 2026 is matter of design and will be dictated by such considerations such as the heat load of the HLW, the thickness of the shells, and the type of insulation used. In some embodiments, the insulation will have a thickness in the range ½ to 6 inches. As shown in FIGS. 28 and 29, inlet ducts 2060 are disposed on the top surface of the upper flange 2077. Each inlet duct 2060 connects to two inlet passageways 2061 which continue from under the upper flange 2077, into the space 2023 between the outer and inner shells 2021, 2022, and then connect to the cavity 2024 by lower openings 2062 in the bottom of the inner shell 2022. Within the space 2023, the inlet passageways 2061 are separated by dividers 2063 to keep cooling air flowing through each inlet passageway 2061 separate from the other inlet passageways 2061 until the cooling air emerges into the cavity 2024. FIGS. 30A and 30B illustrate the configuration of the inlet passageways 2061 and the dividers 2063. Each inlet passageway 2061 connects with the space 2023 by openings 2064 in the top of the outer shell 2021. From the openings 2064, the cooling air continues down the in the space, via the individual inlet passageways 2061 created by the dividers 2064, and into the cavity 2024, where it is used to cool a placed HLW canister. The dividers 2063 are equispaced within the space 2023 to aid in balancing the air pressure entering the space 2023 from each inlet duct and inlet passageway. Also, as shown in the figures, each of the lower ribs 2053 is integrated with one of the dividers 2063, such that the lower ribs form an extension of the dividers, extending into the cavity 2024. Referring back to FIG. 29, each inlet duct 2060 includes a duct cover 2065, to help prevent rain water or other debris from entering and/or blocking the inlet passageways 2061, affixed on top of an inlet wall 2066 that surrounds the inlet passageways 2061 on the top surface of the upper flange 2077. The inlet wall 2066 is peripherally perforated around the entire periphery of the opening of the inlet passageways 2061. At least a portion of the lower part of the inlet ducts are left without perforations, to aid in preventing rain water from entering the HLW storage container. Preferably, the inlet wall 2066 is perforated over 60% or more of its surface, and the perforations can be made in any shape, size, and distribution in accordance with design preferences. When the inlet ducts 2060 are formed with the inlet wall 2006 peripherally perforated, each of the inlet ducts has been found to maintain an intake air pressure independently of each of the other inlet ducts, even in high wind conditions, and each of the inlet ducts has been found to maintain an intake air pressure substantially the same as each of the other inlet ducts, again, even in high wind conditions. The lid 2030 rests atop and is supported by the upper flange 2077 and a shell flange 2078, the latter being disposed on and connected to the tops edge of the inner shell 2022. The lid 2030 encloses the top of the cavity 2024 and provides the necessary radiation shielding so that radiation does not escape from the top of the cavity 2024 when a canister loaded with HLW is stored therein. The lid 2030 is designed to facilitate the release of heated air from the cavity 2024. FIG. 31A illustrates the HLW storage container 2010 with a canister 2013 placed within the cavity 2024. As shown in the FIG. 31B detailed view, the bottom of the canister 2013 sits on the pedestal 2052, and the lower ribs 2053 maintain a space between the bottom of the canister 2013 and the inner shell 2022. Similarly, the FIG. 31C detailed view shows that the upper ribs 2054 maintain a space between the top of the canister 2013 and the inner shell 2022. The FIG. 31D detailed view shows the lid 2030 resting atop the upper flange 2077 and the shell flange 2078. The lid 2030 includes a closure gasket 2031 which forms a seal against the upper flange 2077 when the lid 2030 is seated, and a leaf spring gasket 2032 which forms a seal against the shell flange 2078. FIGS. 32 and 33 illustrate the lid 2030 removed from the body of the HLW storage container. Referring first to FIG. 32, the lid 2030 is preferably constructed of a combination of low carbon steel and concrete (or another radiation absorbing material) in order to provide the requisite radiation shielding. The lid 2030 includes an upper lid part 2033 and a lower lid part 2034. The upper lid part 2033 preferable extends at least as high as, if not higher than, the top of each inlet duct 2060. Each lid part 2033, 2034 includes an external shell 2035, 2036 encasing an upper concrete shield 2037 and a lower concrete shield 2038. One or more outlet passageways 2039 are formed within and around the body parts 2033, 2034 to connect the cavity with the outlet duct 2040 formed on the top surface of the lid 2030. The outlet passageways 2039 pass over the lower lid part 2034, between the upper and lower lid parts 2033, 2034, and up through a central aperture within the upper lid part 2034. The outlet duct 2040 covers this central aperture to better control the heated air as it rises. By being disposed on the top of the lid 2030, the outlet duct 2040 may also be raised up significantly higher than the inlet ducts, using any desired length of extension for the outlet duct. By raising up the outlet duct higher, mixing between the heated air emitted from the outlet duct and cooler air being drawn into the inlet ducts can be significantly reduced, if not eliminated altogether. The outlet duct 2040, which is constructed similar to the inlet ducts, includes a duct cover 2041, to help prevent rain water or other debris from entering and/or blocking the outlet passageways 2039, affixed on top of an outlet wall 2042 that surrounds the outlet passageways 2039 on the top surface of the upper lid part 2033. The outlet wall 2042 is peripherally perforated around the entire periphery of the opening of the outlet passageways 2039. At least a portion of the lower part of the outlet duct is left without perforations, to aid in preventing rain water from entering the HLW storage container. Preferably, the outlet wall 2042 is perforated over 60% or more of its surface, and the perforations can be made in any shape, size, and distribution in accordance with design preferences. The external shell of the lid 2030 may be constructed of a wide variety of materials, including without limitation metals, stainless steel, aluminum, aluminum-alloys, plastics, and the like. The lid may also be constructed of a single piece of material, such as concrete or steel for example, so that it has no separate external shell. When the lid 2030 is positioned atop the body 2020, the outlet passageways 2039 are in spatial cooperation with the cavity 2024. As a result, cool ambient air can enter the HLW storage container 2010 through the inlet ducts 2060, flow into the space 2023, and into the bottom of the cavity 2024 via the openings 2062. When a canister containing HLW having a heat load is supported within the cavity 2024, this cool air is warmed by the HLW canister, rises within the cavity 2024, and exits the cavity 2024 via the outlet ducts 2040. Because the inlet ducts 2060 are placed on different sides of the lid 2030, and the dividers separate the inlet passageways associated with the different inlet ducts, the hydraulic resistance to the incoming air flow, a common limitation in ventilated modules, is minimized. This configuration makes the HLW storage container less apt to build up heat internally under high wind conditions. A plurality of HLW storage containers 2100 can be used at the same ISFSI site and situated in arrays as shown in FIG. 34. Although the HLW storage containers 2100 are closely spaced, the design permits a canister in each HLW storage container 2100 to be independently accessed and retrieved easily. In addition, the design of the individual storage containers 2100, and particularly the design and positioning of the inlet and outlet ducts, enables the inlet ducts of a first of the storage containers to maintain air pressure independently of the inlet ducts of a second of the storage containers. Each storage container therefore will operate independently of each of the other storage containers, such that the failure of one storage container is unlikely to lead directly to the failure of other surrounding storage containers in the array. With reference to FIGS. 35-47, a fourth inventive concept will be described. Referring to FIG. 35, a dual-walled DSC 3099 according to one embodiment of the present invention is disclosed. The dual-walled DSC 3099 and its components are illustrated and described as an MPC style structure. However, it is to be understood that the concepts and ideas disclosed herein can be applied to other areas of high level radioactive waste storage, transportation and support. Moreover, while the dual-walled DSC 3099 is described as being used in combination with a specially designed fuel basket 3090 (which in of itself constitutes an invention), the dual-walled DSC 3090 can be used with any style of fuel basket, such as the one described in U.S. Pat. No. 5,898,747, issued Apr. 27, 1999. In fact. In some instances it may be possible to use the dual-walled DSC 3099 without a fuel basket, depending on the intended function. Furthermore, the dual-walled DSC 3099 can be used to store and/or transport any type of high level radioactive materials and is not limited to SNF. As will become apparent from the structural description below, the dual-walled DSC 3099 contains two independent containment boundaries about the storage cavity 3030 that operate to contain both fluidic (gas and liquid) and particulate radiological matter within the cavity 3030. As a result, if one containment boundary were to fail, the other containment boundary will remain intact. While theoretically the same, the containment boundaries formed by the dual-walled DSC 3099 about the cavity 3030 can be literalized in many ways, including without limitation a gas-tight containment boundary, a pressure vessel, a hermetic containment boundary, a radiological containment boundary, and a containment boundary for fluidic and particulate matter. These terms are used synonymously throughout this application. In one instance, these terms generally refer to a type of boundary that surrounds a space and prohibits all fluidic and particulate matter from escaping from and/or entering into the space when subjected to the required operating conditions, such as pressures, temperatures, etc. Finally, while the dual-walled DSC 3009 is illustrated and described in a vertical orientation, it is to be understood that the dual-walled DSC 3099 can be used to store and/or transport its load in any desired orientation, including at an angle or horizontally. Thus, use of all relative terms through this specification, including without limitation “top,” “bottom,” “inner” and “outer,” are used for convenience only and are not intended to be limiting of the invention in such a manner. The dual-walled DSC 3099 includes a first shell that acts as an inner shell 3010 and a second shell that acts as an outer shell 3020. The inner and outer shells 3010, 3020 are preferably cylindrical tubes and are constructed of a metal. Of course, other shapes can be used if desired. The inner shell 3010 is a tubular hollow shell that includes an inner surface 3011, an outer surface 3012, a top edge 3013 and a bottom edge 3014. The inner surface 3011 of the inner shell 3010 forms a cavity/space 3030 for receiving and storing SNF. The cavity 3030 is a cylindrical cavity formed about a central axis. The outer shell 3020 is also a tubular hollow shell that includes an inner surface 3021, an outer surface 3022, a top edge 3023 and a bottom edge 3024. The outer shell 3020 circumferentially surrounds the inner shell 3010. The inner shell 3010 and the outer shell 3020 are constructed so that the inner surface 3021 of the outer shell 3020 is in substantially continuous surface contact with the outer surface 3012 of the inner shell 3010. In other words, the interface between the inner shell 3010 and the outer shell 3020 is substantially free of gaps/voids and are in conformal contact. This can be achieved through an explosive joining, a cladding process, a roller bonding process and/or a mechanical compression process that bonds the inner shell 3010 to the outer shell 3020. The continuous surface contact at the interface between the inner shell 3010 and the outer shell 3020 reduces the resistance to the transmission of heat through the inner and outer shells 3010, 3020 to a negligible value. Thus, heat emanating from the SNF loaded within the cavity 3030 can efficiently and effectively be conducted outward through the shells 3010, 3020 where it is removed from the outer surface 3022 of the outer shell via convection. Even though the interface is formed in any of these manners, there still remains an interstitial space 3097 between the inner shell 3010 and the outer shell 3020. Alternatively, the interstitial space may be formed without the inner surface of the outer shell being in substantially continuous surface contact with the outer surface of the inner shell. As is discussed in more detail below, the presence of this interstitial space is used advantageously during a leak testing process. The inner and outer shells 3010, 3020 are preferably both made of a metal. As used herein, the term metal refers to both pure metals and metal alloys. Suitable metals include without limitation austenitic stainless steel and other alloys including Hastelloy™ and Inconel™. Of course, other materials can be utilized. The thickness of each of the inner and outer shells 3010, 3020 is preferably in the range of 5 mm to 25 mm. The outer diameter of the outer shell 3020 is preferably in the range of 1700 mm to 2000 mm. The inner diameter of the inner shell 3010 is preferably in the range of 1700 mm to 1900 mm. The specific size and/or thickness of the shells 3010, 3020, however, is a matter of design choice. In some embodiments, it may be further preferable that the inner shell 3010 be constructed of a metal that has a coefficient of thermal expansion that is equal to or greater than the coefficient of thermal expansion of she metal of which the outer shell 3020 is constructed. Thus, when the SNF that is stored in the cavity 3030 and emits heat, the outer shell 3020 will not expand away from the inner shell 3010. This ensures that the continuous surface contact between the outer surface 3012 of the inner shell 3010 and the outer surface 3021 of the outer shell 3020 will be maintained and a gaps will not form under heat loading conditions. The dual-walled DSC 3099 also includes a first lid that acts as an inner top lid 3060 for the inner shell 3010 and a second lid that acts as an outer top lid 3070 for the second shell 3020. The inner and outer top lids 3060, 3070 are plate-like structures that are preferably constructed of the same materials discussed above with respect to the shells 3010, 3020. Preferably the thickness of the inner top lid 3060 is in the range of 99 mm to 300 mm. The thickness of the outer top lid is preferably in the range of 50 mm to 150 mm. The invention is not, however, limited to any specific dimensions, which will be dictated on a case-by-case basis and the radioactive levels of the SNF to be stored in the cavity 3030. Referring to FIG. 36, the inner top lid 3060 includes a top surface 3061, a bottom surface 3062 and an outer lateral surface/edge 3063. The outer top lid 3070 includes a top surface 3071, a bottom surface 3072 and an outer lateral surface/edge 3073. When fully assembled, the outer lid 3070 is positioned atop the inner lid 3060 so that the bottom surface 3072 of the outer lid 3070 is in substantially continuous surface contact with the top surface 3061 of the inner lid 3060. The outer lid 3070 also includes a test port 3095, to which one end of conduit is coupled (see FIGS. 44 and 45) in fluidic communication therewith. As is discussed below, the other end of the conduit is fitted with both a removable seal, to enable leak testing, and valve, both being included to comply with ASME Code. During an SNF underwater loading procedure, the inner and outer lids 3060, 3070 are removed. Once the cavity 3030 is loaded with the SNF, the inner top lid 3060 is positioned so as to enclose the top end of the cavity 3030 and rests atop the brackets 3015. Once the inner top lid 3060 is in place and seal welded to the inner shell 3010, the cavity 3030 is evacuated/dried via the appropriate method and backfilled with nitrogen, helium or another inert gas. The drying and backfilling process of the cavity 3030 is achieved via the holes 3064 of the inner lid 3060 that form passageways into the cavity 3030. Once the drying and backfilling is complete, the holes 3061 are filled with a metal or otherwise plugged so as to hermetically seal the cavity 3030. Referring now to FIGS. 35 and 37 concurrently, the outer shell 3020 has an axial length L2 that is greater than the axial length L1 of the inner shell 3010. As such, the top edge 3013 of the inner shell 3010 extends beyond the top edge 3023 of the outer shell 3020. Similarly, the bottom edge 3024 of the outer shell 3020 extends beyond the bottom edge 3013 of the inner shell 3010. The offset between the top edges 3013, 3023 of the shells 3010, 3020 allows the top edge 3013 of the inner shell 3010 to act as a ledge for receiving and supporting the outer top lid 3070. When the inner lid 3060 is in place, the inner surface 3011 of the inner shell 3010 extends over the outer lateral edges 3063. When the outer lid 3070 is then positioned atop the inner lid 3060, the inner surface 3021 of the outer shell 3020 extends over the outer lateral edge 3073 of the outer top lid 3070. The top edge 3023 of the outer shell 3020 is substantially flush with the top surface 3071 of the outer top lid 3070. The inner and outer top lids 3060, 3070 are welded to the inner and outer shells 3010, 3020 respectively after the fuel is loaded into the cavity 3030. Conventional edge groove welds can be used. However, it is preferred that all connections between the components of the dual-walled DSC 3099 be through-thickness weld. The dual-walled DSC 3099 also includes a first plate that acts as an inner base plate 3040 and a second plate that acts as an outer base plate 3050. The inner and outer base plates 3040, 3050 are rigid plate-like structures having circular horizontal cross-sections. The invention is not so limited, however, and the shape and size of the base plates 3040, 3050 is dependent upon the shape of the inner and outer shells 3010, 3020. The inner base plate 3040 includes a top surface 3041, a bottom surface 3042 and an outer lateral surface/edge 3043. Similarly, the outer base plate 3050 includes a top surface 3051, a bottom surface 3052 and an outer lateral surface/edge 3053. The top surface 3041 of the inner base plate 3040 forms the floor of the cavity 3030. The inner base plate 3040 rests atop the outer base plate 3050. Similar to the other corresponding components of the dual-walled DSC 3099, the bottom surface 3042 of the inner base plate 3040 is in substantially continuous surface contact with the top surface 3051 of the outer base plate 3050. As a result, the interface between the inner base plate 3040 and the outer base plate 3050 is free of gaseous gaps/voids for thermal conduction optimization. An explosive joining, a cladding process, a roller bonding process and/or a mechanical compression process can be used to effectuate the contact between the base plates 3040, 3050. Preferably, the thickness of the inner base plate 3040 is in the range of 50 mm to 150 mm. The thickness of the outer base plate 3050 is preferably in the range of 99 mm to 200 mm. Preferably, the length from the top surface of the outer top lid 3070 to the bottom surface of the outer base plate 3050 is in the range of 4000 mm to 5000 mm, but the invention is in no way limited to any specific dimensions. The outer base plate 3050 may be equipped on its bottom surface with a grapple ring (not shown) for handling purposes. The thickness of the grapple ring is preferably between 50 mm and 150 mm. The outer diameter of the grapple ring is preferably between 350 mm and 450 mm. Referring now to FIGS. 36 and 38 concurrently, the inner shell 3010 rests atop the inner base plate 3040 in a substantially upright orientation. The bottom edge 3014 of the inner shell 3010 is connected to the top surface 3041 of the inner base plate 3040 by a through-thickness single groove (V or J shape) weld. The outer surface 3012 of the inner shell 3010 is substantially flush with the outer lateral edge 3043 of the inner base plate 3040. The outer shell 3020, which circumferentially surrounds the inner shell 3010, extends over the outer lateral edges 3043, 3053 of the inner and outer base plates 3040, 3050 so that the bottom edge 3024 of the outer shell 3020 is substantially flush with the bottom surface 3052 of the outer base plate 3050. The inner surface 3021 of the outer shell 3020 is also connected to the outer base plate 3050 using a through-thickness edge weld. In an alternative embodiment, the bottom edge 3024 of the outer shell 3020 could rest atop the top surface 3051 of the ratter base plate 3050 (rather than extending over the outer later edge of the base plate 3050). In that embodiment, the bottom edge 3024 of the outer shell 3020 could be welded to the fop surface 3051 of the outer base plate 3050. When all of the seal welds discussed above are completed, the combination of the inner shell 3010, the inner base plate 3040 and the inner top lid 3060 forms a first hermetically sealed structure surrounding the cavity 3030, thereby creating a first pressure vessel. Similarly, the combination of the outer shell 3020, the outer base plate 3050, and the outer top lid 3070 form a second sealed structure about the first hermetically sealed structure, thereby creating a second pressure vessel about the first pressure vessel and the cavity 3030. With the inclusion of the test port 3095, the seal of the second pressure vessel also effectively includes the conduit, sealed at the end not coupled to the test port. Theoretically, the first pressure vessel is located within the internal cavity of the second pressure vessel. Each pressure vessel is engineered to autonomously meet the stress limits of the ASME Code with significant margins. Unlike the prior art DSC, all of the SNF stored in the cavity 3030 of the dual-walled DSC 3090 share a common confinement space. The common confinement space (i.e., cavity 3030) is protected by two independent gas-tight pressure retention boundaries. Each of these boundaries can withstand both sub-atmospheric supra-atmospheric pressures as needed, even when subjected to the thermal load given off by the SNF within the cavity 3030. In the event of a failure of the first hermetically sealed structure surrounding the cavity 3030, at least some of the backfilled helium will leak into the interstitial space 3097. Because helium is both an inert gas and a small molecule, the testing equipment and processes, described in greater below, are able to draw helium through the interstitial space 3097 for detection and determination of whether the first hermetically sealed structure has failed. A ventilated system 3101 is shown in FIGS. 39A & 39B. The cask lid 3107 of a ventilated cask 3103 is shown in FIG. 39A, and a cross section of the ventilated cask 3103 is shown in FIG. 39B. As can be seen in FIG. 39B, the ventilated cask 3103 includes a cylindrical cask body 3105 and a cask lid 3107. The cylindrical cask body 3105 includes a set of air inlet ducts 3109 near its bottom and a set of air outlet ducts 3111 near its top. A dual-walled DSC 3099 containing decaying spent nuclear fuel stands upright inside the ventilated cask 3103, with a small diametrical clearance, in the form an annular gap 3113, being formed between an inner surface of the cylindrical cask body 3105 of the ventilated cask 3103 and the outer surface 3115 of the DSC 399. The outer surface 3115 of the DSC 3099 becomes heated due to the thermal energy being generated by the spent nuclear fuel sealed in the DSC 3099. The heat of the outer surface 3115 causes the surrounding air column to heat and rise, resulting in a continuous natural convective ventilation action. The cold air entering the air inlet ducts 3111 at the bottom of the cylindrical cask body 3105 is progressively heated as it rises in the annular gap 3113, reaching its maximum value as it exits the cylindrical cask body 3105. Different designs of such casks are known and described in greater detail in the prior art, e.g., U.S. patent publication No. 2003/0147486, published Aug. 7, 2003, and WO 2013/115881, published Aug. 8, 2013, the disclosures of which are incorporated herein by reference in their entirety. An assembled cask 3151 is shown in FIG. 40. The cask lid 3153 includes ventilation ducts 3155, through one of which the conduit 3157 runs to the outside of the cask 3151. The conduit 3157 extends down the side of the cask body 3159, and into an enclosure 3161 which is affixed to the exterior of the cask body 3159. Although not shown, the conduit may be secured to the cask body 3159 by appropriate brackets affixed to the cask body 3159. As an alternative, the conduit may extend away from the cask body entirely, to an enclosure that is affixed to an independent support (such as a nearby pole or other wall). The conduit 3157 is preferably ¼ inch stainless steel conduit, as such conduit can be evacuated without collapsing. Other conduit materials and sizes that exhibit a similar strength and properties as stainless steel conduit may also be used. Also, the conduit 3157 follows a tortuous path from the first end, where it is coupled to the test port, to the second end, to which the seal, valve, and alternately the testing equipment are coupled. The tortuous path is included so that there is no line of sight path for radiation to escape from the DSC to the outside of the cask 3151. Also, by running the conduit to the outside of the cask, the testing described below may be performed while the cask remains in its storage position and the cask lid remains on the cask, thereby minimizing the amount of time needed to perform the test and significantly reducing the amount of radiation to which workers are exposed. FIG. 41 shows a detailed view of the enclosure 3161 with a cover 3163 in place, which serves to protect contents of the internal chamber of the enclosure 3161, and may be used to make the enclosure waterproof, if desired. One sidewall 3165 of the enclosure 3161 and cover 3163 may include features for locking the cover in place—as shown these features are a pair of aligned rings 3167 on the sidewall 3165 and on the cover 3163, which enable a lock or other security feature (e.g., a tag) to be placed on the enclosure 3161. The conduit 3157 passes through sidewall 3169 and into the internal chamber 3171 of the enclosure 3161, as shown in FIG. 42. Within the enclosure 3161, the second end 3173 of the conduit 3157 includes one test apparatus connector 3175 and a secondary connector 3177. The two connectors 3175, 3177 provide a dual failsafe boundary in compliance with ASME Code. When no test is being performed, a retrievable seal 3179 is coupled to the test apparatus connector 3175. The removable seal 3179 may be of any type suitable for sealing the test apparatus connector 3175 and for use under the operating conditions described herein. The test apparatus connector 3175 is otherwise configured for coupling to the test apparatus to be used, which may be a mass spectrometer leak detector (MSLD) of the kind which are readily available on the market today, and one of ordinary skill in the art would be aware of the types of different MSLDs available. The secondary connector 3177 is regulated by a valve 3181 which is suitable for the operating conditions described herein. During the testing process, once tests are performed by the MSLD, a source of a second inert gas (different from the inert gas which is filled in the canister) may be connected to the secondary connector so that the conduit and at least part of the interstitial space are backfilled with this second inert gas. An alternative for extending the conduit 3157 to the outside of the cask 3151 is shown in FIG. 43. In this embodiment, a groove 3191 is formed in the cask lid 3153, and the conduit 3157 is positioned in the groove 3191, with the cask lid 3153 in place on the cask body 3159 so that the conduit 3157 may extend to the outside of the cask 3151. FIG. 44 shows this same embodiment without the cask lid in place. As shown, the conduit 3157 extends across the top of the cask body 3159 from the test port 3193 formed in the outer top lid 3195 of the second pressure vessel. The conduit 3157 is coupled to the test port 3193 with an appropriate pressure fitting 3199, which may also be constructed from stainless steel. FIGS. 45 and 46 illustrate the test port 3193 in greater detail—in FIG. 46, the cask is not shown for additional clarity. A portion of the interstitial space 3201 exists between the inner top lid 3203 and the outer top lid 3195. As indicated above, although the interstitial space 3201 may be very small, in such a small space, small, inert helium atoms may still move around within such a space. In the event that larger inert atoms are used to fill the cavity of the canister, the choices of how to form the interstitial space may be more limited to take into consideration the presently disclosed system and method of leak detection. The test port 3193 extends through the outer top lid 3195 so that it is in fluidic communication with the interstitial space 3201. Thus, when the vacuum is created in the conduit, if helium molecules are present within the interstitial space, at least some of them will be drawn into the conduit, and from there into the attached MSLD, so that they may be detected. A block diagram showing the leak detection system and illustrating the method for detecting leaks is depicted in FIG. 47. The interstitial space 3251 is formed between the inner pressure vessel 3253 and the outer pressure vessel 3255. The first end 3257 of the conduit 3259 is coupled to the test port 3261, and the second end 3263 of the conduit 3259 is coupled to the leak detector 3265, so that the interstitial space 3251, the test port 3261, the conduit 3259, and the leak detector 3265 are all in fluidic communication. The leak detector 3265 includes a vacuum system 3267, which is used to draw gas from the conduit 3259, and thus also from the interstitial space 3251, into the leak detector 3265 for analysis. The leak detector also includes a gas sensor 3269, which is preferably a mass spectrometer, and a pressure sensor 3271 to monitor the state of the vacuum established in the conduit 3259. The gas sensor 3269 is configured to detect the presence of the inert gas backfilled into the cavity 3273 of the inner pressure vessel 3253. During operation of the leak detector 3265, in one embodiment, the mass spectrometer of an MSLD is used to analyze the gas being drawn from the interstitial space while the vacuum is being established. An analysis is performed to determine if the gas being drawn contains helium atoms, and the number of helium atoms are counted. Depending upon the conditions existing at the time of testing, once the count of helium atoms passes a predetermined number, then a leak in the fluidic containment boundary that is formed by the inner pressure vessel may be said to exist. This predetermined number may vary, depending upon the particular storage container, conditions at the time the storage container was manufactured, or the conditions existing at the storage site. In other words, the presence of a single helium atom is not necessarily indicative of a leak in the inner storage container. However, a count of several helium atoms may be indicative of a leak. Further, because of the ease of the testing procedures, a particular canister might be tested two or more times to confirm the presence of excess helium in the interstitial space before a leak is determined to be positively identified. Also during operation of the leak detector 3265, in one embodiment, the pressure sensor of the MSLD is used to monitor the established vacuum in the conduit and in the interstitial space. In the event that the vacuum decreases over a short period of time from its initially established level, or alternatively if the MSLD needs to perform additional work to maintain the vacuum once established, then a leak in the fluidic containment boundary that is formed by the outer pressure vessel may be said to exist. In one embodiment, an MSLD is able to establish a vacuum in the conduit and in the interstitial space at about 10−8 atms, and if that established vacuum changes by about an order of magnitude, to about 10−7 atms within a time period of about 1 second, then this is an indicator that there is a breach in the containment provided by the outer pressure vessel. Once a test is complete, and whether or not a potential or actual leak is identified, the MSLD is decoupled from the conduit, and the removable seal may be put back in place on the test apparatus connector. Alternatively, before the removable seal is put back in place, the conduit may be backfilled with an inert gas that is different from the inert gas used to backfill the cavity of the inner pressure vessel. The two tests performed by the leak tester are very accurate, and unlike current testing systems, they do not require further investigation to determine if the test resulted in a false positive identification of a leak. The simplicity of the leak testing system and processes described above enables testing of radioactive materials containment on a regular basis, such as monthly, semi-annually, annually, or at any other chosen interval, without requiring dedicated (and costly) test equipment being connected to every individual containment system. Although dedicated equipment permits constant monitoring, it has been found that intermittent testing is sufficient and more cost effective. In addition, testing a single radioactive materials canister may be performed quickly, meaning that a reduction in manpower may be realized by implementing such systems and methods. Finally, the additional equipment that is added to a canister for performing these leak tests is not complex and requires little maintenance, thereby enabling further cost savings to be realized. With reference to FIGS. 48-52B, a fifth inventive concept will be described. The lid 4011 and top portion of a side wall 4013 for an MPC of the prior art are shown in FIG. 48. The top surface 4015 of the lid 4011 includes a beveled edge 4017, and the closure weld 4019 joining the lid 4011 to the side wall 4013 is formed in the space between the half V-shaped space between the beveled edge 4017 and the top portion of the side wall 4013. As shown, the weld is a through-thickness single groove weld V-shaped groove, although the groove could instead be J-shaped. Due the physical configuration of the lid, the sidewall, and the closure weld, this type of closure weld is not susceptible to 100% volumetric examination. A dual-walled MPC 4201 is illustrated in FIG. 49A, and this MPC 4201 is configured so that the closure weld may be subjected to 100% volumetric examination. The dual-walled MPC 4201 may be used with any style of fuel basket, such as the one described in U.S. Pat. No. 5,898,747, issued Apr. 27, 1999. In some instances it may be possible to use the dual-walled MPC 4201 without a fuel basket, depending on the intended function. Furthermore, the dual-walled MPC 4201 may be used to store and/or transport any type of high level radioactive materials and is not limited to spent nuclear fuel. As will become apparent from the structural description below, the dual-walled MPC 4201 creates two independent containment boundaries about the storage cavity 4203 which operate to contain both fluidic (gas and liquid) and particulate radiological matter within the cavity 4203. As a result, if one containment boundary were to fail, the other containment boundary will remain intact. While theoretically the same, the containment boundaries formed by the dual-walled MPC 201 about the cavity 4203 can be literalized in many ways, including without limitation a gas-tight containment boundary, a pressure vessel, a hermetic containment boundary, a radiological containment boundary, and a containment boundary for fluidic and particulate matter. These terms are used synonymously throughout this application. In one instance, these terms generally refer to a type of boundary that surrounds a space and prohibits all fluidic and particulate matter from escaping from and/or entering into the space when subjected to the required operating conditions, such as pressures, temperatures, etc. Finally, while the dual-walled MPC 4201 is illustrated and described in a vertical orientation, it is to be understood that the dual-walled MPC 4201 can be used to store and/or transport its load in any desired orientation, including at an angle or horizontally. Thus, use of all relative terms through this specification, including without limitation “top,” “bottom,” “inner” and “outer,” are used for convenience only and are not intended to be limiting of the invention in such a manner. The dual-walled MPC 4201 includes a first shell that acts as an inner shell 4205 and a second shell that acts as an outer shell 4207. The inner and outer shells 4205, 4207 are preferably cylindrical tubes and are constructed of a metal. Of course, other shapes can be used if desired. The inner shell 4205 is a tubular hollow shell that includes an inner surface 4209, an outer surface 4210, a top edge 4212 and a bottom edge 4215. The inner surface 4209 of the inner shell 4205 forms a cavity/space 4203 for receiving and storing SNF. The cavity 4203 is a cylindrical cavity formed about a central axis. The outer shell 4207 is also a tubular hollow shell that includes an inner surface 4221, an outer surface 4223, a top edge 4225 and a bottom edge 4227. The outer shell 4207 circumferentially surrounds the inner shell 4205. The inner shell 4205 and the outer shell 4207 are constructed so that the inner surface 4221 of the outer shell 4207 is in substantially continuous surface contact with the outer surface 4223 of the inner shell 4205. In other words, the interface between the inner shell 4205 and the outer shell 4207 is substantially free of gaps/voids such that the two shells 4205, 4207 are in conformal contact. This can be achieved through an explosive joining, a cladding process, a roller bonding process and/or a mechanical compression process that bonds the inner shell 4205 to the outer shell 4207. The continuous surface contact at the interface between the inner shell 4205 and the outer shell 4207 reduces the resistance to the transmission of heat through the inner and outer shells 4205, 4207 to a negligible value. Thus heat emanating from the spent nuclear fuel loaded within the cavity 4203 can efficiently and effectively be conducted outward through the shells 4205, 4207 where it is removed from the outer surface 4223 of the outer shell via convection. The inner and outer shells 4205, 4207 are preferably both made of a metal. As used herein, the term metal refers to both pure metals and metal alloys. Suitable metals include without limitation austenitic stainless steel and other alloys including Hastelloy™ and Inconel™. Of course, other materials can be utilized. The thickness of each of the inner and outer shells 4205, 4207 is preferably in the range of 5 mm to 35 mm. The outer diameter of the outer shell 4207 is preferably in the range of 1700 mm to 2000 mm. The inner diameter of the inner shell 4205 is preferably in the range of 1700 mm to 1600 mm. The specific size and/or thickness of the shells 4205, 4207, however, is a matter of design choice. In some embodiments, it may be further preferable that the inner shell 4205 be constructed of a metal that has a coefficient of thermal expansion that is equal to or greater than the coefficient of thermal expansion of the metal of which the outer shell 4207 is constructed. Thus, when the spent nuclear fuel that is stored in the cavity 4203 emits heat, the outer shell 4207 will not expand away from the inner shell 4205. This ensures that the continuous surface contact between the outer surface 4210 of the inner shell 4205 and the outer surface 4223 of the outer shell 4207 will be maintained and a gaps will not form under heat loading conditions. The dual-walled MPC 4201 also includes a first top plate that acts as an inner top lid 4229 for the inner shell 4205 and a second top plate that acts as an outer top lid 4231 for the outer shell 4207. The inner and outer top lids 4229, 4231 are plate-like structures that are preferably constructed of the same materials discussed above with respect to the shells 4205, 4207. Preferably the thickness of the inner top lid 4229 is in the range of 99 mm to 300 mm. The thickness of the outer top lid 4231 is preferably in the range of 50 mm to 150 mm. The invention is not, however, limited to any specific dimensions, which will be dictated on a case-by-case basis and the radioactive levels of the spent nuclear fuel to be stored in the cavity 4203. The inner top lid 4229 includes a top surface 4233 with a beveled edge 4235, a bottom surface 4237, an outer lateral surface/edge 4239, and a channel 4241 formed in the top surface 4233 and set in from the beveled edge 4235. The outer top lid 4231 includes a top surface 4243 with a beveled edge 4245, a bottom surface 4247, an outer lateral surface/edge 4249, and a channel 4251 formed in the top surface 4243 and set in from the beveled edge 4245. When fully assembled, the outer lid 4231 is positioned atop the inner lid 4229 so that the bottom surface 4247 of the outer lid 4231 is in substantially continuous surface contact with the top surface 4233 of the inner lid 4229. Both the inner top lid 4229 and the outer top lid 4231 also include vent and/or drain ports 4253, 4255. During loading procedure involving spent nuclear fuel, the cavity 4203 is loaded with the spent nuclear fuel, then the inner top lid 4229 is positioned so as to enclose the top end of the cavity 4203 and rests atop brackets (not shown). Once the inner top lid 4229 is in place, a closure weld is formed to seal the inner top lid 4229 to the inner shell 4205. The top lid 4229 may be welded to the inner shell 4205 using any suitable welding technique or combinations of techniques that use a filler material. Examples of suitable welding techniques include resistance seam welding, manual metal arc welding, metal inert gas welding, tungsten inert gas welding, submerged arc welding, plasma arc welding, gas welding, electroslag welding, thermit welding. After the cavity 4203 is sealed by the closure weld, it may then be evacuated/dried via the appropriate method and backfilled with nitrogen, helium or another inert gas using the ports 4249 of the inner lid 4229 that form passageways into the cavity 4203. The ports 4249 may thereafter be filled with a metal or other wise plugged so as to hermetically seal the cavity 4203. The outer shell 4207 has an axial length that is greater than the axial length of the inner shell 4205. As such, the top edge 4225 of the outer shell 4207 extends beyond the top edge 4211 of the inner shell 4205. Similarly, the bottom edge 4227 of the outer shell 4207 extends beyond the bottom edge 4215 of the inner shell 4205. The offset between the top edges 4211, 4225 of the shells 4205, 4207 allows the top edge 4211 of the inner shell 4205 to act as a ledge for receiving and supporting the outer top lid 4231. When the inner top lid 4229 is in place, the inner surface 4209 of the inner shell 4205 extends over the outer lateral edges 4239. When the outer top lid 4231 is then positioned atop the inner lid 4229, the inner surface 4221 of the outer shell 4207 extends over the outer lateral edge 4249 of the outer top lid 4231. The top edge 4225 of the outer shell 4207 is substantially flush with the top surface 4253 of the outer top lid 4231. The inner and outer top lids 4229, 4231 are welded to the inner and outer shells 4205, 4207 respectively after the fuel is loaded into the cavity 4203. Similar to the inner top lid 4229, once the outer top lid 4231 is in place, a closure weld is formed to seal the outer top lid 4231 to the outer shell 4207. The outer top lid 4231 may be welded to the outer shell 4207 using any suitable welding technique or combinations of techniques that use a filler material. Examples of suitable welding techniques include resistance seam welding, manual metal arc welding, metal inert gas welding, tungsten inert gas welding, submerged arc welding, plasma arc welding, gas welding, electroslag welding, thermit welding. The closure welds sealing the inner and outer top lids 4229, 4231 to the inner and outer shells 4205, 4207 may be subjected to 100% volumetric examination once the welds are formed. It is to be understood that the closure weld for the inner top lid 4229 is to undergo volumetric examination before the outer top lid 4231 put in place. The dual-walled MPC 4201 also includes a first plate that acts as an inner base plate 4265 and a second plate that acts as an outer base plate 4267. The inner and outer base plates 4265, 4267 are rigid plate-like structures having circular horizontal cross-sections. The invention is not so limited, however, and the shape and size of the base plates is dependent upon the shape of the inner and outer shells. The inner base plate 4265 includes a top surface 4269, a bottom surface 4271 and an outer lateral surface/edge 4273. Similarly, the outer base plate 4267 includes a top surface 4275, a bottom surface 4277 and an outer lateral surface/edge 4279. The top surface 4269 of the inner base plate 4265 forms the floor of the cavity 4203. The inner base plate 4265 rests atop the outer base plate 4267. Similar to the other corresponding components of the dual-walled MPC 201, the bottom surface 4271 of the inner base plate 4265 is in substantially continuous surface contact with the top surface 4275 of the outer base plate 4267. As a result, the interface between the inner base plate 4265 and the outer base plate 4267 is free of gaseous gaps/voids for thermal conduction optimization. An explosive joining, a cladding process, a roller bonding process and/or a mechanical compression process can be used to effectuate the contact between the base plates 4265, 4267. Preferably the thickness of the inner base plate 4265 is in the range of 50 mm to 150 mm. The thickness of the outer base plate 4267 is preferably in the range of 99 mm to 200 mm. Preferably, the length from the top surface of the outer top lid 4231 to the bottom surface of the outer base plate 4267 is in the range of 4000 mm to 5000 mm, but the invention is in no way limited to any specific dimensions. The outer base plate 4267 may be equipped on its bottom surface with a grapple ring (not shown) for handling purposes. The thickness of the grapple ring is preferably between 50 mm and 150 mm. The outer diameter of the grapple ring is preferably between 350 mm and 450 mm. The inner shell 4205 rests atop the inner base plate 4265 in a substantially upright orientation. The bottom edge 4215 of the inner shell 4205 is connected to the top surface 4275 of the inner base plate 4265 by a through-thickness single groove (V or j shape) weld. The outer surface 4210 of the inner shell 4205 is substantially flush with the outer lateral edge 4273 of the inner base plate 4265. The outer shell 4207, which circumferentially surrounds the inner shell 4205, extends over the outer lateral edges 4273, 4279 of the inner and outer base plates 4265, 4267 so that the bottom edge 4227 of the outer shell 4207 is substantially flush with the bottom surface 4277 of the outer base plate 4267. The inner surface 4221 of the outer shell 4207 is also connected to the outer base plate 4267 using a through-thickness edge weld. In an alternative embodiment, the bottom edge 4227 of the outer shell 4207 could rest atop the top surface 4275 of the outer base plate 4267 (rather than extending over the outer later edge of the base plate 4267). In such an embodiment, the bottom edge 4227 of the outer shell 4207 could be welded to the top surface 4275 of the outer base plate 4267. When all of the seal and closure welds discussed above are completed, the combination of the inner shell 4205, the inner base plate 4265 and the inner top lid 4229 forms a first hermetically sealed structure surrounding the cavity 4203, thereby creating a first pressure vessel. Similarly, the combination of the outer shell 4207, the outer base plate 4267, and the outer top lid 4231 form a second sealed structure about the first hermetically sealed structure, thereby creating a second pressure vessel about the first pressure vessel and the cavity 4203. Theoretically, the first pressure vessel is located within the internal cavity of the second pressure vessel. Each pressure vessel is engineered to autonomously meet the stress limits of the ASME Code with significant margins. FIG. 49B illustrates a single-walled MPC 4285 which is constructed in a similar manner as each pressure vessel of the double-walled MPC 4201 discussed above. This single-walled MPC 4287 includes a side wall 4289 seal welded to a base plate 4291, and a top plate 4293. The top surface 4295 of the top plate 4293 includes a beveled top edge 4297 and a channel 4299 set in from the top edge 4297. Having the lid configured with the channel 4299 makes it so that the closure weld may be subjected to 100% volumetric examination. All other parts of the single-walled MPC 285 may be constructed in the same manner described above. A detailed view a top plate 4311 and the closure weld 4313 sealing the top plate 4311 to a side wall 4315 of an MPC are illustrated in FIG. 49C. The channel 4317 in the top surface 4319 is set in from the beveled top edge 4321. The channel 4317 extends below the top surface 4319 at least as much as does the bevel of the beveled top edge 4321. In some embodiments, depending upon the configuration of the probe being used, it may be desirable to have the channel 4317 extend deeper below the top surface than the bevel in order to accommodate the probe. The channel 4317 is sufficiently wide so that a probe used for examining the closure weld may be placed within the channel 4317 and moved circumferentially around the top plate 4311 for purposes of achieving 100% volumetric examination of the closure weld. For some types of probes, the channel may be as wide as 2″ to 3″, although these dimensions may vary significantly to accommodate the configuration of the probe used to examine the closure weld. The side wall 4323 of the channel 4317 nearest the beveled top edge 4321 is placed at an angle that is approximately parallel to the angle of the beveled top edge 4321. However, in some embodiments the angle of this channel side wall may vary from the angle of the top beveled edge by 5°-20° or more, depending upon the configuration of probe being used. The side wall 4323, however, may be formed at any angle relative to the beveled top edge 4321. The opposite wall 4325 of the channel 4317 may have any configuration, from a well-defined wall, as is shown, to a curved or flat surface adjoining the bottom 4327 of the channel 4317. One embodiment of a top plate 4331 is shown in FIG. 50 with ports 4333 positioned in the central portion 4335 of the top surface 4337 of the top plate 4331, radially inward from the channel 4339. The ports 4333 may serve any desired purpose for the MPC for which the top plate 4331 is used, and the different ports may be used for different purposes. Examples of purposes for the ports include their use as vent ports, as vacuum ports, as drain ports, as backfill ports, as test ports, among others. Another embodiment of a top plate 4341 is shown in FIG. 51. In this embodiment, the ports 4343 are positioned within the channel 4345. In other embodiments, ports may be positioned both within the channel and in the central portion of the top surface of the top plate. FIGS. 52A and 52B illustrate the process of performing the 100% volumetric examination of the closure weld alter it has been formed. With the top plate in place on the top opening of the sidewall, the top plate having a channel as described above, the closure weld may be formed by automated equipment, such as is well known in the art. In order to volumetrically examine the closure weld, probes are mounted on a support arm capable of rotating and positioning the probes to perform the volumetric examination of the closure weld. For example, the probes may be mounted on the same type of weld arm that is used in the automated process for forming the closure weld. The volumetric examination may be carried out once the entire weld is formed. Only the end of the support arm 4371 is illustrated in FIG. 52A to simplify the drawing. It is to be understood that the support arm may have any appropriate configuration that is capable of supporting the probes and moving them around the top plate to perform the volumetric examination, as mam different types and configurations of such support arms are well-known in the arts, including combination rotary/articulating robotic arms. Two probes 4373, 4375 are affixed to the end of the support arm 4371, and the support arm is configured for automated or remote positioning of the probes so that the volumetric examination of the closure weld may be performed. The first probe 4373 is positioned on the outside of the top of the side wall 4377, and the second probe 4375 is shown just prior to being positioned within the channel 4379 formed in the top surface 4381 of the top plate 4383. This is second probe 4375 is shown positioned within the channel 4379 in FIG. 52B. Once the two probes are in position, the entire volume of a portion of the closure weld is disposed between the two probes, and that entire volume may be volumetrically examined. By activating the two probes and moving them synchronously around the top plate, maintaining their relative position with respect to the closure weld, the entirety of the weld is passed between the two probes in one circumscription of the top plate. It is therefore possible, with the appropriate examination technology, to perform a 100% volumetric examination of the closure weld. Using well-known processes associated with the selected examination technology, the integrity of the entire closure weld may be determined from the examination. In the embodiment of FIG. 52A, the entire closure weld is formed first, followed by the volumetric examination of the closure weld. In the embodiment of FIG. 52B, the weld head 4385 extends from the same support arm (not shown in FIG. 52B) as the probes 4373, 4375. The weld arm then moves the weld head around the top edge of the top plate to form the closure weld, and the probes trail the weld head to perform the volumetric examination. This embodiment may be used to form the weld and substantially concurrently volumetrically examine the weld. For a multi-pass closure weld, having the probes trail the weld head in this manner enables a separate volumetric examination of each pass of the closure weld. Due to the heat generated from the welding process, which may interfere with the examination process, this embodiment may be best suited for use in pools or in the presence of a coolant, such as a flow of demineralized water In certain embodiments, a Linear Scan-Phased Array UT system may be used to examine the closure weld, and for such embodiments the probes are ultrasound transducer probes. Such a UT system is capable of conducting the 100% volumetric examination of the closure weld within a matter of minutes. Beneficially, with the top plate configured as described above and with use of the two probes, no human activity needs to be directly involved for placing the top plate, forming the closure weld, or examining the integrity of the closure weld, so that work crews are not exposed to any significant doses of radiation. In embodiments where a UT system is used outside of a pool of water or other fluid, a coupling agent, such as demineralized water or an appropriate gel, may be introduced between the transducer probes and the top plate and/or side wall to increase the amount of ultrasound energy that passes into the closure weld, thereby improving the volumetric examination. As is well known in the art of UT, only small amounts of the coupling agent are needed to form a thin film, minimizing air gaps, between the transducer probe and the parts of the MPC into which the ultrasound energy is being directed. Therefore, a simple drip system suffices to introduce a coupling agent such as demineralized water to the process of volumetric examination described herein. In embodiments involving a high heat load canister, to ensure that the metal temperature of the weld mass is not too high for an accurate UT reading, it may be necessary to circulate cooling water through the MPC using the vent and drain ports in the lid before performing the volumetric examination. As an alternative, the use of a coupling agent for ultrasound energy, such as demineralized water, between the transducer probes and the MPC helps to insure that the volumetric examination is performed at a uniform temperature, thereby preserving the UT calibration integrity. With reference to FIGS. 53-59, a sixth inventive concept will be described. FIG. 53 illustrates an apparatus for transferring spent nuclear fuel in the form of a transfer cask 5011. The transfer cask 5011 includes a cylindrical inner shell 5013 which forms a cavity 5015 along with the top lid 5017 and the bottom lid 5019. As shown, a canister 5021 for holding spent nuclear fuel is disposed within the cavity 5015. The inner shell 5013 has a longitudinal axis 5023, and the inner shell 5013 has a slightly larger radius, measured from the longitudinal axis 5023, as compared to the canister 5021, to create an annulus 5025 of space between the inner shell 5013 and the canister 5021 disposed in the cavity 5015. This annulus 5025, as discussed in greater detail below, serves to enable cooling of the canister 5021 by ventilation with atmosphere. The transfer cask further includes an intermediate shell 5027 and an outer shell 5029. Each of the inner shell 5013, the intermediate shell 5027, and the outer shell 5029 are preferably made from carbon steel, with the top of each welded to a top flange 5031, and the bottom of each welded to a bottom flange 5033. The intermediate shell 5027 is disposed concentrically around and spaced apart from the inner shell 5013, thereby forming a second annulus 5035. This second annulus 5035 is capable of holding a gamma absorbing material such as concrete, lead, or steel. Lead is preferred because it most effectively provides gamma shielding for the radioactive spent nuclear fuel once it is placed within cavity 5015. The outer shell 5029 is disposed concentrically around and spaced apart from the intermediate shell 5027, thereby forming a third annulus 5037. This third annulus 5037 is capable of holding a neutron absorbing material such as water or the aforementioned aluminum trihydrate-boron carbide-epoxy mixture. As shown, the third annulus 5037 includes panels of a metal matrix composite. For alternative embodiments in which water is to be used in the third annulus, U.S. Pat. No. 7,330,525 describes a manner in which the outer shell may be formed, in order to contain water, and a process for using water as a neutron absorber in the transfer cask during transfer of a canister containing spent nuclear fuel. The top lid 5017 is securable to the top flange 5031 by extending bolts (not shown) through the top lid 5017 to engage the top flange 5031. The top lid 5017 is typically only secured to the top flange 5031 once the canister 5021 is in place within the cavity 5015 during the transfer process. A central opening 5039 in the top lid 5017 provides access to the canister 5021 for performing certain handling operations with respect to the canister 5021 while the top lid 5017 is secured to top flange 5031. Referring to FIG. 54A, the top flange 5031 is integrally formed through forging and machining so that it does not include any joints, welds, or seams, and so that it does not include parts that are separately formed and then subsequently joined together. The top flange 5031 is machined to include two trunnions 5041 to be used for lifting the transfer cask with a crane. As shown in FIGS. 54A-54C, the trunnions may be of a variety of cross sections such as round trunnions 5041 (FIG. 54A), rectangular trunnions 5041b (FIG. 54B), obround trunnions 5041c (FIG. 54C), oblong trunnions, and the like. The cross-sectional form of the trunnions may be any shape according to design choice, with specific implementations limited only by the equipment used to hoist the transfer cask. More than two trunnions may be machined as part of the top flange, based upon design choices and the lifting system with which the transfer cask is to be used. For purposes of stability during lifting, the trunnions are distributed approximately equidistantly around the top flange. The top flange 5031 also includes a seating groove 5043 for a sealing ring (not shown), which serves as a seal, against the canister and within the annulus, when the canister is placed in the cavity. A plurality of ventilation channels 5045 are included in the top flange 5031, with internal channel inlets 5047 on the interior surface 5049 of the top flange 5031 located below the seating 5043 so that when a canister is placed, air is directed through the ventilation channels 5045. The ventilation channels 5045 open up to the exterior of the top flange 5031, and to the exterior of the transfer cask, at external channel outlets 5051 so that the ventilation channels fluidically connect the annulus 5025 with the exterior of the top flange 5031 and the transfer cask. The ventilation channels 5045 through the top flange 5031 may have a variety of forms or paths, however, because air is being used to ventilate the transfer cask, and unlike water, air is not a good neutron absorber, the one design constraint for the ventilation channels is that the paths of the ventilation channels preclude a direct line of travel from within the cavity to the exterior of the top flange. With this design constraint on the ventilation channels of the top flange, emissions from the canister cannot pass through an all-air pathway from the canister to the exterior of the transfer cask. The integral design of the trunnions 5041 as part of the top flange 5031 serves to eliminate joints between the top flange and the trunnions, thereby significantly improving the fidelity of structural integrity of the overall lifting system (as compared to the prior art, in which the trunnions are joined to the top flange by welding or a threaded joint). The top flange 5031 is also enlarged as compared to top flanges of the prior art, but still keeping within the constraints of the size of the cask pit in the pool and the lifting limit of the cask crane. Even though enlarged, the top flange 5031, inclusive of the integral trunnions 5041, has a smaller outer diameter as compared to the outer shell 5029. To aid in preventing damage that may be caused by protruding trunnions in the event of a transfer cask accidentally tipping into other casks, each trunnion 5041 is disposed within a recess 5053 of the top flange 5031. The larger top flange 5031 also serves to provide increased shielding in the top region of the cask where most human activity (to weld and dry the canister) occurs. Turning back to FIG. 53, the bottom lid 5019 is secured to the bottom flange 5033 by a plurality of bolts (not shown) that extend through holes in the bottom flange 5033 the engage the bottom lid 5019. The bottom lid 5019 includes an impact zone 5061 positioned directly beneath the cavity 5015. The bottom lid 5019 also includes a gamma-absorbing layer 5063, such as lead, below the impact zone 5061. To be most effective in absorbing impacts from accidental falls of the transfer cask, the impact zone 5061 extends substantially under the entirety of the cavity 5015. The impact zone includes an impact absorbing structure 5065 which can serve to cushion the fall of a canister loaded into the transfer cask, thereby providing some damage protection to the fuel in the event of a handling mishap while the transfer cask is being moved around the building or plant site. As shown, the impact absorbing structure 5065 is formed by a plurality of cylindrical tubes 5067 within the bottom lid 5019. These tubes 5067 are distributed throughout the impact zone 5061, with their longitudinal axes aligned with a major dimension (i.e., the diameter) of the bottom lid 5019. The thickness, number of tubes, and the cross-sectional shape of the tubes are a matter of design choice based upon the particular implementation. Factors that may be taken into consideration for these design choices include estimated drop height (based on the operational procedures of the facility), the weight of the canister, and the weight of the loaded transfer cask. Computations have shown that a set of parallel 2-inch tubes distributed throughout the impact zone 5061 can limit the impact load experienced by a 40-ton canister, placed with a transfer cask, falling from 18 inches onto a concrete pad to a g-force of less than 25 (in the absence of the impact limiter, the g-force may shoot up to over 100). A plurality of ventilation channels 5071 are included in the bottom lid 5029, with external channel inlets 5073 on the external surface 5075 of the bottom lid and internal channel outlets 5077 located so that the ventilation channels 5071 can direct an air flow into the annulus 5025. A plurality of ventilation channels configured in this manner are formed approximately equidistantly around the bottom lid to provide cooling ventilation to the canister 5021 outside of the storage pool. At the point of intersection between the channel outlets 5077 and the annulus 5025, the bottom flange 5053 is configured with a chamfered surface 5079 to broaden out the annulus 5025, thereby providing an enlarged space about the base of the canister 5021 into which air may be drawn through the ventilation channels 5071. Each channel inlet 5073 is configured to receive a sealing plug (not shown), which may threadably engage the channel inlet 5073 to provide a seal and turn the ventilation channel and annulus into a “blind” cavity that does not have ingress through the bottom lid. Similar plugs may be placed in the channel outlets of the top flange, thereby rendering the entire annuls cavity into a “blind” cavity. Such plugs may be placed under circumstances where it is desirable to protect the ventilation channel from ingress of contaminated water or other matter, either solely at the bottom of the transfer cask, or at the top and the bottom. A second example of a ventilation channel 5081 is shown in FIG. 57, and a plurality of ventilation channels 5081 configured in this manner are formed approximately equidistantly around the bottom lid to provide cooling ventilation to the canister 5021 outside of the storage pool. Again, at the point of intersection between the channel outlets 5083 and the annulus 5025, an enlarged space 5085 is included about the base of the canister 5021 into which air may be drawn through the ventilation channels 5081. The channel inlets 5087 may also be configured to receive a sealing plug (not shown). The ventilation channels 5071 through the bottom lid 5029 may have a variety of forms or paths, however, because air is being used to ventilate the transfer cask, and unlike water, air is not a good neutron absorber, the one design constraint for the ventilation channels is that the paths of the ventilation channels preclude a direct line of travel from within the cavity to the exterior of the bottom lid. With this design constraint on the ventilation channels of the bottom lid, emissions from the canister cannot pass through an all-air pathway from the canister to the exterior of the transfer cask. FIGS. 56 and 57 illustrate another alternative embodiment of the bottom lid 5091 and an integrated ventilation channel. In this embodiment, the ventilation channel is a toroidal-shaped distribution channel 5093 having a single channel inlet 5095 and a plurality of channel outlets 5097 which are positioned to fluidically connect the annulus, formed between the inner shell of the transfer cask and the canister placed in the cavity, with the exterior of the bottom lid 5091 and the transfer cask. The radial position of the channel inlet 5095 is different than the radial position of the channel outlets 5097 so that the configuration of the ventilation channel 5093 precludes a direct line of travel from within the cavity to the exterior of the bottom lid. A transfer cask which includes the annulus between the inner shell and the canister, the ventilation channels in the top flange, and the ventilation channels in the bottom flange, configured in any of the manners discussed above, when out of a storage pool allows ambient air to ventilate up the annulus to enhance the heat removal efficacy of the cask. Calculations have shown that a mere ¾ inch wide annulus can reduce the fuel cladding temperature by as much as an additional 20° C., in comparison to a blind annulus with stagnant air (which is the state-of-the-art). And, as compared to a water-cooled annulus, a passive ambient air-cooled annulus is much simpler, easier to use, and easier to maintain, thereby resulting in greater operational reliability. Such a transfer cask will remove decay heat from the canister by ventilation action. For low heat canisters (those generating less than about 18 kW), the natural ventilation through the annulus coupled with heat dissipation from the external surfaces of the cask are sufficient to keep the contents of the canister from overheating. In circumstances where additional cooling is needed for higher heat load canisters, beyond the cooling that can be provided by ventilation of ambient air, chilled air can be forced through the annulus. One such system is shown schematically in FIG. 58. And, even a forced air system is simpler and easier to use and maintain than a cooled water system. A forced air system is most easily used when the bottom lid includes an integrated ventilation channel with a single channel inlet, such as is shown in FIG. 56. During use, an air compressor 5111 operates to store compressed air in a compressed air tank 5113, and the air outlet 5115 of the compressed air tank 5113 is fluidically coupled though an appropriate air line 5117 to the channel inlet 5119 of the bottom lid of the transfer cask 5121. The compressed air tank 5113 itself may be cooled by ambient air, or it may be cooled by an active cooling or refrigeration system 5123. As those of skill in the art will recognize, decompression of air naturally decreases the temperature of that air, so that the amount of cooling needed for the compressed air tank 5113 will depend upon the heat dissipation needs of the transfer cask. For example, a refrigeration system may be used to cool the compressed air tank to a temperature as low as 5° C., thereby causing the decompressed air from the compressed air tank to be cooler still when it is directed into the annulus of the transfer cask. The decompressed air is delivered into the ventilation channel of the bottom lid, and then into the annulus, by the positive pressure of expansion upon release from the compressed air tank. The air compressor and compressed air tank are sized to provide the cooled air at a sufficiently high velocity to ensure turbulent How conditions within the annulus. Calculations have shown that a 50 MP compressor is adequate to cool a canister with as much as 35 kW heat load. The chilled air is heated within the annulus and exits the transfer cask through the ventilation channels in the flange. As an alternative to using a compressed air tank and an air compressor, chilled air may alternatively be forced into the annulus by use of a blower. The advantages of a forced air cooling system include greater simplicity, as compared to a water cooled system, use of single phase cooling medium (air rather than water) and mitigation of the concerns of leakage (no water spillage) at the flanged or screwed joints. The performance of the system is easily monitored by measuring the temperature of the exiting heated air from the cask. FIG. 59 is a flowchart showing, the process of moving a transfer cask, as described above with ventilation channels, loaded with a canister from a pool for transport or storage of the canister. The process starts 5121 with a fully loaded canister in the cavity of transfer cask without the top lid in place. The process of loading the canister is well-known to those of skill in the art, and so they are not discussed herein. As the transfer cask sits in the pool, one or more plugs may be in place in the bottom lid to seal off the ventilation channels to make the ventilation channels and the annulus a “blind” cavity, thereby protecting from ingress of contaminated water. Without the plugs in place, water fills the annulus and helps to remove heat generated by the spent nuclear fuel in the canister. The hoist of a crane is lowered into the pool and secured to the trunnions of the transfer cask. Once the hoist is secured to the trunnions, the crane lifts 5123 the transfer cask, along with the canister payload, out of the storage pool. The transfer cask is designed so that at this stage in the process, the combined weight of the transfer cask and payload is equal to or less than the rated lifting capacity of the crane. Once lifted out of the storage pool, the crane sets transfer cask down 5125 in a staging area. At this point, the canister contains pool water in addition to the spent nuclear fuel. This pool water acts as a neutron absorber as long as it is in the canister, and it removed from the canister in order to store the spent nuclear fuel in a dry-state. In the event that one or more plugs are in place in the bottom lid, they are removed 5127 to allow ventilated cooling by circulation of atmospheric air through the annulus. As an alternative, at this point, a compressed air tank is fluidically coupled to the channel inlet of the bottom lid using an appropriate hose and coupling. The compressed air tank is coupled to an air compressor so that compressed air is maintained in the tank during use. Compressed air from the tank is decompressed and passed into the channel inlet during the remaining steps of moving the transfer cask while it is loaded with the canister. Once the transfer cask is ventilated, the pool water in the canister is pumped out 5129, and the spent nuclear fuel in the canister is allowed to dry. The canister is then backfilled with an inert gas, such as helium, and sealed. The cask lid is then secured 5131 to transfer cask. The transfer cask is then lifted by the crane and moved to a position above another cask 5133, at which point the bottom lid is removed and the canister is lowered into the other cask 5135. The other cask may be a storage cask, if the spent nuclear fuel is to be stored long-term, or it may be a transport cask suitable for moving spent nuclear fuel over long distances. Once the canister is removed from the transfer cask, the transfer cask may be reused to perform the above described procedure again. To reuse the transfer cask, the one or more plugs are again put in place in the bottom lid to seal off the ventilation channels. As used throughout, ranges are used as shorthand for describing each and every value that is within the range. Any value within the range can be selected as the terminus of the range. In addition, all references cited herein are hereby incorporated by referenced in their entireties. In the event of a conflict in a definition in the present disclosure and that of a cited reference, the present disclosure controls. While the invention has been described with respect to specific examples including presently preferred modes of carrying out the invention, those skilled in the art will appreciate that there are numerous variations and permutations of the above described systems and techniques. It is to be understood that other embodiments may be utilized and structural and functional modifications may be made without departing from the scope of the present invention. Thus, the spirit and scope of the invention should be construed broadly as set forth in the appended claims.
	claims	1. A cylindrical neutron source, comprising: a cylindrical chamber; a plurality of alternatingly nested concentric or coaxial neutron generating targets and plasma ion generating regions, wherein each target is a cylinder and each plasma ion generating region is annular and has an inner and an outer plasma boundary surface; an RF antenna induction coil positioned in each plasma ion generating region; cylindrical extraction electrodes at both the inner and outer boundary surfaces of each plasma ion generating region having a voltage difference applied relative to the adjacent target, the extraction electrodes containing a plurality of longitudinal slots through which the ions are extracted from plasma ion generating regions and directed onto the targets in consequence of the applied voltage. 2. The neutron source of claim 1 wherein the plasma generating regions are deuterium ion sources. 3. The neutron source of claim 2 wherein the targets have titanium surfaces. 4. The neutron source of claim 1 wherein the source has a length of about 35 cm and a diameter of about 50 cm. 5. A cylindrical neutron source, comprising:a cylindrical chamber;at least two plasma ion generating regions and at least two neutron generating targets, a first neutron generating target disposed concentrically as a cylinder inside and spaced from a first plasma ion generating region, a second neutron generating target disposed concentrically as a cylinder outside and spaced from said first plasma ion generating region, and a second plasma ion generating region disposed concentrically outside and spaced from said second neutron generating target, all of said plasma generating regions and said neutron generating targets being disposed within said cylindrical chamber;an RF antenna induction coil disposed within each plasma generating region; anda plurality of cylindrically shaped extraction electrodes having longitudinal slots, wherein each said plasma ion generating region is annular and has an inner and an outer plasma boundary surface, each of said plasma ion generating regions having a first extraction electrode disposed adjacent said inner plasma boundary surface of said plasma ion generating regions, each of said inner plasma boundary surfaces being disposed in each of said plasma ion generating regions so as to be positioned within a portion of said plasma ion generating region that is located toward a center longitudinal axis of said neutron source, and a second extraction electrode disposed adjacent said outer plasma boundary surface of said plasma ion generating region, each of said outer plasma boundary surfaces being disposed in each of said plasma ion generating regions so as to be positioned within a portion of said plasma ion generating region that is located farther from the center longitudinal axis than is said inner plasma boundary surface, and each said first and second extraction electrodes having a voltage difference applied relative to the adjacent target. 6. The cylindrical neutron source of claim 5 wherein the plasma generating regions are deuterium ion sources. 7. The cylindrical neutron source of claim 6 wherein the targets have titanium surfaces. 8. The cylindrical neutron source of claim 5 wherein the source has a length of about 35 cm and a diameter of about 50 cm. 9. The cylindrical neutron source of claim 5 further comprising a plurality of coaxially nested alternating ion sources and targets. 10. The cylinder neutron source of claim 5, wherein each plasma generating region forms an unintermittent plasma.
	description	This application claims priority to and the benefit of co-pending U.S. provisional patent application Ser. No. 61/241,366, filed Sep. 10, 2009, which application is incorporated herein by reference in its entirety. The present invention is directed generally to X-ray spectrometry systems, and in particular, but not necessarily exclusively to a multiple element short working distance spectrometer. Spectrometry systems can be used to confirm the presence of or to determine the concentration, electronic or magnetic properties, or local chemical environment of a given chemical species in a sample, such as in physical or analytical chemistry. Two common spectroscopy methods are absorption spectrometry and fluorescence spectrometry. In absorption spectrometry, a beam of light is sent through a sample to be analyzed, and certain wavelengths of the light are absorbed by the sample. By comparing the wavelengths of the absorbed light to known chemical absorption spectra, the components of the sample may be identified. In fluorescence spectrometry, a sample is bombarded by high energy light or other radiation capable of inducing electronic transitions to higher energy levels. As the excited electrons fall back to lower energy levels, the wavelength of the emitted light can be used to identify numerous atomic-scale properties of the sample. The Rowland circle and either the Johann or Johannson geometries can be employed in spectrometry systems. In this arrangement, a curved, crystal-based (e.g., silicon or germanium) diffraction element is used for wavelength-specific focusing of X-rays. For applications involving higher-energy X-rays, a radius of curvature of one meter or more is typically required. This in turn creates a large working distance (which is a function of the radius of curvature) that results in a poor collection solid angle and creates the need for precision tolerances with little margin for error during fabrication, calibration, and operations. Further, every time such a device is moved, it must be calibrated to these tolerances, meaning that use of these devices can be expensive and time consuming. Similar considerations apply to the von Hamos geometry, which makes use of a simpler cylindrical design with only partial focusing properties. The following examples, references and description provide specific details sufficient for a thorough understanding of embodiments of the disclosure. One skilled in the art will understand, however, that the disclosure may be practiced without certain details. In other instances, well known structures and functions have not been shown or described in detail to avoid unnecessarily obscuring the description of embodiments of the disclosure. As used herein, the term “orientation” refers to the unit normal vector defining a planar face. For example, two planar faces with identical unit normal vectors have the same orientation, despite being in different locations. As used herein, the term “planar face” refers to the single two dimensional surface of an object. For example, the planar face of a cubic-shaped diffraction crystal includes the two dimensional surface facing an X-ray source but does not include the surface of the 5 other faces of the cubic-shaped crystal. As used herein, the term planar refers to a generally flat surface. However, in some embodiments, planar can include weakly bent or slightly curved surfaces or substantially planar surfaces with localized regions of bending, curvature, or damage. As used herein, the term “X-ray” refers to the band of the electromagnetic spectrum that extends from 0.01 nm to 10 nm in wavelength and 120 eV to 120 keV in energy. In one embodiment, the X-ray energy range extends from about 3 keV to 10 keV in energy. In general, embodiments of spectrometers and spectrometry systems can employ a variety of X-ray energy ranges. Moreover, embodiments of spectrometers and spectrometry systems can be configured for high energy resolution. For example, embodiments of spectrometers and spectrometry systems can have high energy resolution on the order of 1 eV or higher. Briefly, the invention is directed to a short working distance spectrometer that is less expensive, smaller, more portable, and easier to fabricate and calibrate in comparison to current spectrometer systems. Embodiments of the spectrometer employ planar Bragg diffraction elements coupled to a rigid body, which can simplify fabrication and reduce cost relative to conventional spectrometers. In addition, embodiments of the spectrometer can be manufactured to be significantly smaller than conventional spectrometers, especially for higher energy X-rays at 1 eV or finer resolution (such conventional spectrometers having working distances on the order of 1 m or more). Embodiments of the spectrometer also employ other features, which offer a substantial improvement over conventional spectrometers, including reduced cost, easier fabrication, customization, and higher collection solid angles. In one embodiment, the spectrometer includes a rigid body having a first planar face with an orientation and a second planar face with a different orientation than the first planar face. The first and second planar faces are configured to position Bragg diffraction elements, and the orientation of the first planar face and the different orientation of the second planar face are arranged to convey a predetermined spectral range (e.g., based in part on the orientation of the planar face) of the electromagnetic radiation to non-overlapping regions of the sensor location via the diffraction elements. In contrast to conventional spectrometers, and in particular, in contrast to conventional systems employing “diced” optics, embodiments of the spectrometer employ detection of non-overlapping regions of X-ray radiation. For example, not only do these spectrometers have a long working distance, they employ X-ray reflections that intentionally overlap at a detector. FIG. 1A is a schematic diagram of a system 100 for conducting short working distance spectrometry. FIG. 1B is a top plan view of a sensor array employable in the system 100 and will be described in conjunction with FIG. 1A. The system 100 includes an X-ray source 102, a housing 110, a sample holder 112 disposed in the housing 110, an apparatus 120 disposed in the housing 110 for positioning Bragg diffraction elements 130, a sensor array 160, and a signal processing component 170 electrically coupled to the sensor array 160 via a signal path 161. In the configuration of FIG. 1A, the system 100 is generally considered to be in a short-working distance configuration because of the smaller dimensions employed relative to conventional spectrometry systems. For example, a distance L1 between the sample holder 112 and the apparatus 120 (or individual Bragg diffraction elements 130 or planar faces holding the Bragg diffraction elements) can be on the order of centimeters. In one embodiment, the dimension L1 can be in the range of about 1 to 15 cm. The housing 110 further includes an entrance aperture 114, an exit spatial filter 115, and an exit aperture 116 located in the exit spatial filter 115. The entrance aperture 114 is typically on the order of several millimeters in diameter and allows fluorescence from a sample to enter while preventing or reducing stray scatter. The sample holder 112 is configured to hold samples to be scanned in a position and at an orientation so that the X-ray beam passes directly through the sample. In general, the position of the sample holder 112 relative to the other components of the system 100 may affect the energy range of the X-ray beam and the diffraction at the Bragg diffraction elements 130. The exit aperture 116 is positioned to filter the diffracted light from the diffraction elements 130, to prevent or reduce stray scatter from reaching the sensor array 160, and to select a predetermined range of diffracted electromagnetic radiation incident at the sensor array 160. The exit aperture 116, for example, may include an opening formed in a material that is opaque to X-ray radiation. In some embodiments, the housing 110 may be omitted from the system 100. For example, the sample holder 112, the entrance aperture 114, and/or the exit aperture 116 may each be arranged as stand-alone components that are not connected to the housing 110. The sensor array 160 is spaced apart from the exit spatial filter 115 by a distance L2. The sensor array 160 is arranged to receive electromagnetic radiation and to output signals indicative of the spectrum of wavelengths (or equivalently of photon energies) associated with detected radiation. In one embodiment, for example, the sensor array 160 include a two-dimensional position-sensitive detector, such as, for example, a camera device or a diode array, that is arranged for detecting X-rays. FIG. 1B shows non-overlapping illumination of regions 162 of the sensor array 160. In some embodiments, there is an exclusive one-to-one correspondence between diffraction elements 130 and the non-overlapping regions 162 of the sensor array 160. In other embodiments, if the reflected radiation from different diffracting elements overlaps at certain sub-regions of the sensor array 160, such sub-regions can be excluded as necessary in subsequent signal processing. Also, while drawn as square-shaped in the figures, the non-overlapping regions 162 will generally have more complex shapes or profiles that can be different from one another in the sensor array 160 based on the overall geometry of the system 100. The locations of the non-overlapping regions 162 can be defined, at least in part, by the configuration of the apparatus 120 and the exit aperture 116 (described further with reference to FIGS. 3A and 3B). The non-overlapping regions 162 of the sensor array 160 can be arranged to receive electromagnetic radiation and to output one or more output signals indicative of the intensities and locations of the received radiation. In some embodiments, the sensor array 160 may be flat, as shown in the figures, or may instead take a more complex shape, such as with tiled components on the surface of a cylinder. The signal processing component 170 is configured to receive the output signals from the sensor array 160 and to provide a variety of data indicative of the electromagnetic radiation 307a and 307b (described further with reference to FIGS. 4A and 4B). FIG. 2 is a partially exploded, isometric diagram of the apparatus 120. The apparatus 120 includes a rigid body 222 and a plurality of planar faces 224 carried by the rigid body 222. The apparatus may be manufactured from a variety of materials. In some embodiments, the rigid body 222 includes a plastic material. In one embodiment, for example, the rigid body 222 may be manufactured by a machine that “prints” in three dimensions via additive manufacturing techniques. For example, RepRap (http://reprap.org/wiki/Mainpage) and Stratasys (http://www.stratasys.com/, based in Eden Prairie, Minn.) provide three-dimensional printers. In such embodiments, the apparatus 120 can be quickly fabricated. For example, a manufacturing engineer can design the apparatus using a computer aided design (CAD) program and create an output file that can be read by a three-dimensional file. In other embodiments, however, conventional subtractive manufacturing techniques, such as electric discharge machining and multi-axis computer-controlled mills, can be used to create the apparatus in several materials, including steel, aluminum, or other metallic materials and/or alloys. The planar faces 224 are configured to position Bragg diffraction elements 130. The planar faces 224, for example, may have a surface area in the range of a few mm to 25 mm square. The Bragg diffraction elements 130 can be attached to the planar faces 224 via an adhesive, epoxy, or the like. The Bragg diffraction elements 130 can include a variety of materials having one or more compositional layers for diffracting electromagnetic radiation. Embodiments of the Bragg diffraction elements 130 include crystalline materials common in the semiconductor industry (e.g., silicon and/or germanium), other highly-crystalline materials (e.g., diamond, quartz, lithium fluoride, or beryl), and multi-layered materials (e.g., artificial multi-layers of silicon and molybdenum). The orientations of the planar faces 224 are configured for directing electromagnetic radiation via the Bragg diffraction elements 130. For example, a first planar face 224a has an orientation N1 for directing a refracted X-ray towards the exit spatial filter 115 and ultimately one of the non-overlapping regions 162 of the sensor array 160 (see FIGS. 3A and 3B). Similarly, a second planar face 224b has a different orientation N2 for directing a refracted X-ray towards the exit spatial filter 115 and ultimately a different non-overlapping region of the sensor array 160. The orientation of the planar faces 224 may be solved as an inverse problem constrained by the desired spatial clearances around the sample 304, the distances L1 and L2, the location and dimensions of the exit spatial filter 115, the spatial resolution of sensor array 160, the desired energy resolution, and the location and dimensions of sensor array 160. This includes, but is not limited to, solutions based on the geometry required by the Rowland circle and either Johannson or Johann geometry, or by the von Hamos geometry. See, for example, the methods for orientating Bragg diffraction elements disclosed in the parent application, U.S. provisional application No. 61/241,366. In one embodiment, the orientations of the planar faces 224 are configured to provide diffracted electromagnetic radiation with generally the same energy or wavelength range to the sensor array. For example, in such an embodiment, the Bragg diffraction elements 130 (when attached to the planar faces 224) and the exit spatial filter 115 (FIG. 1A) may be arranged to provide a generally similar range of energy to each of the non-overlapping regions 162. In another embodiment, the orientations of the planar faces 224 are configured to provide diffracted electromagnetic radiation with different energy ranges. For example, the Bragg diffraction elements 130 (when attached to the planar faces 224) can each provide a different portion of an electromagnetic spectrum to different non-overlapping regions of the sensor array 160. In such an embodiment, the apparatus 120 can be configured for the detection of a specific sample with several different specific bands of emitted radiation. In such cases, the system 100 can be configured to include additional exit spatial filters for accommodating the different bands of emitted radiation. FIG. 3A is a schematic diagram showing paths of electromagnetic radiation in the spectrometry system 100. FIG. 3B is a top plan view of the sensor array 160 and will be described in conjunction with FIG. 3A. To simplify the description, FIG. 3A only shows diffraction from two diffraction elements. In general, however, embodiments of the spectrometry system 100 can employ numerous diffraction elements for refracting numerous, non-overlapping paths of electromagnetic radiation at a sensor array. The X-ray source 102 provides X-ray beam 303 to a sample 304 carried by the sample holder 112. The sample 304 emits electromagnetic radiation through the entrance aperture 114, as stimulated by the details of the incident radiation and the microscopic composition and structural details of the sample 304. A portion 305a of the electromagnetic radiation is incident onto a Bragg diffraction element 330a, and another portion 305b of the electromagnetic radiation is incident onto a Bragg diffraction element 330b. The Bragg diffraction elements 330a and 330b diffract electromagnetic radiation 306a and 306b towards the exit spatial filter 115. As discussed above, the direction of diffraction at individual Bragg diffraction elements is based on the orientation of the planar faces 224 of the apparatus 120 (see FIG. 2) and the choice of material and crystalline orientation of the respective diffraction elements 130. The exit spatial filter 115, in turn, receives and filters the electromagnetic radiation 306a and 306b to provide electromagnetic radiation 307a and 307b to the sensor array 160. The aperture 116 is generally configured to allow a portion of the electromagnetic radiation 306a and 306b to pass through the exit spatial filter 115 while substantially blocking the remaining portion of the electromagnetic radiation 306a and 306b. In particular, the exit aperture 116 is configured to direct the electromagnetic radiation 307a and 307b to non-overlapping regions of the sensor array 160. FIG. 3B shows the illuminated non-overlapping regions 362a and 362b corresponding to the electromagnetic radiation 307a and 307b, respectively. As shown, the electromagnetic radiation 307a and 307b overlaps (at region 308) upon exiting the exit spatial filter 115. The electromagnetic radiation 307a and 307b, however, does not overlap (at region 309) when reaching the sensor 160. The locations of the region 308 and the region 309 can be controlled by the size of the aperture 116. For example, the exit spatial filter 115 will block less electromagnetic radiation as the size of the aperture is increased. If the aperture is too large, however, the region 309 will not exist and the non-overlapping regions 162 will not be spaced apart from one another. The locations of the region 308 and the region 309 can also be controlled by the distance L2. Decreasing the distance L2 decreases the size of the region 309 and the spacing distance S1 between the non-overlapping regions 162. Increasing the distance L2 increases the size of the region 309 and the spacing distance S1 between the non-overlapping regions 162. In some embodiments, the exit aperture 116 is configured to filter a predetermined energy range of the electromagnetic radiation 307a and 307b. A larger aperture, for example, will filter less electromagnetic radiation than a smaller aperture. FIG. 4A is a schematic diagram of the signal processing component 170. The signal processing component 170 is arranged to output information corresponding to an emission spectrum. For example, the signal processing component can output information corresponding to an absorption spectrum or a fluorescence spectrum of a sample at the sample holder 112 (FIG. 1A). In general, the signal processing component 170 may have a variety of configurations. For example, a personal computer may serve as the signal processing component 170 and may run one or more software applications in conjunction with various hardware for processing signals and outputting information, such as to a display, a printer, or the like. Alternatively, the signal processing component may be a standalone device or the like, including a microcontroller or other hardware device. As shown in the figure, the signal processing component 170 is arranged to receive input signals from the sensor array 160 via one or more signal paths 472 from the sensor array 160 (FIG. 1A). The processing component 170 is further arranged to output signals via one or more signal paths 474 based on the electromagnetic radiation detected at the sensor array 160. In general, the signal processing component 170 employs a processing unit 476 and a memory 478. The memory 478 may include RAM, ROM, and the like to provide processor executable instructions for calibration, for storing data, for outputting data, and for performing various operations based on the input signals 472. FIG. 4B is a flow diagram showing a process 480 for employing the signal processing component 170 to providing data indicative of an emission spectrum. The process 480 begins at block 481 where the signal processing unit receives signals from the sensor array 160. The signals can be voltage, current, charge, or optical signals representing time-varying information indicative of the spatial distribution of X rays incident at one or more of the non-overlapping regions 162. The non-overlapping regions 162 may each convey an image showing the spatial distribution of X-ray intensity incident on the sensor array 160. The process 480 continues to block 482, where the signal processing component 170 calculates an electromagnetic spectrum based on the input signals from the sensor array 160. The signal processing component 170 may employ any one of a myriad of algorithms for determining a spectrum, including statistical inferential methods or other signal processing algorithms. The signal processing component 170 can also apply band-pass or other filters for pre-processing of signals from sensor array 160 or post-processing of the calculated spectrum. The central aspect to be considered is the calibration of the sensor array 160, specifically the determination of the wavelength or energy of X-ray photons expected to be incident at each spatially resolved point on sensor array 160. In some cases, this can be obtained as a pure consequence of conventional ray-tracing analysis of the entire system 100. In some applications, however, this will be directly determined by use of reference radiation sources with known distributions of X-ray wavelengths (spectra). Such reference experiments will give highly reliable calibration of some subset of spatial subregions (e.g. pixels) on the sensor array 160. With such partial information, the entire sensor array 160 can then be reliably calibrated using interpolation methods, statistical inferential methods, or other standard numerical techniques. The process 480 continues to block 483, wherein the signal processing component 170 aggregates the spectrum collected at each of the non-overlapping regions 162. For example, referring to FIG. 3A, the signal processing component 170 can aggregate the spectrum determined for each of the non-overlapping regions 362a and 362b. In one embodiment, the statistical accuracy of the detection can be enhanced by analyzing similar spectral ranges at each of the non-overlapping regions 162. In another embodiment, a spectral range is based on different spectral ranges detected at each of the non-overlapping regions 162. In other embodiments, the non-overlapping regions 162 may represent the same energy range but may not be combined, as the different regions give different spectral information as a consequence of polarization of the emitted electromagnetic radiation or other advanced physical phenomenon. The process 480 continues to block 484, where a spectrum indicative of the sample is output. For example, the output can be provided at a display, stored in memory, or output to a printer. At decision block 485, processing can return to block 480 for analyzing another sample (or re-analyzing the sample). Alternatively, processing can terminate. It should be noted that the invention is not limited to the specific processing blocks or order thereof. For example, in some embodiments, aggregation at the block 483 can be performed prior to the calculation of the electromagnetic spectrum at block 482. From the foregoing it will be appreciated that representative embodiments have been described for purposes of illustration. However, well known characteristics often associated with spectrometry systems have not been described to avoid unnecessarily obscuring the various embodiments. In addition, various modifications may be made to the various embodiments, including adding or eliminating various features. For example, while the apparatus 120 is shown as having multiple arrays of planar faces for a supporting Bragg diffraction elements, other embodiments of the apparatus 120 may only employ a single array. Also, the apparatus 120 may include more or fewer planar faces than those illustrated. In some embodiments, the planar faces may be slightly non-planar such that the planar face have a weak bend or have a slight curvature. For example, it is contemplated that a weak bend or slight curve in the (slightly non-planar) face and/or corresponding Bragg diffraction element can enhance the diffraction of X-rays from a Bragg diffraction element. The apparatus 120 may also incorporate other modifications, such as to increase reflectivity of the diffracting elements by the application of appropriate strains. For example, the diffraction elements can be substantially strained, including being weakly bent or otherwise intentionally damaged to decrease crystallinity, so as to advantageously influence their integrated reflection efficiency. Moreover, embodiments of the system 100 may include other modifications or components. For example, a variety of X-ray emitters may be used in lieu of the X-ray source 102 and the sample 304. The X-rays emanating from the X-ray emitter may, for example, be fluorescence excited by continuous or pulsed ionizing radiation such as X-rays or high energy charged particle beams, such as in an electron microscope, or may be the result of elastic or inelastic scattering of incident X-ray radiation. For example, an electron microscopy system may use embodiments the apparatus 120 in an imaging modality with extremely high spatial resolution (e.g., the imaging modality is sensitive to the local magnetic or chemical properties of a sample scanned by the electron beam). Alternatively, the X-ray emitter may itself be radioactive or can be artificially induced by several means. The X-ray emitter may itself constitute a high temperature plasma, such as in fusion experiments or some extrasolar bodies (such as studied by X-ray telescopes), where several physical effects induce fluorescence or other emission modes at X-ray wavelengths.
053217314	description	DETAILED DESCRIPTION OF THE INVENTION In order to appreciate the conventional wisdom of steam separator and dryer design, reference is made FIG. 1 which shows a simplified BWR generally at 10. BWR 10 may be seen to comprise a reactor pressure vessel (RPV), 12, configured to admit feedwater via a feedwater inlet, 14, and to exhaust steam via a steam outlet, 16, thus providing for the ultimate operation of the BWR. A nuclear core, 18, is provided and is disposed within a core shroud, 20. Core shroud 20 and RPV 12 define a core annulus region, 22, through which sub-cooled water flowing downwardly through a downcomer region, 24, may enter a core lower plenum region identified at 26 before flowing through core 18. A water and steam mixture exits core 18 and flows into a core upper plenum region identified at 28, which is defined by a shroud head, 30, disposed atop core 18. From core upper plenum 28, the two-phase mixture enters a plurality of standpipes, 32, which, together with RPV 12, define downcomer region 24. Standpipes 32 are disposed atop shroud head 30 and extend in fluid communication between core upper plenum 28 and a corresponding number of individual steam separators, 34. Steam separators 34 have outlet communications for water separated from the two-phase mixture flowing therethrough to enter downcomer region 24. The separated water in downcomer region 24 combines with the feedwater entering from inlet 14 to provide an accumulation of water for continuous and endless flow through core 18. A representative water level is shown at 36. Steam separators 34 also have outlet communications for wet steam to pass into a wet steam plenum region identified at 38. Conventionally, a separate dryer assembly, 40, is provided having inlets, 42, in fluid communication with wet steam plenum 38. The collective steam throughput of steam separators 34 is, accordingly, passed into wet steam plenum 38 and then, via inlets 42, into dryer assembly 40. Water removed from the steam is returned to downcomer region 24 via drains, 44. The dried steam, essentially water-free, is passed via outlets, 46, into a steam dome region, identified at 48, to be withdrawn from RPV 12 via steam outlet 16. A wall, 45, separates dryer assembly 40 and steam dome region 48 from wet steam plenum 38. Looking now to FIG. 2, the instant, inventive modular steam separator with integrated dryer is shown generally at 50 as a steam separator, 54, having an integrated steam dryer, 56. Steam separator 54, for example, may be of a conventional centrifugal type. See U.S. Pat. No. 3,902,876 and Wolf et al., "Advances in Steam-Water Separators for Boiling Water Reactors", ASME Paper No. 73-WA/Pwr-4, November 1973. Steam dryer 56, for example, may have internal vanes. Modular steam separator with integrated dryer 50 may be especially adapted for employment in either forced or natural circulation BWRs such as, for example, conventional BWRs, the advanced BWR (ABWR), and the simplified BWR (SBWR). Reactor internals, construction and operation are well known in the art, such as illustrated by reference to the following publications: Glasstone and Sesonske, Nuclear Reactor Engineering, pp. 748-753, 3d Edition, VanNostrand, Reinholt (New York, N.Y., 1981); Wolfe and Wilkens, "Improvements in Boiling Water Reactor Designs and Safety", presented at the American Nuclear Society Topical Meeting, Seattle, Wash., May 1-5, 1988; Duncan and McCandless, "An Advanced Simplified Boiling Water Reactor", presented at the American Nuclear Society Topical Meeting, Seattle, Wash., May 1-5, 1988; and Lahey and Moody, The Thermal Hydraulics of a Boiling Water Nuclear Reactor, especially Chapter 2, pp. 15-44, American Nuclear Society (LeGrange Park, Ill. 1977). Conventional BWRs, the ABWR and the SBWR all are described and discussed in the foregoing references, each of which is expressly incorporated herein by reference. Advantageously, a multiplicity of modular steam separators with integrated dryers 50 may be incorporated into a BWR, thereby eliminating the need for a wet steam plenum, and discrete steam separator and dryer components. The incorporation of modular steam separator with integrated dryer 50 into a BWR may be effected by connecting the proximal end of standpipe 58 of steam separator 54 to an outlet in core upper plenum 60. Core upper plenum 60 may be seen to be defined by shroud head 61 of core shroud 65. A two-phase water and steam mixture, 62, formed from the passage of feedwater and recycled coolant in heat transport relationship with the reactor core may be passed into steam separator 54 from core upper plenum 60 via standpipe 58. As water and steam mixture 62 is transported through barrel 64 of steam separator 54, a helical motion may be imparted thereto by an inlet swirler positioned at the proximal end of barrel 64 for the purpose of generating centrifugal forces to effect separation of the entrained water from the steam. As a result of the helical motion imparted to water and steam mixture 62 by the inlet swirler, a portion of the denser water phase, 66, of water and steam mixture 62 is directed to vanes 68a-c in each of three representative stages of steam separator 54. In each of the representative stages shown, separated water phase 66 is removed form barrel 64 via, respectively, pickoff rings 70a-c and is passed via annuli 72a-c into downcomer region 74 for addition to the feedwater of the reactor via core annulus 75 which is defined by the region bounded by shroud 65 and reactor pressure vessel 77. Annuli 72a-c are formed within the regions bounded by outer skirts 76a-c and vanes 68a-c. To prevent any entrained steam separated by a free surface separation mechanism from water phase 66 from bypassing steam dryer 56, steam connection 73 is provided for its passage to steam dryer 56 via barrel 64. Separated steam phase 78 is passed from the distal end of steam separator 54 into steam dryer 56 via a steam inlet distribution channel, 80. Preferably, steam inlet distribution channel 80 decreases in cross-sectional area as it extends in fluid communication from barrel 64 into dryer 56. Within dryer 56, any unseparated water, 82, is removed from steam phase 78 and is passed into downcomer region 74 via dryer drain 84. Drain 84 may be seen to extend below water level 85 for addition of unseparated water 82 to the feedwater of the reactor via core annulus 75. Dried steam phase 90, now essentially moisture-free, is passed via vent 86 into steam dome region 88 for ultimate use as the working fluid for a turbine used, for example, in the generation of electric power. A wall, 76, separates steam dome region 88 from downcomer region 74. Preferably, vent 86 increases in cross-sectional area as it extends away from steam dryer 56. Regarding materials of construction, preferably all components are manufactured from materials appropriate for their use within a nuclear BWR. Since certain modification can be made in accordance with the precepts of the present invention, the description herein is illustrative rather than limitative.
	summary
	claims	1. A nuclear reactor containment system comprising:a nuclear reactor;a container enclosing the nuclear reactor, the container including:at least one heat removal system having an active state and an inactive state, wherein the at least one heat removal system dissipates heat from the container more efficiently in the active state than in the inactive state, and wherein the at least one heat removal system is structured to switch from the inactive state to the active state based on a temperature of the container,wherein the at least one heat removal system includes a first heat removal system including:fins disposed in an outer portion of the container and forming a plurality of cooling channels; anda plurality of air regulating mechanisms structured to block air from flowing through the cooling channels when the first heat removal system is in the inactive state and to allow air to flow through the cooling channels when the first heat removal system is in the active state,wherein at least one of the air regulating mechanisms includes:a first plate structured to be disposed over a first cooling channel of the plurality of cooling channels to block airflow through the first cooling channel;a pivot structured to support a first side of the first plate;an electromagnetic element structured to support a second side of the first plate via an electromagnetic force and being structured to support the second side of the first plate via the electromagnetic force at a predetermined temperature,wherein when the first heat removal system is in the inactive state, the electromagnetic element supports the second side of the first plate and the first plate blocks airflow through the first cooling channel, andwherein when the first heat removal system is in the active state, the electromagnetic element melts and allows the second side of the plate to fall into the first cooling channel and allow airflow through the first cooling channel. 2. The nuclear reactor containment system of claim 1, wherein the at least one heat removal system is structured to switch from the inactive state to the active state at one or more predetermined temperatures of the container above temperatures of the container corresponding to normal operation of the nuclear reactor. 3. The nuclear reactor containment system of claim 1, wherein at least one of the air regulating mechanisms includes:a bimetallic strip disposed in a second cooling channel of the plurality of cooling channels, the bimetallic strip being structured to bend based on temperature,wherein the bimetallic strip is structured to be bent in a direction to block airflow in the second cooling channel when the first heat removal system is in the inactive state and to be bent in a direction to allow airflow through the second cooling channel when the first heat removal system is in the active state. 4. The nuclear reactor containment system of claim 1, wherein at least one of the air regulating mechanisms includes:a second plate structured to be disposed over a second cooling channel of the plurality of cooling channels to block airflow through the second cooling channel;a pivot structured to support a first side of the second plate;a fusible element structured to support a second side of the second plate and being structured to melt at a predetermined temperature,wherein when the first heat removal system is in the inactive state, the fusible element supports the second side of the second plate and the second plate blocks airflow through the second cooling channel, andwherein when the first heat removal system is in the active state, the fusible element melts and allows the second side of the second plate to fall into the cooling channel and allow airflow through the second cooling channel. 5. The nuclear reactor containment system of claim 1, wherein at least one of the air regulating mechanisms includes:a second plate structured to be disposed over a second cooling channel of the plurality of cooling channels to block airflow through the second cooling channel,wherein the second plate is comprised of a melting material,wherein when the first heat removal system is in the inactive state, the second plate blocks airflow through the second cooling channel, andwherein when the first heat removal system is in the active state, the second plate is structured to melt and allow airflow through the second cooling channel. 6. The nuclear reactor containment system of claim 1, wherein the at least one heat removal system includes a second heat removal system structured to dissipate heat to ground. 7. The nuclear reactor containment system of claim 6, wherein the second heat removal system includes:a primary chamber disposed in a bottom portion of the container, the primary chamber having a first material disposed therein; anda secondary chamber disposed above the primary chamber and around a perimeter of the container, the secondary chamber having a second material disposed therein,wherein the first material is a porous material having a higher melting point than the second material,wherein when the second heat removal system is in the inactive state, the second material does not melt, andwherein when the second heat removal system is in the active state, the second material melts and flows into the primary chamber. 8. The nuclear reactor containment system of claim 7, wherein the first material has a porosity in a range of about 30-80%. 9. The nuclear reactor containment system of claim 7, wherein the second material as at least one of chips, spheres, or powder in its solid state. 10. The nuclear reactor containment system of claim 6, further comprising:a heat spreader base plate in contact with a bottom surface of the container;a base area disposed below the container; anda number of heat conductive plates extending from the heat spreader base plate into the base area. 11. The nuclear reactor containment system of claim 10, wherein the base area is comprised of concrete. 12. The nuclear reactor containment system of claim 1, wherein the container has a substantially cylindrical shape and is comprised of a plurality of layers, the plurality of layers including:a neutron absorber layer;a gamma shield layer; anda container vessel wall. 13. The nuclear reactor containment system of claim 12, wherein the plurality of layers further includes:a neutron reflector layer. 14. The nuclear reactor containment system of claim 1, further comprising:an underground vault structured to receive the container, the underground vault including a recess to receive the container and a surrounding barrier area having a plurality of heat conductive plates formed therein. 15. The nuclear reactor containment system of claim 1, further comprising:a cask structured to receive the container; anda number of impact limiters structured to limit impacts to the container during transport. 16. The nuclear reactor containment system of claim 1, wherein a top portion of the container includes a lid having a lug, wherein the container is structured to be lifted via the lug. 17. A nuclear reactor containment system comprising:a nuclear reactor;a container enclosing the nuclear reactor, the container including:at least one heat removal system having an active state and an inactive state, wherein the at least one heat removal system dissipates heat from the container more efficiently in the active state than in the inactive state, and wherein the at least one heat removal system is structured to switch from the inactive state to the active state based on a temperature of the container,wherein the at least one heat removal system includes a first heat removal system structured to dissipate heat to groundwherein the first heat removal system includes:a chamber disposed in a bottom portion of the container, the chamber having a top side, a bottom side, and sidewalls;a flexible liner disposed in the chamber, the flexible liner having a thermally conducting fluid disposed therein; andthermally expanding elements disposed on the sidewalls of the chamber, the thermally expanding elements being structured to expand inward and reduce an interior volume of the chamber as temperature rises,wherein when the first heat removal system is in the inactive state, the interior volume of the chamber is large enough that the flexible liner and thermally conducting fluid do not contact the top of the chamber, andwherein when the first heat removal system is in the active state, the thermally expanding elements expand and reduce the interior volume of the chamber such that the flexible liner and thermally conducting fluid are pressed upward to contact the top of the chamber.
	description	X-ray tubes for the production of x-rays for imaging including the medical imaging of human patients are typically provided with filters which condition the x-ray beam used for imaging. These filters attenuate certain x-rays in the beam to better suit the x-ray beam to a particular imaging task. For instance, for the medical imaging of human patients the softer x-rays which are likely to be absorbed by the tissue of the patient are filtered out of the beam. Some conditioning requires the use of metal filters which are adhered to the x-ray tube. For instance, in x-ray tubes for the medical imaging of human patients a 75 micron sheet of pure copper is adhered to the aluminum window which allows the x-rays to exit the tube for imaging. In some cases this copper sheet is protected by a polymer sheet which is in turn affixed to the aluminum window such that the x-rays used for imaging pass through the thickness of the copper sheet. In one embodiment an x-ray system includes x-ray tube containing a cathode which supplies electrons and an anode which can be maintained at high positive electrical potential to the cathode and has a target area which is impacted by electrons from the cathode when the positive potential is maintained generating x-rays. The x-ray tube is enclosed in a radiation resistant casing. The radiation resistant casing has a window which allows some of the x-rays generated at the target area to exit the system. The window has a vapor deposited layer of a filtering metal of sufficient thickness to effectively condition the x-rays passing through the window and located to intercept the x-rays passing through the window. Another embodiment includes a process for the construction of an x-ray system. The x-ray system includes one or more conditioning filters having an x-ray tube. The x-ray tube includes a cathode which supplies electrons and an anode which can be maintained at high positive electrical potential to the cathode and has a target area which is impacted by electrons from the cathode when the negative potential is maintained generating x-rays which is enclosed in a radiation resistant casing. The casing has a window which allows some of the x-rays generated at the target area to exit the system. The window has one or more layers of a filtering metal of sufficient thickness to effectively condition the x-rays passing through the window and located to intercept the x-rays passing through the window. A layer of filtering metal is vapor deposited on a surface of the window. Referring to FIG. 1 one embodiment involves an X-ray system with an x-ray tube 100 which contains a cathode 110 and an anode 120. A fairly high vacuum is maintained in the interior of the x-ray tube. The cathode 110 is heated to provide a source of free electrons. The anode 120 is maintained at a high positive potential relative to the cathode 110 which causes the free electrons to accelerate and strike the anode 120 at a high velocity generating x-rays. Some of these x-rays pass through a radiation emission passage 130 in the electron collector 150. These x-rays then pass through the x-ray tube wall 140 via a beryllium window 170, which is essentially transparent to x-rays but provides structural integrity to the gap in the x-ray tube wall 140. The beryllium window 170 provides for the emission of the x-rays at the high operating temperatures of the x-ray tube 100. The x-rays then pass through the dielectric oil circulation path 160 and through the radiation resistant casing 180 via an aluminum window 200 and an Ultem window 210. In one embodiment the Ultem window is polyetherimmide, however other materials may also be used. The radiation resistant casing 180 contains a dielectric oil which cools the x-ray tube and minimizes the probability of arcing between the electrical connections for the anode 120 and the cathode 110. The aluminum window 200 provides a path for the x-rays through the radiation resistant casing 180. The Ultem window 210 is essentially transparent to the x-rays but provides protection for the exterior surface of the aluminum window 200. Referring to FIG. 2 a physical vapor deposited copper filter 220 of about 75 microns thickness is shown on the exterior surface of the aluminum window 200. The exposed surface of this filter 220 is protected from handling damage by the Ultem window 210 and by the fact that the aluminum window 200 has a recessed exterior surface. In one embodiment Ultem window 210 may be removed. Referring to FIG. 3 the x-ray beam 230 is shown passing through the dielectric oil circulation path 160 and then the aluminum window 200 and the physical vapor deposited copper filter 220, which both condition the beam by the absorption or attenuation of some of the x-rays. In this way the X-ray beam is conditioned in this embodiment to be suitable for medical imaging of human patients. Among other things the “softer” x-rays likely to be absorbed by the tissue of the patient have been attenuated or removed. Finally, the beam 230 passes through the Ultem filter 210, which is essentially transparent to the x-rays. The layer of filtering metal may be deposited by any of the known physical or chemical vapor deposition methods with the modification that the deposited layer is sufficiently thick to effectively condition the x-rays passing through it. Vapor deposition methods provide for a reproducible deposit of a layer of reasonably uniform composition and thickness so that the x-ray conditioning behavior is fairly consistent within each window and between multiple windows manufactured to the same specification. This deposition technique minimizes the formation of mixtures of the substrate material, i.e. the surface of the window on which the layer is being deposited, and the metal being deposited. This facilitates the design of windows with particular conditioning characteristics as these characteristics can be reproducibly predicted from the identity and thickness of the filtering layer without having to account for the effect of unintended mixtures of materials. Suitable physical vapor deposition techniques include those involving creating a vapor of the metal to be deposited under reduced pressure of an inert gas and causing the vapor to condense on the substrate which will carry the layer by the application of an electrical potential. In one embodiment a magnetron is used to generate the vapor. In one embodiment the surface to which the layer is to be applied is cleaned by bombardment of ionized atoms of the inert gas before exposing this surface to the vapor. It is convenient to use an inert gas whose atomic number is reasonably close to that of the metal being deposited. For instance it is convenient to use argon when creating a layer of copper. Suitable chemical vapor deposition techniques include those involving a chemical reaction which results in a vapor of the metal to be deposited being condensed on the substrate which will carry the layer. It is convenient to create a vapor of a chemical compound involving the desired metal and then to release the metal from the compound creating a vapor of the metal itself. The layer of vapor deposited metal should be thick enough to cause a significant attenuation of the x-rays passing through it. Conditioning the x-ray beam with a filter removes or at least greatly reduces the presence of certain components of the x-ray spectrum generated by the impact of electrons on the anode target. For instance, X-rays used for the medical imaging of human patients are commonly filtered through a thick copper layer to remove all or a substantial portion of the “softer” x-rays which are likely to be absorbed by the tissue of the human patient as opposed to passing through this tissue. In one embodiment the vapor deposited metal layer is at least about 10 microns in thickness. In another embodiment it is between about 10 and 200 microns in thickness. In a further embodiment it is between about 50 and 150 microns in thickness. It may be convenient to employ a thickness of about 75 microns, particularly if the deposited metal is copper. The thickness may be readily tailored to achieve a desired x-ray conditioning effect. The thickness of the layer of vapor deposited metal should be fairly uniform and reproducible between windows carrying such layers. In one embodiment the thickness is within plus or minus 2 microns of the nominal thickness intended. Thus for this embodiment a layer with a target thickness of 33 microns the thickness observed across a number of windows carrying such layers should be between 31 and 35 microns. This may be contrasted to the plus or minus seven micron tolerance common when the filter is formed from a rolled sheet material as opposed to a vapor deposited layer. The layer may be of any metal which is amenable to one or more vapor deposition techniques and has desirable x-ray conditioning properties. These metals include Aluminum, Copper, Molybdenum, Tin, Titanium, Tungsten and Zirconium. It is not necessary that the metal have good cold or hot workability or ductility. More than one layer may be deposited on the window to condition the x-rays. By varying the identity and thickness of multiple layers of vapor deposited metal conditioning effects can be tailored to meet particular needs. In one embodiment a layer of copper is vapor deposited followed by a layer of carbon and then followed by a layer of titanium. In this case one of the filtering layers was not a metal but it was sandwiched between two vapor deposited metals. In one embodiment the copper layer is about 50 microns in thickness while that of the carbon is about 25 microns and that of the titanium is about 40 microns. The layer is deposited on a surface of the window through which the x-ray beam used for imaging passes in exiting the radiation resistant casing. In one embodiment this is the external surface of the window. Because the vapor deposited metal forms a good bond with the surface on which it is deposited it is also possible to place it on the interior surface of the window which faces the dielectric oil circulation path without undue concern that the oil will wick between the layer and the window surface. The window may be constructed of any of the materials commonly used to allow the emission of an x-ray beam from an x-ray system with a radiation resistant casing. In one embodiment the window is fabricated of an appropriate material and has an appropriate thickness to contribute to the conditioning of the x-ray beam for its intended use. In one embodiment the window is fabricated of high purity aluminum. In one embodiment the window is inset such that its exterior surface is closer to the interior of the system than the exterior surface of the adjacent portion of the radiation resistant casing into which it is placed. This provides a recess which protects the surface of the vapor deposited layer from handling damage and also minimizes the thickness of the dielectric oil circulation path which passes in front of the window. This in turn minimizes the probability that turbulence or air bubbles in the dielectric oil will cause x-ray artifacts when the system is used for imaging. The combination of the vapor deposited layer or layers and the window conditions the x-ray beam exiting the radiation resistant casing for imaging a particular type or class of target. In one embodiment the x-ray beam is conditioned for the medical imaging of human patients. A copper layer of about 33 microns was generated on a high purity aluminum window fixture using a magnetron based physical vapor deposition process to yield composite suited to serve as a window for the radiation resistant casing enclosing an x-ray tube and to condition the x-rays passing through it for medical imaging of human patients. The window fixture was ultrasonically cleaned in alcohol for 5 minutes . and all surfaces of window fixture that were not to be coated with copper were masked. The window fixture was then placed in a vacuum chamber which was then evacuated to 3.0×10−6 Torr. The window fixture was held in vacuum at less than 3.0×10−6 Torr for one hour. The window fixture was then subject to 2 minutes of Argon ion scrubbing at 2.0 kV and 89 mA in a 17.5 mTorr Argon atmosphere. A copper vapor was supplied to the coating chamber by energizing a magnetron to 500 Watts for 60 minutes using a ramp rate of 8 seconds. A Torus Magnetron system from Kurt J Lesker™ Vacuum (Product Number TM3FS10XBS) with a 3″ diameter target was used. A 500VDC bias voltage was applied to the window fixture. The chamber pressure was adjusted to about 5 mTorr so that the magnetron plasma current and voltage were 1.59A and 256V, respectively. The deposition rate on the fixture was in excess of about 0.11 Å/sec. After about 60 minutes a copper layer of 33 microns had been generated. An elemental analysis of a cross section of a composite prepared in the manner described in the coating example at various distances from the aluminum/copper junction was performed by scanning electron microscopy (SEM) using a sampling square of 1.6 microns. The results are reported in the table below. Distance in Microns fromWt % of Elements DetectedBoundary Starting from Cu sideAlCuSiAgFe12100.0090.6599.3560.5799.4336.4493.56141.8457.410.180.57049.6950.160.150.557.9240.890.160.370.66173.3024.280.211.051.16280.7114.810.321.842.31384.329.020.302.703.663-491.742.660.335.28488.884.070.474.002.595.591.941.370.465.560.67 The results show that the degree of intermixing of the aluminum and copper is minimal enough that the two layers can function effectively in conditioning x-rays as distinct layers. It is important to note that the construction and arrangement of an x-ray system as described herein is illustrative only. Although only a few embodiments of the present invention have been described in detail in this disclosure, those skilled in the art who review this disclosure will readily appreciate that many modifications are possible (e.g. variations in sizes, dimensions, structures, shapes and proportions of the various elements, values of parameters, mounting arrangements, use of materials, colors, orientations, etc.) without materially departing from the novel teachings and advantages of the subject matter recited in the claims. For example, elements shown as integrally formed may be constructed of multiple parts or elements and vice versa, the position of elements may be reversed or otherwise varied, and the nature of number of discrete elements or positions may be altered or varied. Accordingly, all such modifications are intended to be included within the scope of the present invention to be included within the scope of the present invention as defined in the appended claims. The order or sequence of any process or method steps may be varied or re-sequenced according to alternative embodiments. Other substitutions, modifications, changes and omissions may be made in the design, operating conditions and arrangement of the exemplary embodiments without departing from the scope of the present inventions as expressed in the appended claims.
	summary
	claims	1. A device comprising:an enclosing structure defining a closed loop flow path, said enclosing structure formed with an inlet and an outlet;a system generating plasma at a plasma site, the site being in fluid communication with said flow path;a gas source connected to said inlet; anda conditioner connected to said outlet to condition gas exiting said enclosing structure. 2. A device as recited in claim 1 wherein the plasma comprises tin and the filter removes a compound selected from the group of compounds comprising tin hydrides, tin oxides and tin bromides. 3. A device as recited in claim 1 wherein said conditioner is selected from the group of conditioners consisting of a gas dilution mechanism, a scrubber or a combination thereof. 4. A device as recited in claim 1 wherein said enclosing structure comprises a vessel in fluid communication with a guideway external to the vessel. 5. A device as recited in claim 1 further comprising a pump forcing said gas through said closed loop flow path. 6. A device as recited in claim 1 further comprising a heat exchanger removing heat from gas flowing in said flow path. 7. A device as recited in claim 1 further comprising a filter removing at least a portion of a target species from gas flowing in said flow path. 8. A device comprising:an enclosing structure defining a closed loop flow path;a system generating plasma from a source material at a plasma site, the site being in fluid communication with said flow path, the source material comprising tin;a gas comprising hydrogen disposed in said enclosing structure;a filter removing at least a portion of a target species from gas flowing in said flow path. 9. A device as recited in claim 8 wherein the plasma comprises tin and the filter removes a compound selected from the group of compounds comprising tin hydrides, tin oxides and tin bromides. 10. A device as recited in claim 9 wherein the filter comprises a cold trap. 11. A device as recited in claim 9 wherein the filter comprises a zeolite filter. 12. A device as recited in claim 9 wherein the filter comprises a chemical absorber. 13. A device as recited in claim 9 wherein the filter comprises a plurality of internally cooled metal plates to condense vapors. 14. A device as recited in claim 9 further comprising a pump forcing said gas through said closed loop flow path. 15. A device as recited in claim 9 further comprising a heat exchanger removing heat from gas flowing in said flow path. 16. A device comprising:an enclosing structure defining a closed loop flow path passing through said through-hole;an EUV reflective optic positioned in said enclosing structure;a temperature control system cooling the optic;a system generating plasma at a plasma site, the site being in fluid communication with said flow path; anda gas disposed in said enclosing structure. 17. A device as recited in claim 16 wherein the optic is a near normal incidence mirror. 18. A device as recited in claim 16 wherein the EUV reflective optic has a substrate and wherein at least one cooling channel is formed in the substrate to pass a heat exchange fluid. 19. A device as recited in claim 16 further comprising a heat exchanger removing heat from gas flowing in said flow path. 20. A device as recited in claim 16 wherein the system generates plasma from a source material comprising tin and the gas comprises hydrogen.
	summary
	claims	1. A computer implemented method of monitoring the health of a system comprising:storing a first plurality of system data associated with at least one steady state capacity level in a memory, wherein the memory is accessible by a building supervisory controller;obtaining a second plurality of system data indicative of system conditions during system operations;identifying from the second plurality of system data that the system was operating at a normal operating level during at least one sample window of the system operations;associating the normal operating level of the at least one sample window with the at least one steady state capacity level;retrieving first health data for parameters associated with a first health condition from the first plurality of system data;retrieving second health data for parameters associated with the first health condition from the second plurality of system data that were obtained during the at least one sample window;comparing the first health data with the second health data;determining by the building supervisory controller if the second health data is indicative of a health condition based upon the comparison; anddisplaying the results of the determination. 2. The method of claim 1, wherein: the at least one sample window comprises a plurality of sample windows; anddetermining comprises:determining, for one of the plurality of sample windows, the number of data of the second health data compared with first health data;determining, for the one of the plurality of sample windows, the number of data of the second health data that indicate a health condition; anddetermining that the second health data is indicative of a health condition for the one of the plurality of sample windows if the number of data of the second health data that were compared with first health data for the one of the plurality of sample windows is the same as the number of data of the second health data during the one of the plurality of sample windows that are indicative of a health condition. 3. The method of claim 1, wherein: storing a first plurality of system data comprises storing, for each of a plurality of steady state capacity levels, an associated first plurality of system data;identifying that the system was operating at a steady state capacity level during at least one sample window of the system operations comprises identifying that the system was operating at a steady state capacity level during a plurality of sample windows;associating the normal operating level comprises associating the normal operating level of each of the plurality of sample windows with one of the plurality of steady state capacity levels;retrieving first health data comprises retrieving first health data for parameters associated with a first health condition from the first plurality of system data for each of the associated one of the plurality of steady state capacity levels;retrieving second health data comprises retrieving, for each of the plurality of sample windows, second health data for parameters associated with the first health condition from the second plurality of system data that were obtained during the respective one of the plurality of sample windows;comparing comprises comparing, for each of the plurality of sample windows, the first health data with the second health data; anddetermining comprises determining, for each of the plurality of sample windows, that the second health data is indicative of a health condition during a particular sample window of the plurality of sample windows if the number of data of the second health data that were compared with first health data for the particular sample window is the same as the number of data of the second health data during the particular sample window that are indicative of a health condition. 4. The method of claim 3, determining further comprising:identifying the total number of the plurality of sample windows;identifying the number of the plurality of sample windows indicative of a health condition; andgenerating a percentage of the total number of the plurality of sample windows indicative of a health condition. 5. The method of claim 4, wherein generating a percentage comprises:generating a percentage of the total number of the plurality of sample windows indicative of a health condition for each data interval of an analysis window. 6. The method of claim 5, wherein displaying comprises displaying the results of the determination in near real time. 7. The method of claim 5, further comprising:providing an alert if the percentage exceeds an alert threshold; andproviding an alarm if the percentage exceeds an alarm threshold. 8. The method of claim 1, wherein storing a second plurality of system data indicative of system conditions during system operations comprises:placing the system in a normal operating mode; andstoring the second plurality of system data indicative of system conditions during normal system operations. 9. The method of claim 1, further comprising:obtaining the first plurality of system data associated with at least one steady state capacity level from design parameters for the system. 10. The method of claim 1, further comprising:operating the system at the at least one steady state capacity level; andobtaining the first plurality of system data while the system is operated at the at least one steady state capacity level. 11. The method of claim 1 wherein comparing comprises:comparing, for each of the parameters associated with a first health condition, each of the second health data corresponding to the parameter with the first health data corresponding to the parameter. 12. The method of claim 11, wherein the first health condition comprises one of the group of health conditions consisting of: noncondensables in a chiller system; low refrigerant; oil in an evaporator; fouled condenser tubes; fouled evaporator tubes; low condenser water flow rate; low chilled water flow rate; water bypassing condenser tubes; and water bypassing evaporator tubes.
063109298	description	DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS Embodiments of an in-core fixed nuclear instrumentation system and a power distribution monitoring system according to the present invention will be described hereinafter with reference to the accompanying drawings. [First embodiment] FIG. 1 is a block diagram showing a schematically structure of a reactor power distribution monitoring system of a boiling water type reactor (BWR) according to a first embodiment of the present invention. In the reactor power distribution monitoring system shown in FIG. 1, the same component or configuration as the power distribution monitoring system of the BWR shown in FIG. 23 to FIG. 25 will be described with the use of like reference numerals. As shown in FIG. 1, a reactor power distribution monitoring system 29 of a boiling water type reactor includes an in-core fixed nuclear instrumentation system 30 having detectors and signal processing units, and a process control computer 31 for monitoring an operating mode of a reactor and a core performance. The process control computer 31, as shown in FIG. 1, comprises, for example, a CPU 60, a memory unit 61, an input console 62 and a display unit 63. The CPU 60 is electrically connected to the memory unit 61, input console 62 and the display unit 63 so as to communicate to each other. The process control computer 31 has a function for simulating a core power distribution of the BWR, and a function for monitoring a core performance of the BWR according to the simulated core power distribution. On the other hand, in the BWR, a reactor pressure vessel 2 is housed in a primary containment vessel 1, and a reactor core 3 is housed in the reactor pressure vessel 2. The core 3 is cooled by a coolant used as a moderator. In the reactor core 3, a large number of fuel assemblies 4 are mounted as shown in FIG. 2 and FIG. 3. In large number of fuel assemblies 4, four fuel assemblies is constructed as one group, a control rod 5 having a cross-like shape in its lateral cross section is mounted between each of the four fuel assemblies constructed as one group so as to be taken in and out from below of each of the fuel assembly group. The reactor core 3 is constructed by mounting large number of groups of four fuel assemblies, and is provided with a plurality of, for example, 52 in-core nuclear instrumentation assemblies 32 which function as a detect unit of the reactor nuclear instrumentation system. Each in-core nuclear instrumentation assembly 32 is arranged at a position different from the place where the control rod 5 is located, and is located at a corner water gap G formed between four fuel assemblies 4 as shown in FIG. 2 and FIG. 3. More specifically, the in-core nuclear instrumentation assembly 32 includes a thin and long tube-like nuclear instrumentation tube 33, a neutron detector assembly (LPRM detector assembly) 34 functioning as a fixed neutron detection means (LPRM) and a .gamma.-ray heating detector assembly (GT assembly) 35 functioning as a fixed .gamma.-ray detecting means (gamma-ray thermometer), which are housed in the nuclear instrumentation tube 33, respectively. The LPRM detector assembly 34 is constructed in a manner that a plurality of LPRM detectors 37 functioning as the fixed neutron detectors are discretely arranged in a core axial direction at equal intervals in the nuclear instrumentation tube 33. In the boiling water type reactor, in general, four LPRM detectors 37 are discretely arranged on a fuel effective portion of the core 3 in the core axial direction at equal intervals. Further, each LPRM detector 37 is adapted to detect neutron flux so as to generate a neutron flux signal (LPRM signal) according to the measured neutron flux. Moreover, each LPRM detector 37 is electrically connected to an LPRM signal processing unit 40 by means of a signal cable 38 through a penetration portion 39, and thus, a power range neutron flux measuring system 41 is constructed. The LPRM signal processing unit 40 include a computer having a CPU, a memory unit and so on, is operative to perform, for example, A/D conversion operation and gain processing operation of each LPRM signal S2 transmitted from each LPRM detector 37 so as to obtain digital LPRM signals (LPRM data) D2, and then, transmits them to the process control computer 31. On the other hand, the GT assembly 35 is constructed in a manner that a plurality of fixed .gamma.-ray heating detectors 44 are discretely arranged in a core axial direction, and a gamma-ray heating value is measured by means of each .gamma.-ray heating detector. The same number or more as LPRM detectors 37 arranged in the core axial direction, for example, eight (8) .gamma.-ray heating detectors 44 are arranged in the core axial direction, and a assembly of the .gamma.-ray heating detectors is constructed as the gamma-ray thermometer assembly (GT assembly) 35. Each .gamma.-ray heating detector 44 of the GT assembly 35 is electrically connected to a gamma-ray thermometer signal processing unit 48 by means of a signal cable 45 through a penetration portion 49, and thus, a gamma-ray thermometer power distribution measuring system 50 is constructed. The gamma-ray thermometer signal processing unit 48 (hereinafter, also described as GT signal processing unit 48) is, as shown in FIG. 4, a computer which includes a CPU 48A, a memory unit 48B, an operation panel 48C and a display panel 48D or the like. The CPU 48A is electrically connected to the memory unit 48B, the operation panel 48C and the display panel 48D so as to communicate to each other. The GT signal processing unit 48 is operative to, according to the GT signals S1 outputted from each gamma-ray heating detector 44 of each GT assembly 35 and each sensitivity S.sub.0 of each gamma-ray heating detector 44, obtain digital .gamma.-ray heating measurement signals (GT signals D1; or described as GT data D1, hereinafter) each representing a .gamma.-ray heating value (W/g) per unit weight. The GT signal processing unit 48 is also operative to transmit the obtained GT data D1 to the process control computer 31. More specifically, the fixed in-core nuclear instrumentation system 30 includes the aforesaid power range neutron measuring system 41 and the gamma-ray thermometer power distribution measuring system 50. The in-core nuclear instrumentation assembly 32 (the LPRM detector assembly 34 and the GT assembly 35) including a group of detectors 37 and 44 of the fixed in-core nuclear instrumentation system 30, is adapted to measure neutron flux and a .gamma.-ray heating value in the core 3 as core nuclear instrumentation data (GT data D1 and LPRM data D2) by means of transmission operation of the detected signals of the detectors 37 and 44 and the signal processing of signal processing units 40 and 48 at predetermined fixed measurement points in the core 3. Moreover, each of the GT assembly 35 has a built-in heater wire. The fixed in-core nuclear instrumentation system 30 has a gamma-ray thermometer heater control unit 53 (hereinafter, also described as GT heater control unit) 53 which is electrically connected to each built-in heater (described later) so as to supply a power source to each built-in heater and is operative to control the power supply to each built-in heater. The GT heater control unit 53 is a power supply unit which includes a power supply circuit, a current measurement circuit, a voltage measurement circuit, a voltage control circuit (microcomputer) and an electrically energizing changeover circuit and is operative to apply a voltage to the built-in heater of the GT assembly 35 which is selected via a power cable 54 so that the heater of the selected GT assembly 35 is heated. In the in-core fixed nuclear instrumentation system 30, since no traversing neutron detector and traversing .gamma.-ray detector is required, it is possible to omit a mechanical driving and operating mechanism included in the conventional reactor nuclear instrumentation system. Therefore, it is possible to simplify a structure of the in-core fixed nuclear instrumentation system 30. In addition, in the nuclear instrumentation system 30, no movable parts are required; therefore, it is possible to achieve a maintenance free, and to avoid or greatly reduce a radiation exposure work of workers. Moreover, in the reactor pressure vessel 2 or a primary system piping (not shown), a core state data measuring device 55 is provided. The core state data measuring device 55 measures core state data S3 (process quantity) including a control rod pattern, a core coolant flow rate, an internal pressure of the reactor pressure vessel, flow of feed water, a temperature of feed water (a core inlet coolant temperature) and so on, which are used as various operating parameters indicative of a reactor operating mode (state) of the reactor. The core state data measuring device 55 is shown as one measuring device in the primary containment vessel 1, simplified in FIG. 1; however, in fact, the core state data (process data) measuring device 55 is a core state data measuring means which is composed of a plurality of measuring devices for measuring or monitoring a plurality of core state data (process data), located inside and outside of the primary containment vessel 1. Further, one part of the core state data measuring devices 55, which is inside of the vessel 1, is connected to a core state data processing unit 58 via a signal cable 57 penetrating through a penetration portion 56, and other part thereof, which is outside of the vessel 1, is connected via the signal cable 57 to the core state data processing unit 58, so that a process data measuring system 59 is constructed. The core state data processing unit 58 receives the core state process data S3 (analog signals or digital signals) measured by the core state data measuring device 55, and then, carries out data processing on the basis of the received core state process data S3 to simulate (calculate) a reactor thermal output, a core inlet coolant temperature and so on. Further, the core state data processing unit 58 converts the core state process data S3 including the calculated reactor thermal output and so on into digital core state data D3, and then, transmits the digital data D3 to the process control computer 31. The core state data processing unit 58 of the process data measuring system 59 is not constructed as a dedicated independent unit, but may be constructed as a part of processing functions (modules) of the process control computer 31. In other words, the process data measuring system 59 may be constructed as a part of processing functions including the in-core power distribution simulating function (module) of the process control computer 31. Furthermore, the process data measuring system 59 may be constructed as a part of the reactor in-core nuclear instrumentation system 30 from the concept of detector and signal processing system. In addition, the core state data processing unit 58 of the process data measuring system 59, the LPRM signal processing unit 40 of the power range neutron flux measuring system 41, the gamma-ray thermometer heater control unit 53 of the gamma-ray thermometer heater control system and the GT signal processing unit 48 of the gamma-ray thermometer power distribution measuring system 50, are individually electrically connected to the process control computer 31. A group of data processed by processing units 40, 48 and 58, that is, the core nuclear instrumentation data (GT data D1 and LPRM data D2) and the core state data D3, are transmitted to the process control computer 31 so as to be inputted therein by an interface process of the process control computer 31. The CPU 60 of the control process computer 31 has a nuclear instrumentation control process module (process function) 60M1 as a part of elements of the in-core nuclear instrumentation system 30 for controlling the LPRM signal processing unit 40, the GT signal processing unit 48 and the gamma-ray thermometer heater control unit 53 constituting the in-core nuclear instrumentation system 30, in addition to the aforesaid interface process on the basis of a nuclear instrumentation control program module PM1 memorized in the memory unit 61. The CPU 60 of the process control computer 31 has a power distribution simulation process module M2 (process function) for simulating neutron flux distribution in the core 3, a power distribution therein and a margin with respect to an operational thermal limit according to a power distribution simulation program module PM2 including a physics model (three-dimensional thermal-hydraulics simulation code), wherein the power distribution simulation program module PM2 is memorized in the memory unit 61. The power distribution simulation process module (process function) M2 is also operative to correct the simulation result simulated thereby so as to obtain a core power distribution reflecting the actually measured core nuclear instrumentation data on the basis of a power distribution learning (adaptive) program module PM3 memorized in the memory unit 61. The process control computer 31 is also capable of receiving various commands such as GT calibration instruction command, power distribution simulating command and the like, which are inputted by an input operation of an operator through the input console 62. The CPU 60 is also operative to output a simulated result including, for example, the power distribution and the margin with respect to the operational thermal limit, a display information such as warning or the like to the operator through the display unit 63. The in-core power distribution simulation process module M2 of the CPU 60 simulates the neutron flux distribution, the core power distribution and the margin with respect to the operational thermal limit in the core 3 so as to store the simulated result including the neutron flux distribution and the core power distribution in the memory unit 61. Then, the power distribution simulation process module M2 of the CPU 60 corrects the simulated result (the neutron flux distribution and the core power distribution) stored in the memory unit 61 according to the inputted GT data D1, or the GT data D1 and the LPRM data D2 so as to determine an accurate core power distribution and an accurately margin with respect to the operational thermal limit, which reflect the actually core nuclear instrumentation data (the GT data D1, the LPRM data D2) and have a high reliability. As described above, each of the modules M1 and M2 of the CPU 60 are realized as processing functions of the CPU 60 on the basis of the program modules PM1, PM2 and PM3. By the way, the in-core nuclear instrumentation assemblies 32 constitute a part of the in-core nuclear instrumentation system 30 of the BWR, as shown in FIG. 1 to FIG. 3, and the core 3 is provided with a large number of, for example, 52 in-core nuclear instrumentation assemblies 32. The in-core nuclear instrumentation assembly 32 is arranged at a corner water gap G position surrounded by four fuel assemblies 4. The in-core nuclear instrumentation assembly 32 includes a nuclear instrumentation tube 33, the neutron detector assembly (LPRM detector assembly) 34 functioning as fixed neutron detecting means, the .gamma.-ray heating detector assembly (GT assembly) 35 functioning as fixed gamma-ray detection means (gamma-ray thermometer). Further, the in-core nuclear instrumentation assembly 32 is constructed in a manner that the LPRM detector assembly 34 and the GT assembly 35 are combined so as to be integrally arranged in the nuclear instrumentation tube 33. The LPRM detector assembly 34 constitutes a local power range monitor (LPRM) as a nuclear fission ionization chamber, and has N (N.gtoreq.4), for example, four LPRM detectors 37 which are discretely arranged in a fuel effective portion in a core axial direction at equal intervals. Incidentally, each interval between each LPRM detector 37 is referred to "L". The GT assembly 35 is inserted into the nuclear instrumentation tube 33 together with the LPRM detector assembly 34. The GT assembly 35 includes eight (8) or nine (9) gamma (.gamma.) ray heating detectors 44 which are discretely arranged in the core axial direction. Each neutron detector 37 of the LPRM detector assembly 34 and each gamma-ray heating detector 44 of the GT assembly 35 are housed in the nuclear instrumentation tube 33, and a coolant is guided so as to flow through the nuclear instrumentation tube 33 from the lower portion of the tube 33 in a mounted state to the upper end thereof. In FIG. 2 and FIG. 3, there is shown in an example of the GT assembly 35 which is constructed in a manner that eight (8) .gamma.-ray heating detectors 44 are arranged in a fuel effective portion H of the core axial direction. As shown in FIG. 3, the fuel effective portion H is indicative of a range where a nuclear fuel is effectively filled along the core axial direction in each fuel element (nuclear fuel filled in a fuel rod), and also, the fuel effective portion H along the core axial direction is described as a fuel effective length. Each arranging distance (or interval) between .gamma.-ray heating detectors 44 in the core axial direction is determined taking each core axial direction arranging distance between neutron detectors 37 of the LPRM detector assembly 34 into consideration. More specifically, if the core axial direction arranging distance between each of the neutron detectors 37 is set as L, the gamma-ray thermometer assembly (GT assembly) 35 is constructed in the following manner that axially center positions of four .gamma.-ray heating detectors of the above 8 .gamma.-ray heating detectors 44 are arranged at the same axial positions of the LPRM detector 37, axially center positions of three .gamma.-ray heating detectors of them are arranged on an intermediate position between LPRM detectors 37 at an interval of L/2, and an axially center position of one lowermost .gamma.-ray heating detector 44 of them is arranged below the lowermost LPRM neutron detector 37 in a fuel effective portion of 15 cm or more upper from a lower end of the fuel effective portion at a distance of L/4 to L/2. In the case where the .gamma.-ray heating detector 44 is located above the uppermost LPRM detector 37, the .gamma.-ray heating detector 44 is arranged above the lowermost LPRM neutron detector 37 in a fuel effective portion of 15 cm or more lower from an upper end of the fuel effective portion at a distance of L/4 to L/2. As described above, the .gamma.-ray heating detector 44 is arranged within a 15 cm or more range separating from the upper and lower ends of the fuel effective portion. The reason is as follows; more specifically, according to .gamma.-ray heating contributing range analysis, a .gamma.-ray contributing range is newly found out; for this reason, it is necessary to accurately detecting a .gamma.-ray heating value close to the upper and lower ends of the fuel effective portion. The lowermost .gamma.-ray heating detector 44 must be arranged in the fuel effective length H and in the vicinity of the lower end of the fuel effective portion as much as possible. For this reason, in the case where the fuel effective length (approximately 371 cm at present) H is divided into 24 nodes in the core axial direction, it is preferable that the axially center position of the lowermost .gamma.-ray heating detector 44 is located on the center position of the axis of an axial node which is a second from the lowermost node. By arranging the .gamma.-ray heating detector 44 as described above, a .gamma.-ray heating value on the lower end side of the core is detected by means of the lowermost .gamma.-ray heating detector 44 of the GT assembly 35. Thus, it is possible to measure a .gamma.-ray heating value over a considerably wide range in the axial direction along the fuel effective length H, and to measure a .gamma.-ray heating value on a lower end range of the core. This results from the following reasons; more specifically, the lowermost node primarily has a low output due to a neutron leakage, the lowermost .gamma.-ray heating detector 44 has a low response, and further, a contributing range of gamma-ray to the .gamma.-ray heating detector 44 is 15 cm or more. Therefore, the lowermost .gamma.-ray heating detector 44 is arranged on a position separating from the lower end of the fuel effective length at a distance of 15 cm or more, and thereby, it is possible to be equally heated upper and lower sides of the location with .gamma.-ray. Moreover, the following disadvantage should be avoided. More specifically, if the lowermost .gamma.-ray heating detector 44 is not arranged at a position separating from the lower end of the fuel effective length at a distance of 15 cm or more, other .gamma.-ray heating detectors 44 arranged in other core axial directions measure a heating effect of .gamma.-ray from the upper and lower sides in the axial direction; while the lowermost .gamma.-ray heating detector 44 detects only .gamma.-ray heating contribution from the upper side. For this reason, a balance of .gamma.-ray heating value measurement between each detector 44 is not proper, and a correlation equation of GT signal to power is also not proper. Therefore, because the lowermost .gamma.-ray heating detector 44 is arranged on a position separating from the lower end of the fuel effective length at a distance of 15 cm or more, it is possible to avoid unbalance of .gamma.-ray heating value measurement of the lowermost .gamma.-ray heating detector 44 so as to prevent the correlation equation of GT signal to power of the lowermost .gamma.-ray heating detector 44 from being different from other .gamma.-ray heating detectors 44 except for the lowermost .gamma.-ray heating detector 44. In an axial design of the latest fuel assembly 4, a natural uranium bracket is frequently used as the lowermost node. For this reason, even if the natural uranium bracket portion having a low output is measured, an output signal of the GT assembly 35 is extremely low; as a result, there is no meaning of interpolating and extrapolating a power distribution at a position below the lowermost LPRM detector 37. By the way, the gamma-ray thermometer assembly (GT assembly) 35 is constructed in combination with the fixed .gamma.-ray heating detector 44, and has a long rod-like structure as shown in FIG. 5 and FIG. 6. The gamma-ray thermometer assembly 35 is a thin and long rod-like assembly having a diameter of e.g., approximately 8 mm.o slashed., and has a length of substantially covering a fuel effective length, for example, 3.7 m (370 cm) to 4 m (400 cm) in the core axial direction. The gamma-ray thermometer assembly (GT assembly) 35 includes a cover tube 65 formed of stainless steel, which is used as a metallic jacket, and a metallic long rod-like core tube 63 is housed in the cover tube 65. Further, the cover tube 65 and the core tube 66 are fixed to each other by shrinkage fit, cooling fit or the like. A sleeve or annular space portion 67 constituting an adiabatic portion is formed between the cover tube 65 and the core tube 66. A plurality of e.g., at least four (4), more specifically, eight (8) or nine (9) annular space portions 67 are discretely arranged at equal intervals in the axial direction. The annular space portion 67 is formed by cutting an outer surface of the core tube 66 along a circumferential direction thereof. Then, a gas having a low heat conductivity, for example, an Ar (argon) gas is filled in the annular space portion 67. The annular space portion 67 may be formed on the cover tube 65 in side wall which is a jacket tube. As the gas having a low heat conductivity, an inert gas such as the Ar gas, a nitrogen gas or the like may be used. The fixed .gamma.-ray heating detector (GT detector) 44 is provided at a position where the annular space portion 67 is formed, and thus, a sensor portion of the gamma-ray thermometer assembly 35 is constructed. The core tube 66 has an internal hole 68 which extends through a center portion of the core tube 66 along an axial direction thereof. In the internal hole 68, a mineral insulated (MI) cable sensor assembly 70 is fixed by brazing, caulking (fastening) or the like. The cable sensor assembly 70 is provided at its center portion with a built-in heater 71 which functions as a rod-like exothermic member of a heater wire for calibrating the gamma-ray thermometer assembly 35, and a plurality of differential type thermocouples 72 which functions as temperature sensors, around the heater 71. A space between the built-in heater 71 and each thermocouple 72 are filled with an electric insulating layer or a metal/metal alloy filler 73, and then, are integrally housed in a metallic cladding tube 74. The metallic cladding tube 74 is closely contacted at outer peripheral surface thereof to the inner peripheral surface of the core tube 66 and at inner peripheral surface thereof to each outer peripheral surface of each thermocouple 72. The built-in heater 71 of the gamma-ray thermometer assembly 35 comprises a sheath heater, and is integrally constructed in a manner that a heater wire 75 is coated with a metallic cladding tube 77 via an electric insulating layer 76. Similarly, each thermocouple 72 is integrally constructed in a manner that thermocouple signal wire 78 are coated with a metallic cladding tube 80 via an electric insulating layer 79. In the differential type thermocouple 72 located in the internal hold 68 of the core tube 66, its low temperature point and a high temperature point are arranged so as to correspond to the annular space portion 67, and thus, the .gamma.-ray heating detector 44 which is the sensor portion of the gamma-ray thermometer assembly 35 is constructed. As shown in FIG. 6, each thermocouple 72 is set in a manner that a high temperature point 81a is located on the sensor portion formed in the annular space portion 67, that is, on the center of the adiabatic portion in the axial direction, and that a low temperature point 81b is located at a downward position slightly separating from the adiabatic portion (the low temperature point 81b may be located at an upward position slightly separating from the adiabatic portion). The thermocouples 72 are coaxially inserted around the built-in heater 71 by the same number as the .gamma.-ray heating detectors 44. The fixed .gamma.-ray heating detectors 44 constitute the gamma-ray thermometer assembly 35 for detecting an in-core power distribution detector, and the principle of measuring the in-core power distribution is shown in FIG. 7A and FIG. 7B. In a reactor such as a boiling water type reactor or the like, a .gamma.-ray is generated in proportional to a local fission rate of a nuclear fuel mounted in the reactor core 3 housed in the reactor pressure vessel 2. The generated .gamma.-ray flux heats a structural element of the gamma-ray thermometer assembly 35, for example, the core tube 66. The heat energy is proportional to a local .gamma.-ray flux; on the other hand, the .gamma.-ray flux is proportional to the fission rate close thereto. In the annular space portion 67 of each .gamma.-ray heating detector 44 which constitutes the gamma-ray thermometer assembly 35, since a performance of removing heat in radial direction by a coolant 82 is poor due to a heat resistance of the annular space portion 67, there is generated a heat flow as shown by an arrow A in FIG. 7B, which makes a detour in the axial direction so that a temperature difference is caused. So, the high temperature point 81a and the low temperature point 81b of the differential type thermocouple 72 are arranged as shown in FIG. 6 and FIG. 7B so that it is possible to detect the temperature difference by a voltage signal. The temperature difference is proportional to the .gamma.-ray heating value, making it possible to obtain a .gamma.-ray heating value which is proportional to a local fission rate from the voltage signal of the differential type thermocouple 72. This is the measuring principle of the gamma-ray thermometer. On the other hand, in the fuel assembly 4, as shown in FIG. 2 and FIG. 3, a large number of fuel rods (not shown) are housed in a rectangular or cylindrical channel box 83. Each fuel rod housed in the fuel assembly 4 is fixed in a manner that an uranium oxide sintered pellet or an uranium-plutonium mixed oxide sintered pellet is filled in a fuel cladding tube made of a zirconium alloy so that upper and lower ends of the fuel cladding tube are welded by end plugs, respectively. Large number of fuel rods are bundled so that a plurality of fuel spacers are arranged in an axial direction at predetermined intervals in order to secure a predetermined distance between the fuel rods. Moreover, an upper tie-plate and a lower tie-plate are arranged at the upper and lower end portions of the fuel assembly 4 so as to engaging with a lower structure and an upper structure of the core 3, respectively. In the fuel assembly 4 of the boiling water type reactor (BWR), the channel box 83 covers an outer side of the bundled fuel assembly 4 so as to form a coolant passage for each fuel assembly 4. The great number of fuel assemblies 4 as described above stand in the core 3 of the reactor, and the process control computer 31 executes a simulation of the in-core power distribution, the margin with respect to the operational limit value {maximum linear heat generating ratio (kW/m) and minimum critical power ratio} of core fuel, according to the power distribution simulation program module, so-called three-dimensional nuclear thermal-hydraulics simulation code. The margin with respect to the operational limit value {maximum linear heat generating ratio (kW/m), referred simply to MLHGR, and minimum critical power ratio, referred simply to MCPR} of core fuel is simulated by the process control computer 31, and then, the simulation result is displayed on the display unit 63 so that the operator monitors the simulation result. Next, monitoring process of core power distribution of the power distribution monitoring system 29 according to the present invention is explained hereinafter, and more particularly, calibration process of detection sensitivity of the in-core fixed nuclear instrumentation system 30 is explained in the central of the monitoring process. In the reactor power distribution monitoring system 29 according to the present invention, a fuel state of the core 3 and a reactor operating mode of the boiling water type reactor (BWR) are monitored by the process control computer 31. That is, various process data {control rod pattern, core coolant flow rate, reactor doom pressure, flow of feed water, a temperature of feed water (a core inlet coolant temperature) and so on} as the reactor state data measured by the core state data measuring device 55 of the boiling water type reactor, are inputted to the state data processing unit 58, and then, these data are collected and processed by the state data processing unit 58 so as to calculate a reactor thermal output or the like. The state data processing unit 58 may be constructed as a part of the process control computer 31; in this case, processing for collecting the core state data is carried out by the process control computer 31. The core state data D3 including the reactor thermal output, which is collected and calculated by the core state data processing unit 58, is transmitted to the process control computer 31 so as to be received in the CPU 60 by the interface process of the nuclear instrumentation control process module 60M1 thereof. On the other hand, neutron flux in the core 3 detected by each LPRM detector assembly 34 of each in-core nuclear instrumentation assembly 32 is converted into the LPRM data D2 via the LPRM signal processing unit 40, and then, each of the LPRM data D2 is transmitted to the process control computer 31 so as to be received in the CPU 60 by the interface process of the nuclear instrumentation control process module 60M1 thereof. Similarly, thermocouple output signal (GT signal) measured by each .gamma.-ray heating detector 44 of each in-core nuclear instrumentation assembly 32 is converted into the GT data D1 representing the .gamma.-ray heating value (W/g) per unit weight by means of the GT signal processing unit 48 on the basis of each sensitivity S.sub.0 of each .gamma.-ray heating detector 44, and then, is transmitted to the process control computer 31 so as to be received in the CPU 60 by the interface process of the nuclear instrumentation control process module 60M1 thereof. The power distribution simulation process module 60M2 of the CPU 60 executes the power distribution simulating process in accordance with the program module (three-dimensional nuclear thermal-hydraulics simulation code) PM2 stored in the memory unit 61 on the basis of the transmitted the GT data D1, the LPRM data D2 and the core state data D3, so that the core power distribution, the core neutron flux distribution, simulation values of the GT signals corresponding to the measured GT data S1, the margin with respect to the operational thermal limit value and so on are simulated. The simulated data including the core power distribution, the simulation values of the GT signals, the margin with respect to the operational thermal limit value and so on are stored in the memory unit 61 as the occasion demands. Incidentally, in this embodiment, in the memory unit 61, at least one of approximate expression data (data set) according to correlation parameters representing the correlation between the nodal power of the fuel assembly 4 and the GT data values D1 based on the GT signals S1 and interpolation and extrapolation lookup table data (data set) according to the above correlation parameters is stored, wherein the correlation parameters includes, for example, a fuel type, a node burn-up, control rod state, a historical relative water density (historical void fraction), an instantaneous relative water density (instantaneous void fraction). That is, the process module 60M2 of the CPU 60 is adapted to simulate the correlation parameters simultaneously with simulating the core power distribution, and to simulate the simulation values of the GT signals by using the at least one of the approximate expression data and the lookup table data according to the simulated correlation parameters. In addition, the process module 60M2 of the CPU 60 corrects the simulated results including the core power distribution and so on by using the actually measured core nuclear instrumentation data (GT data D1) from the core 3 and the three-dimensional nuclear thermal-hydraulics simulation code) in accordance with the program module PM3. At this time, in order to measure the power distribution in an axial direction of the core 3, each GT assembly 35 has fixed GT detectors 44 which is the same N-th (number) as the fixed LPRM detector 37 less than 24 nodes, for example, four or more, and then, the core power distribution or the like simulated by the power distribution simulation process module 60M2 of the CPU 60 is learnt so as to be corrected on the basis of the three-dimensional nuclear thermal-hydraulics simulation code and the core nuclear instrumentation data (GT data D1) corresponding to the GT signals measured by each GT detector 44 of each GT assembly 35. Incidentally, the power distribution adaptation correction process of the CPU 60 will be detailedly explained by referring to FIG. 13 and FIG. 14 in sixth embodiment of the specification. Namely, the actual thermocouple output signals (GT signals) S1 from the GT assemblies 35 are converted from the voltage signals into the GT data D1 corresponding to the gamma-ray heating values (W/g) by the GT signal processing unit 48 so as to be inputted to the process control computer 31. At this time, by the power distribution simulation process module 60M2 of the CPU 60, a .gamma.-ray heating value for each axial node of each GT assembly is obtained according to the core power distribution simulated by the power distribution simulation process module 60M2 of the CPU 60 on the basis of the three-dimensional nuclear thermal-hydraulics simulation code of the program module PM2. Each .gamma.-ray heating value is temporally stored in the memory unit 61. Related to some nodes in the axial direction at which the GT detector 44 is provided, difference between each simulation value of the part of nodes stored in the memory unit 61 and each actual measured value (GT data D1 value) thereof is obtained by ratio. Then, by the power distribution simulation process module 60M2 of the CPU 60, data indicative of differences (ratios) between the respected actual .gamma.-ray heating values (GT data D1 values) of the GT detectors 44 having the limited number in the core axial direction and the respected simulation values of the .gamma.-ray heating values corresponding to the GT detectors 44 are interpolated and extrapolated in other (remained) nodes in the axial direction, respectively, wherein the GT detector 44 is not provided at the other nodes in the axial direction, thereby obtaining the correction data of the .gamma.-ray heating value differences with respect to the whole axial nodes. Incidentally, in addition to interpolation and extrapolation in the axial direction, it is possible to interpolate and extrapolate the .gamma.-ray heating value difference corrections (correction ratios; correction factors) with respect to radial positions at which the GT assemblies are not provided along a core radial direction. Further, the power distribution simulation process module 60M2 of the CPU 60 corrects the core power distribution simulated by the power distribution simulating process so that each .gamma.-ray heating value difference correction data value for each node of each GT assembly is "1.0"; that is, the GT data D1 value of each node in the axial direction of each GT assembly and each simulation value of each .gamma.-ray heating value corresponding to each node are coincident with each other, and whereby, it is possible to obtain a high accurate reactor power distribution and a high accurate margin with respect to an operational thermal limit value, or, in addition to them, a high accurate neutron flux distribution. As described above, in the process control computer 31 for monitoring the reactor operating mode and the core power distribution, the CPU 60 always continuously receives the core state data D3, and periodically (e.g., one time per hour) or always carries out the core power distribution simulation process (three-dimensional nuclear thermal-hydraulics simulation process) on the basis of the latest operating parameters (core state data D3) and the three-dimensional nuclear thermal-hydraulics simulation code (program module PM2) in accordance with a simulation request command inputted from the input console 62 by the input operation of the operator. More specifically, in accordance with the power distribution adaptation process module PM3, according to the GT data D1 (W/g) based on the GT signals S1 at that point of time at which the power distribution is simulated, the simulated core power distribution is corrected so that the actually measured core nuclear instrumentation data (the GT data D1) are reflected the simulated power distribution, whereby it is possible to simulate a high accurate reactor power distribution and a high accurate margin with respect to the operational thermal limit value, or, in addition to them, a high accurate neutron flux distribution. Meanwhile, the nuclear instrumentation control process module 60M1 as a part of processing functions of the CPU 60 has a function of computing a reactor operating time, for example, in-core elapse time after each GT assembly 35 is loaded (mounted) in the reactor (core 3), a function of updating and storing each in-core elapse time (in-core mounted time) of each GT assembly 35 in the memory unit 61 of the process control computer 31, and a function of storing a plurality of preset heater calibration time intervals in the memory unit 61 which will be described later. Further, in a state that there is no change in the core state process S3 such as operating parameters (core power, core coolant flow rate, control rod pattern, etc.), the nuclear instrumentation control process module 60M1 has a function of transmitting an execution instruction of output voltage sensitivity measurement processing (computation processing) by the built-in heater 71 of each fixed GT detector 44 of each GT assembly 35, to the GT heater control unit 53 every predetermined time. Incidentally, the aforesaid processing (computation processing) for measuring a sensitivity of each fixed GT detector 44 by the built-in heater 71, is called as heater calibration processing, wherein the sensitivity of each fixed GT detector 44 represents a value for determining a relationship between thermocouple output voltage and .gamma.-ray heating value (unit: (W/g) per unit weigh) of each GT detector 44. In addition, the nuclear instrumentation control process module 60M1 of the CPU 60 has a function of storing a time (calibration processing start time) at the point of time of transmitting the executive instruction of output voltage measurement processing (heater calibration instruction) of each GT assembly 35 in the memory unit 61. That is, each calibration processing start time of each GT assembly 35 is stored in each different address of the memory unit 61. At this time, the nuclear instrumentation control process module 60M1 of the CPU 60 previously sets different transmission intervals of the heater calibration instruction (hereinafter, the transmission interval is referred to "heater calibration time interval") to the GT assemblies 35 according to the in-core elapse times of the GT assemblies 35. That is, the following different heater calibration time intervals are stored in the memory unit 61 of the process control computer 31. Namely, 48-hours is set as a first heater calibration time interval with respect to the GT assembly in which a reactor operating time after the GT assembly mounted in the core 3 is within 500 hours (that is, in-core elapse time within 500 hours), 168-hours is set as a second heater calibration time interval with respect to the GT assembly in which the in-core elapse time ranges from 500 to 1000 hours, 336-hours is set as a third heater calibration time interval with respect to the GT assembly in which the in-core elapse time ranges from 1000 to 2000 hours, and one month (or 1000 hours) is set as a fourth heater calibration time interval with respect to the GT assembly in which the in-core elapse time exceeds 2000 hours. By referring to the memory unit 61, as shown in FIG. 8, the nuclear instrumentation control process module 60M1 calculates the in-core elapsed times of the GT assemblies and discriminates and selects a GT assembly or GT assemblies as a heater calibration target from all GT assemblies on the basis of each elapse time from the previous calibration processing time of each GT assembly 35 to the present time and a heater calibration time interval corresponding to the present in-core elapse time of each GT assembly 35 (step S1). Next, the nuclear instrumentation control process module 60M1 of the CPU 60 registers each heater calibration time interval corresponding to each discriminated in-core elapse time of each GT assembly 35 in the memory at each of the discrimination processes (step S2). Then, the nuclear instrumentation control process module 60M1 transmits each registered heater calibration time interval corresponding to each in-core elapse time corresponding to each GT assembly 35 to the display unit 63 so as to display each heater calibration time interval of each GT assembly 35 as a heater calibration time interval registration image on the display unit 63 (step S3). Furthermore, the nuclear instrumentation control process module 60M1 automatically transmits a heater calibration processing execution instruction {including an address (positional address) of the selected heater calibration target of the GT assembly 35} with respect to the selected heater calibration target of the GT assembly 35, to the GT heater control unit 53 and the GT signal processing unit 48, or transmits a GT calibration instruction command transmission request with respect to the heater calibration target of the GT assembly 35, to the display unit 63 so as to display and output the GT calibration instruction command transmission request with respect to the operator via the display unit 63 (step S4). At this time, the operator operates the input console 62 in accordance with the GT calibration instruction command transmission request displayed on the display unit 63 so as to transmit a GT calibration instruction command corresponding to the aforesaid heater calibration target of the GT assembly 35. In response to the transmitted GT calibration instruction command, the nuclear instrumentation control process module 60M1 automatically transmits the aforesaid heater calibration processing execution instruction to the GT heater control unit 53 and the GT signal processing unit 48. At this time, the GT heater control unit 53 starts to supply a power (applies a voltage) to the built-in heater 71 of at least one of the target GT assembly 35 corresponding to the positional address of the transmitted heater calibration processing execution instruction with delay time, and controls the applied heater voltage so that a current value flowing through the built-in heater 71 becomes a predetermined value. Next, the GT heater control unit 53 measures the applied voltage value of the built-in heater 71 of the at least one of the target GT assembly 35 and the current value flowing through the built-in heater 71 thereof so as to transmit the measured values to the GT signal processing unit 48. On the other hand, the CPU 48A of the GT signal processing unit 48, as shown in FIG. 9, receives the heater calibration processing execution instruction (positional address) transmitted from the nuclear instrumentation control process module 60M1 (step S10), and then, in response to a receiving timing of the heater calibration processing execution instruction, concurrently measures a thermocouple output voltage signal (mV) of each GT detector 44 of the GT assembly 35, in a no-heated state, having the positional address along the core axial direction (step S11). Incidentally, the thermocouple output voltage signal of each GT detector 44 which is not heated by the built-in heater 71 is referred to "no-heated output voltage signal" hereinafter. Furthermore, simultaneously to Step S11, the CPU 48A of the GT signal processing unit 48 receives the applied voltage to the heater 71 and the measured current value in the heater 71 transmitted from the GT heater control unit 53 (step S12), and then, in accordance with the receiving timing of the applied voltage and the measured current value, concurrently measures a thermocouple output voltage signal (mV) of each GT detector 44 of the GT assembly 35 having the positional address with each GT detector 44 heating by the heater 71 (step S13). Incidentally, the thermocouple output voltage signal of each GT detector 44 which is heated by the built-in heater 71 is referred to "heated output voltage signal" hereinafter. Then, the CPU 48A of the GT signal processing unit 48 stores the measured no-heated output voltage signal and the measured heated voltage signal of each GT detector 44, and voltage and current of heater power supply of the heater calibration target of the GT assembly 35 with the positional address in each separated address of the memory unit 48B (step S14). The above heater calibration process according to the heater calibration process execution instruction is continued while changing the positional addresses of the target GT assemblies until the heater calibration process to all of the target GT assemblies which have need of the heater calibration is finished. Next, the CPU 48A of the GT signal processing unit 48 measures (computes) a sensitivity S.sub.0 {a value for determining a relationship between a thermocouple output voltage and a .gamma.-ray heating value (unit: W/g) per unit weight of each GT detector 44} of each present heater calibration target of each GT detector 44 on the basis of the no-heated output voltage signal and the heated output voltage signal of heater power of each present heater calibration target of each GT detector 44 (step S15). The following is a description about the sensitivity measurement process by the CPU 48A of the GT signal processing unit 48 with reference to FIG. 10. The following equation (1) is established as an expression of relation between a thermocouple output voltage of the GT detector 44 and a gamma-ray heating value W.gamma. per unit weight of the GT detector 44. EQU U.gamma.=S.sub.0 (1+.alpha.U.gamma.)W.gamma. (1) where, S.sub.0 : output voltage sensitivity (mV/(W/g)) PA1 .alpha.: non-linear coefficient (mV.sup.-1) PA1 U.gamma.: output signal (mV) PA1 W.gamma.: gamma heating value (W/g) PA1 U: no-heated output voltage (mV) PA1 U': heated output voltage (mV) PA1 P.sub.H : additional heating value (W/g) by built-in heater PA1 where, the GAF.sup.L.sub.(n-1) represents (n-1) time integration gain adjustment factor and the GAF.sup.L.sub.n represents n time integration gain adjustment factor. As described above, by multiplying the last time GAF.sup.L.sub.(n-1) by the RGAF.sup.L, it is possible to determine the gain adjustment factor (the GAF.sup.L.sub.n) by which the LPRM signal S2 is multiplied so as to obtain the LPRM data D2. Incidentally, the integration gain adjustment factor the GAF.sup.L.sub.n by which the LPRM signal S2 is multiplied is referred to "first gain adjustment factor" and the above gain adjustment factor RGAF.sup.L is referred to "second gain adjustment factor". PA1 Where, "a" and "b" are constants to be approximated, and "t" is an elapse time (second or minute). In addition, ".lambda." is a nuclear time constant (see Table 1) set each node according to a time constant library previously stored in the memory unit 61 of the process control computer 31. In this case, the above non-linear coefficient .alpha. is a fixed value computed taking a temperature dependency of a physical property value of a structural material of the GT detector 44 into consideration. Moreover, the output voltage sensitivity S.sub.0 of the present GT detector 44 is computed by the following equation (2) with the use of the measured no-heated output voltage signal and the heated output voltage signal. EQU S.sub.0 =[{U'/(1+.alpha.U')}-{U/(1+.alpha.U)}]/P.sub.H (2) where, More specifically, in a state that the operator confirms that a state of the core 3 is constant and stable, when adding a heating value P.sub.H by means of the built-in heater 71 shown in FIG. 5 and FIG. 6, a change of thermocouple output signal (difference between U and U') is caused in accordance with the added heating value P.sub.H. Thus, the sensitivity S.sub.0 of the GT detector 44 can be computed with the use of a previously measured mass (weight) of the GT detector 44, a heater resistance value thereof and the above equation (2). In this case, preferably, the built-in heater 71 of the GT assembly 35 is manufactured so that the resistance value of the built-in heater 71 is constant in the axial direction independent of each GT detector. However, because of considering a manufacture error for each GT assembly 35 or a manufacture error with respect to an axial distribution of each GT assembly 35, its manufacturing data is reflected so that the additional heating value P.sub.H of each of the aforesaid built-in heaters 71 is determined on the basis of the supply current and the resistance value of the detecting section of each of the built-in heaters 71. By the way, a gamma-ray heating value W.gamma. per unit weight of the GT detector 44 is computed from the aforesaid output voltage sensitivity S.sub.0 of the GT detector 44 and the output voltage signal (mV signal) of the GT detector 44 with the use of the following equation (3). EQU W.gamma.=U.gamma./{S.sub.0 (1+.alpha.U.gamma.)} (3) As described above, the CPU 48A of the GT signal processing unit 48 can compute the sensitivities S.sub.0 of all GT detectors 44 of the heater calibration target GT assemblies 35 on the basis of controlling the built-in heaters 71 of the heater calibration target GT assemblies 35 by the GT heater control unit 53. Then, the CPU 48A of the GT signal processing unit 48 stores the computed sensitivity S.sub.0 of each GT detector 44 of the heater calibration target in the memory unit 48B, and then, transmits the computed sensitivity S.sub.0 of each GT detector 44 of the heater calibration target to the nuclear instrumentation control process module 60M1 of the process control computer 31 (step S16). As described above, the sensitivity measuring process by the GT heater control unit 53 and the GT signal processing unit 48 according to the heater calibration process execution instruction, is repeatedly carried out in a predetermined sequence until all GT assemblies 35 of the heater calibration target are processed. Moreover, the sensitivity measuring process of each GT detector 44 of each GT assembly 35 of the heater calibration target is repeatedly carried out in accordance with the heater calibration processing execution instruction transmitted by the discrimination process of the nuclear instrumentation control process module 60M1 at each corresponding heater calibration time interval. As is evident from the above description, the sensitivity S.sub.0 of each GT detector 44 of each GT assembly 35 is periodically measured (computed) by the GT signal processing unit 48 in accordance with each heater calibration interval determined based on each in-core elapse time of each GT assembly 35, and then, the measured sensitivity S.sub.0 of each GT detector 44 of each GT assembly 35 is transmitted to the nuclear instrumentation control process module 60M1. At that time, the nuclear instrumentation control process module 60M1 stores the sensitivity S.sub.0 of each GT detector 44 of each GT assembly 35 periodically transmitted in accordance with each heater calibration interval, that is, time sequentially change data of the sensitivity S.sub.0 of each GT detector 44 of each GT assembly 35 in the memory unit 61 for each GT detector 44. Then, the nuclear instrumentation control process module 60M1 executes a process of making a comparison between a sensitivity S.sub.0 ' of each GT detector 44 newly transmitted to the nuclear instrumentation control process module 60M1 and past several sensitivity S.sub.0 data (sensitivity change data) before the point of time of the sensitivity S.sub.0 ' transmission, and displaying the comparative result via the display unit 63, or a process of displaying a elapsed change trend (graph) of sensitivity including the sensitivity S.sub.0 ' on the basis of the time sequentially change data of the sensitivity stored in the memory unit 61 via the display unit 63. To give an example of displaying process of elapse time change trend, with the use of all sensitivity S.sub.0 data of the GT detector 44 after the GT detector 44 is mounted in the core 3, or past several sensitivity S.sub.0 data from the present time, according to least square approximation, a function of computing the following equation (4) for each GT detector 44 may be incorporated into the nuclear instrumentation control process module 60M1. EQU S.sub.0 =a+b.multidot.e.sup.-.lambda.t (4) In this case, .lambda. of the above equation (4) can be determined on the basis of the sensitivity S.sub.0 data according to the least square approximation, and also, can be used as a representative value of past actual data value. Now, referring to FIG. 11, there is shown an example of an elapse time change trend graph of the sensitivity S.sub.0 of each GT detector 44 of each GT assembly 35, which is dependent on in-core elapsed (mounted) time based on the above equation (4). In the graph, a symbol X denotes an actually measured GT detector sensitivity S.sub.0, and a symbol Y denotes an approximation curve represented by a predictive approximate equation of the equation (4) obtained by using the actually measured sensitivity S.sub.0. The operator monitors the comparative result between a newly transmitted sensitivity S.sub.0 ' and sensitivity change data, or the elapse time change trend graph (see FIG. 11), which is displayed on the display unit 63. In the monitoring result, in the case where the operator judges that the newly transmitted sensitivity S.sub.0 ' changes not less than a predetermined value {first judgement value (abnormal judgement value; allowable sensitivity change judgement value, for example 10% of the sensitivity S.sub.0)}, the operator judges the transmitted sensitivity S.sub.0 ' to be wrong so as to transmit a bypass instruction to the process control computer 31 via the input console 62. Moreover, for example, in the case where any sensitivities of plural GT detectors 44 of plural GT assemblies 35 change within a predictable range so that any sensitivities of them are normal, the operator is adapted to transmit a sensitivity update instruction for collectively updating the corresponding sensitivities of plural GT detectors 44 of plural GT assemblies 35, and further, is adapted to transmit a sensitivity update instruction for individually updating the corresponding sensitivity for each GT detector 44 and for each GT assembly 35. The nuclear instrumentation control process module 60M1 of the CPU 60 of the process control computer 31 transmits a sensitivity update instruction for updating the sensitivity S.sub.0 of the GT detector 44 of the GT assembly 35 corresponding to the transmitted sensitivity update instruction from the input console 62 into a new sensitivity S.sub.0 ', to the GT signal processing unit 48. The GT signal processing unit 48 updates the sensitivity S.sub.0 of GT detector 44 of the corresponding GT assembly 35 on the basis of the transmitted sensitivity update instruction into a new sensitivity S.sub.0 ', and then, converts the output voltage signal from the GT detector 44 into the GT data D1 with the use of the updated new sensitivity S.sub.0 '. In addition, the nuclear instrumentation control process module 60M1 of the process control computer 31, according to a second judgement value (second allowable sensitivity change judgement value) which is previously stored in the memory unit 61 and is within the predictable range and less than the first judgement value, for example 0.2% of the sensitivity S.sub.0 and the comparative result between the obtained sensitivity S.sub.0 ' and the sensitivity change data, or the sensitivity elapse time change trend graph, automatically makes a decision whether or not the newly transmitted sensitivity S.sub.0 ' changes more than the second judgement value and less than the first judgement. Then, in the case where it is judged that the newly transmitted sensitivity S.sub.0 ' changes more than the second judgement value and less than the first judgement value by the nuclear instrumentation control process module 60M1, the nuclear instrumentation control process module 60M1 of the CPU 60 of the process control computer 31 judges that the change of the newly sensitivity S.sub.0 ' is a normally and preferable change so as to transmit a sensitivity update instruction for updating the sensitivity S.sub.0 of the GT detector 44 of the corresponding GT assembly 35 into a new sensitivity S.sub.0 ', to the GT signal processing unit 48. In addition, in the case where it is judged that the newly transmitted sensitivity S.sub.0 ' changes less than the second judgement value by the nuclear instrumentation control process module 60M1, the nuclear instrumentation control process module 60M1 judges that it is not necessary to update the sensitivity S.sub.0 of the GT detector 44 of the corresponding GT assembly 35 into a new sensitivity S.sub.0 ' so as not to execute the aforesaid sensitivity update process. Furthermore, in the case where the newly transmitted sensitivity S.sub.0 ' changes not less that the first judgement value (e.g., 10% of the sensitivity S.sub.0), the nuclear instrumentation control process module 60M1 judges the GT detector 44 of the corresponding to the GT assembly 35 to be wrong so as to output warning information including the address of the GT detector 44 and the GT assembly 35 which are wrong to the operator directly or via the display unit 63. As a result, according to the outputted warning information, the operator determines that the GT detector 44 corresponding to the warning information or the GT assembly 35 including the GT detector 44 corresponding to the warning information is wrong, thereby registering the GT detector 44 or the GT assembly 35 as a failure bypass. In addition, the nuclear instrumentation control process module 60M1 of the CPU 60 of the process control computer 31 periodically evaluates the in-core elapse time of each GT assembly 35 successively updated and registered in the memory unit 61. In accordance with the elapse time of the in-core elapse time of each GT assembly 35, in the case where the GT heater calibration time interval set at the present point of time (within the in-core elapse time) is changed over into the next heater calibration time interval set in the next in-core elapse time, the nuclear instrumentation control process module 60M1 changes a display mode of display symbol of the corresponding GT assembly 35 (e.g., flushing the display symbol, etc.) on the heater calibration time interval registration image on the display unit 63 thereby informing a change-over of the above heater calibration time interval to the operator. As described above, according to this first embodiment, in accordance with each heater calibration interval set according to each in-core elapse time of each GT assembly 35 (GT detector 44), the sensitivity of each GT detector 44 is measured by the operations of the nuclear instrumentation control process module 60M1, the GT signal processing unit 48, the GT heater control system 53 and the built-in heater 71, and the sensitivity of each GT detector 44 is updated according to the time change data of the measured sensitivity S.sub.0, whereby it is possible to correct a drop-down or saturation of the output voltage of the GT detector 44 with respect to the gamma-ray heating value and to obtain an very accurate .gamma.-ray heating value (GT data D1). Therefore, it is possible to further improve an accuracy of the process for correcting the core power distribution, thereby obtaining the core power distribution having a high reliability. In particular, in an operation of a reactor, the sensitivity S.sub.0 of each GT detector 44 of each GT assembly 35 slowly changes accompanying with an elapse time as shown by a broken line Y of FIG. 11, and then, becomes an equilibrium state. For this reason, in the nuclear instrumentation assembly 32 having a shorter in-core elapse time, which is mounted in the core 3 of the reactor and is started to be operated, the sensitivity S.sub.0 thereof rapidly changes during its operation cycle. On the other hand, in the nuclear instrumentation assembly 32 which has been mounted from the previous operation cycle or before that, that is, having a longer in-core elapse time, the sensitivity S.sub.0 of the GT detector 44 is almost stable. Therefore, in the nuclear instrumentation assembly 32, in which the in-core elapse time is long and the sensitivity S.sub.0 of the GT detector 44 of the GT assembly 35 is stable, a confirmation may be made whether or not the GT assembly 35 is wrong. Thus, since, for example, the heater calibration interval than 1000 hours or more as described above is set to the GT assembly 35 with stable sensitivity, it is possible to prevent an unnecessary in-ope state (bypass state) of the GT assembly from becoming long without carrying out unnecessary GT heater calibration process. Further, only GT assembly 35 having a relatively shorter in-core elapse time is calibrated by the built in heater at a relatively short heater calibration interval (e.g., 48 hours), thereby carrying out an effectively short-time GT heater calibration with respect to the GT heater. In addition, it is possible to make low a provability of heater breakdown in an operation life of the GT assembly 35. Moreover, the heater calibration time interval of each GT assembly (each GT detector) is displayed on the heater calibration time interval registration image of the display unit 63; therefore, it is easy to make a heater calibration frequency management of the GT assembly 35, and it is possible to readily identify the presence of a short-time calibration target of the GT assembly 35. In this first embodiment, the computation of the sensitivity S.sub.0 of the GT detector 44 is carried out by means of the GT signal processing unit 48. However, the present invention is not limited to the structure. That is, the output voltage signal (mV) of the GT detector 44 is directly transmitted to the nuclear instrumentation control process module 60M1 of the process control computer 31, and then, the computation of the sensitivity S.sub.0 of the GT detector 44 may be carried out by means of the nuclear instrumentation control process module 60M1. Namely, it is not substantial problem whether the computation of the sensitivity S.sub.0 of the GT detector 44 is carried out by the CPU of which computer. Moreover, in the above first embodiment, the sensitivity calibration time interval of the GT assembly 35 stored in the memory unit 61 of the process control computer 31 has four stages in accordance with the in-core elapse time history; however, the present invention is not limited to the structure. Namely the sensitivity calibration time interval may have plural stages, for example, three stages, or two stages. As seen from a degree of sensitivity change by individual in-core elapse times of the actual GT signal detectors, for example, referring to a sensitivity change graph as shown in FIG. 11, the sensitivity calibration time interval has characteristics such that the change is great in the initial mounted state, and is saturated accompanying with an elapse time. For this reason, in view of this time interval characteristics, for example, in the case where a third judgement value representing a gap between the actually measured sensitivity of each GT detector and prediction sensitivity thereof in heater calibration state, which is predicted from the time sequential data of the sensitivity, or a change value from the sensitivity result of each GT detector measured while the preceding sensitivity update process is executed, is set to a relatively small value (e.g., 1% of sensitivity), it is preferable that the sensitivity calibration time interval has multi stages such as four states described above. Incidentally, in this embodiment, as a parameter for setting the calibration time interval of each GT detector, the in-core elapse time is used. However, the present invention is not limited to the structure. That is, an in-core neutron irradiation quantity of the sensor portion of each GT detector may be employed in place of the above in-core elapse time. In this case, the in-core neutron irradiation quantity for each GT detector 44 is accurately computed (simulated) by means of the power distribution simulation module 60M2 of the CPU 60. Furthermore, the simulated in-core neutron irradiation quantity of each GT detector 44 is stored in the memory unit 61 of the process control computer 31 in place of the in-core elapse time so that each heater calibration time interval of each GT detector is set in accordance with a range of 3 each in-core neutron irradiation quantity of each GT detector so as to store each heater calibration time interval in the memory unit 61. In addition, there is no need of accurately computing (simulating) the above in-core neutron irradiation quantity, and a parameter substantially proportional to the in-core neutron irradiation quantity may be replaced. For example, an average burn-up increment of each fuel nodes surrounding the GT detector 44 can be employed as the parameter. According to this first embodiment, in the in-core fixed nuclear instrumentation system 30 of the reactor, the plurality of LPRM detectors 37 for detecting the local power distribution of power range in the reactor core 3 and the fixed GT detectors 44 of the gamma-ray thermometer assembly 35 for detecting the .gamma.-ray heating value, are housed in the nuclear instrumentation tube 33. Further, the in-core fixed nuclear instrumentation system 30 includes: the in-core nuclear instrumentation assembly 32 in which the GT detectors 44 are arranged at least in the vicinity of the LPRM detectors 37; the LPRM signal processing unit 40 for processing the LPRM signal S2 from the LPRM detector 37; the GT signal processing unit 48 for processing the output voltage signal (GT signal) S1 from the GT assembly 35; and the GT heater control unit 53 for carrying out the electrically energizing control with respect to the built-in heater 71 built in the GT assembly 35. On the other hand, the in-core fixed nuclear instrumentation system 30 is monitored and controlled by the nuclear instrumentation control process module 60M1 of the CPU 60 of the process control computer 31. The nuclear instrumentation control process module 60M1 constitutes a monitor and control module, and controls the operations of the GT heater control unit 53 and the GT signal processing unit 48. The GT heater control unit 53 carries out the electrically energizing control with respect to the built-in heater 71 of the GT assembly 35 so that the output voltage sensitivity of each GT detector 44 can be calibrated by heater heating. Moreover, in the GT assembly 35, during reactor operation, the built-in heater 71 is electrically energized and heated, and then, the GT signal processing unit 48 measures an increase of output voltage of each GT detector 44 by heater heating (additional heating value) and the heating voltage and current to the built-in heater 71, and further, calibrates the thermocouple output voltage sensitivity of the GT detector 44 per unit heating value (W/g) by gamma-ray with the use of a previously measured heater resistance and the mass of the fixed GT detector 44. The interval of calibrating the output voltage sensitivity is set in accordance with the in-core elapse time of the gamma-ray thermometer assembly 35 as the parameter. The in-core elapse time of the gamma-ray thermometer assembly 35 is computed and so as to be stored in the memory unit 61 by the nuclear instrumentation control process module 60M1. In accordance with each in-core elapse time of each GT assembly 35, the following heater calibration time intervals are previously prepared; for example, a first time interval, a second time interval, a third time interval . . . in the order of the shorter time interval. Then, the corresponding time interval is automatically selected in accordance with the computed in-core elapse time, and the output voltage sensitivity measurement by heater heating is carried out at the selected time interval. The nuclear instrumentation control process module 60M1 outputs the warning information to the display unit 63 at the point of time when the heater calibration time interval of the output voltage sensitivity by heater heating of the fixed GT detector 44 of the GT assembly 35 is changed over from the previous time interval into a new time interval, and thus, the change-over of the calibration time interval is informed of the operator by the warning information. [Second embodiment] The following is a description on a second embodiment of an in-core fixed nuclear instrumentation system, a power distribution simulating system and a power distribution monitoring system according to the present invention. This second embodiment has the same construction and operation of the reactor power distribution monitoring system 29, the in-core fixed nuclear instrumentation system 30 and the process control computer 31, as the first embodiment (see FIG. 1 to FIG. 11) of the present invention. Therefore, like reference numerals are used to designate the same elements as the first embodiment, and the details are omitted. In the in-core fixed nuclear instrumentation system 30 and the power distribution simulating system 31 of this second embodiment, discrimination processing of the gamma-ray thermometer assembly 35 for calibrating an output voltage sensitivity by heater heating of the fixed GT detector 44 is basically different from that shown in the first embodiment of the present invention. That is, the nuclear instrumentation control process module 60M1 as a part of processing functions of the CPU 60 has a function of computing a reactor operating time (hereinafter, defined as in-core elapse time) after each GT assembly 35 is mounted in the core 3 and a function of updating and storing each in-core elapse time of each GT assembly 35 in the memory unit 61 of the process control computer 31. Similarly to the first embodiment, a plurality of heater calibration time intervals {first heater calibration time interval (for example, 48 hours), second heater calibration time interval (for example, 165 hours), third heater calibration time interval (for example, 336 hours) and fourth heater calibration time interval (for example, 1000 hours)} are previously stored in the memory unit 61 so as to correspond to the in-core elapse times. Moreover, the nuclear instrumentation control process module 60M1 has a function of storing an elapse time change data of the output voltage sensitivity S.sub.0 of each GT detector 44 (time sequential data of the output voltage sensitivity) in the memory unit 61 of the process control computer 31 at each sensitivity calibration process of the GT assembly 35, which is carried out at each time interval selected in accordance with the in-core elapse time of the GT assembly 35 so that the time sequential data of the output voltage sensitivity correspond to the in-core elapse times. That is, in the memory unit 61, each time sequential data of each output voltage sensitivity of each GT detector 44 is stored as a table with respect to each in-core elapse time. The nuclear instrumentation control process module 60M1 captures at least two or more latest time series data points from the present point of time with the use of the time series data of the sensitivity S.sub.0 of the GT detector 44, and then, estimates a curve of an output voltage sensitivity change by a linear extrapolation or quadratic curve extrapolation on the basis of the time series data point, or estimates a future output voltage sensitivity change curve by approximating the data points to a linear approximate equation, a quadratic curve approximate equation or the least square approximation equation of a curve represented by the above equation (4) "a+b.multidot.e.sup.-.lambda.t ". In the case of estimating the output voltage sensitivity change curve with the use of the above Equation (4) "a+b.multidot.e.sup.-.lambda.t ", "t" is an in-core elapse time, "a", "b" and ".lambda." are constant to be approximated. In this case, the ".lambda." may take a value representatively selected from the past sensitivity characteristic of the GT detector). Then, the nuclear instrumentation control process module 60M1, as shown in FIG. 12, estimates a future value of the sensitivity S.sub.0 of the GT detector 44 according to the above output voltage sensitivity change curve so as to calculate a value of the sensitivity S.sub.0 after a predetermined time from the last heater heating calibration, that is, after a time interval (e.g., the second time interval of 168 hours) which is longer than a time interval to be selected next to the present selected time interval (e.g., the first time interval of 48 hours) (step S20). In the case where the estimated value of the sensitivity S.sub.0 changes to exceed the predetermined value, that is, the third judgement value, for example, in the case where it is estimated that the third judgement value is 1% of the sensitivity and the estimated value of the output voltage sensitivity S.sub.0 changes more than the 1% from the sensitivity output voltage sensitivity measured at the point of the last heater heating calibration time (the judgement of step S21 is YES), a predetermined short time interval shorter than the next time interval (e.g., in the case where the present time interval is 48 hours, the predetermined short time interval is first time interval of 48 hours), is selected from the memory unit 61, and the selected time interval is set as the heater calibration time interval for the next heater calibration so as to be registered in the memory unit 61 (step S22). On the other hand, in the case where it is estimated that the estimated value of the output voltage sensitivity S.sub.0 changes not to exceed the third judgement value, that is, it is estimated that the estimated value of the output voltage sensitivity S.sub.0 changes not more than the 1% from the sensitivity output voltage sensitivity measured at the point of the last heater heating calibration time (the judgement of step S21 is NO), a predetermined next time interval (e.g., the second time interval of 168 hours) longer than the present time interval (e.g., the first time interval of 48 hours) is selected from the memory unit 61, and the selected time interval is set as the heater calibration time interval for the next heater calibration so as to be registered in the memory unit 61 (step S23). In this case, the nuclear instrumentation control process module 60M1 can automatically computes the heater calibration time interval with a predetermined limit range from the shortest heater calibration time interval (for example, first time interval of 48 hours) to the longest heater calibration time interval (for example, fourth time interval of 1000 hours) so that the sensitivity change is limited within a predetermined sensitivity range (within a range of the above third judgement value) in accordance with the change quantity of the estimated value of the output voltage sensitivity S.sub.0 by the aforesaid estimating process of the nuclear instrumentation control process module 60M1. According to the registration values of the heater calibration time intervals of the axial GT detectors 44 of one GT assembly 35 (excluding the assembly which is synthetically determined as being failure and bypass-registered in the nuclear instrumentation control process module 60M1), the nuclear instrumentation control process module 60M1 refers to the memory unit 61 so as to research the registered heater calibration time intervals of all axial GT detectors 44 of the GT assembly 35. Furthermore, the nuclear instrumentation control process module 60M1 transmits a heater calibration instruction of each GT detector 44 to the GT signal processing unit 48 and the GT heater control unit 53 on the basis of the heater calibration time interval having the shortest heater calibration time interval of all GT detectors 44 along the axial direction of the GT assembly 35. The aforesaid function and heater calibrating method are necessary in the case of using the neutron irradiation quantity of the GT detector 44 as the parameter for heater calibration in place of the in-core elapse time of the GT assembly 35. According to this second embodiment, only shipment (shipping) data value of an out-pile test or only initialization data is registered as an output voltage sensitivity S.sub.0 of the GT detector 44 of the GT assembly 35 newly loaded in a regular inspection construction before an operation cycle. For this reason, the initial sensitivity S.sub.0 of the GT detector 44 is detected, and then, the point of time when the core is a steady state is selected in a temporarily turbine operation at start-up, at the point of time of partial output after turbine operation, at the point of early time after rate output, and thus, the GT heater heating calibration processing is carried out three or four times by means of the operations of the aforesaid nuclear instrumentation control process module 60M1, the GT signal processing unit 48 and the GT heater control unit 53. For example, by four-time or more heater calibration processing, concerning the output voltage sensitivity S.sub.0 data of the GT detector 44 of the GT assembly 35 having four or more sensitivity S.sub.0 data of the GT detector 44, a sensitivity S.sub.0 value of the GT detector 44 after the predetermined time interval selected at present (here, is temporarily set as a second predetermined time interval) and a sensitivity S.sub.0 value of the GT detector 44 after the predetermined time interval longer than the second predetermined time interval (here, is temporarily set as a third predetermined time interval), are estimated by a linear extrapolation or a quadratic curve extrapolation, or are estimated by using a liner approximation, quadratic curve approximation or the "a+b.multidot.e.sup.-.lambda.t " curve approximation with the use of at least two or more latest time series data points from the present point of time of the sensitivity S.sub.0 of the GT detector 44. As a result of these estimations, in the case where it is estimated that the estimated change quantity of the sensitivity S.sub.0 of the GT detector 44 after the second predetermined time interval exceeds the third judgement value 1%, the first calibration time interval shorter than the second predetermined time interval is registered as the next heater calibration time interval in the memory unit 61. On the other hand, in the case where it is estimated that the estimated change quantity of the sensitivity S.sub.0 of the GT detector 44 after the third predetermined time interval does not exceed the third judgement value 1%, the third predetermined time interval longer than the second predetermined time interval is registered as the next heater calibration time interval in the memory unit 61. Moreover, in the case where it is estimated that although the estimated change quantity of the sensitivity S.sub.0 of the GT detector 44 after the second predetermined time interval does not exceed the third judgement value 1%, the estimated change quantity of the sensitivity S.sub.0 of the GT detector 44 after the third predetermined time interval exceeds the third judgement value 1%, the second predetermined time interval is registered as the next heater calibration time interval in the memory unit 61. Thereafter, automatic research process of each heater calibration time interval registered in the memory unit 61 is carried out in the GT detectors 44 discretely arranged in the axial direction of the identical GT assembly 35. Even in the event that in GT detector of the GT detectors 44 which are not bypass-registered, there is only one of GT heater calibration time interval having a short time interval, the heater calibration time interval of all GT detectors 44 of the corresponding GT assembly 35 is automatically corrected and registered as the aforesaid short heater calibration time interval in the memory unit 61. In the finally registered GT calibration time interval, in the case where the time interval is changed over from the previous heater calibration time interval to a new heater calibration time interval, concerning the GT assembly 35, the nuclear instrumentation control process module 60M1 has a function of flushing and displaying the corresponding heater calibration time interval on the heater calibration time interval registration image on the display unit 63 so as to give a caution to the operator. Then, the operator checks the flushing and displaying heater calibration time interval data referring to the trend graph of the output voltage sensitivity S.sub.0 of the GT detector 44 displayed on the display unit 63, and thus, can compare with the previous output voltage sensitivity S.sub.0. In the aforesaid manner, the final registration result of the GT heater calibration time interval registered in the memory unit 61 is processed by means of the nuclear instrumentation control process module 60M1, and then, is displayed on the display unit 63 so as to be informed of the operator. That is, when reaching a predetermined heater calibration time corresponding to the heater calibration time interval of the predetermined GT assembly 35, the nuclear instrumentation control process module 60M1 gives a warning for carrying out the GT heater calibration process to the operator via the display unit 61, and then, displays the target GT assembly 35 on the display unit 61 on the basis of the registration result. In this case, although a GT heater calibration timing is different every GT assembly, heater calibration of all GT detectors 44 is necessarily carried out before adjustment process of the LPRM detector sensitivity by the GT signal is carried out in a cycle start-up (low power) or in the vicinity of a rated reactor output after start-up. Whereby it is possible to unify a timing base point of the heater calibration time interval of each GT assembly 35. The GT heater heating calibration processing is carried out by the instruction from the process control computer 31 according to the same procedures as the first embodiment, and then, a newly obtained output voltage sensitivity S.sub.0 of the GT detector 44 is transmitted to the process control computer 31. The operator's procedures after that is the same as the first embodiment. This second embodiment has described the case of estimating the output voltage sensitivity S.sub.0 of the GT detector 44 by extrapolating from at least two points of the output voltage sensitivity S.sub.0 of the latest GT detector 44, e.g., three points, by using a linear equation approximation, a quadratic equation approximation or "a+b.multidot.e.sup.-.lambda.t " equation approximation. The present invention is not limited to the structure, but the output voltage sensitivity S.sub.0 of the GT detector 44 may be estimated by extrapolating from four points of the output voltage sensitivity S.sub.0 of the latest GT detector 44 by using a cubic equation approximation or a least square approximation of other function. As described above, according to this second embodiment, the sensitivity S.sub.0 changes of the GT detector 44 in predetermined future time points are estimated from the latest S.sub.0 trend data with respect to the present sensitivity S.sub.0 of the GT detector 44. Then, in the case where the estimated change value for the present selected interval is larger than the third judgement value, the short time interval is registered as the GT heater calibration time interval; therefore, it is possible to flexibly cope with an unexpected output voltage sensitivity change of the GT assembly 35, and to prevent a power distribution accuracy by the GT assembly 35 from being deteriorated. Moreover, in the case where the GT heater calibration time interval is changed over from the previous time interval into a new time interval, the above change-over is informed via the display unit 63. Thus, it is possible to give a warning of inspection and caution for any failures to the operator, and to improve an efficiency of maintaining a reliability of the in-core instrumentation of the reactor. In addition, it is possible to automatically calculate the heater calibration time interval within a predetermined range from the minimum heater calibration time interval to the maximum heater calibration time interval so that the sensitivity change is limited within a predetermined sensitivity change (within a range of the third judgement value) in accordance with the estimated sensitivity change. According to this second embodiment, in the in-core fixed nuclear instrumentation system 30 of the reactor, the plurality of fixed neutron detectors (LPRM detectors) 37 for detecting the local power distribution of power range in the reactor core and fixed GT detectors 44 gamma-ray thermometer assembly 35 for detecting the .gamma.-ray heating value, are housed in the nuclear instrumentation tube 33. Further, the in-core fixed nuclear instrumentation system 30 includes: the in-core nuclear instrumentation assembly 32 in which the GT detectors 44 are arranged at least in the vicinity of the fixed LPRM detectors 37; the LPRM signal processing unit 40 for processing the LPRM signal S2 from the LPRM detector 37; the GT signal processing unit 48 for processing the output voltage signal (GT signal) S1 from the gamma-ray thermometer assembly 35; the GT heater control unit 53 for carrying out an electrically energizing control with respect to the heaters 71 built in the GT assembly 35; and the nuclear instrumentation control process module 60M1 for simulating (computing) and storing the in-core elapse time or in-core irradiation quantity (burn-up quantity) of the GT assembly 35. The nuclear instrumentation control process module 60M1 controls the operations of the GT heater control unit 53 and the GT signal processing unit 48. During a reactor operation, when the heater wire of the built-in heater 71 is electrically energized by means of the GT heater control unit 53, the thermocouple output voltage increasing sensitivity of the fixed GT detector 44 of the GT assembly 35 with respect to heater heating (additional heating value), is measured by means of the GT signal processing unit 48 on the basis of a heating voltage and current of the built-in heater 71 (heater wire). Then, an output voltage sensitivity per unit heating value (W/g) by a gamma-ray is calibrated by means of the GT signal processing unit 48 on the basis of the previously measured (already known) heater resistance value and the mass (heating conversion mass) of the fixed GT detector 44. As the time interval of calibrating the output voltage sensitivity of the GT assembly 35, the nuclear instrumentation control process module 60M1 stores, when the GT assembly 35 is initially mounted in the core 3, the sensitivity time series data of each GT detector 44 computed by the GT signal processing unit 48 as the predetermined shortest time interval (for example, the first time interval) as the table with respect to the in-core elapse time in the memory unit 61. Furthermore, the nuclear instrumentation control process module 60M1 estimates and computes an output voltage sensitivity change curve from the latest two or more points time series data from the present point of time, and then, compares the output voltage sensitivity change value with the third judgement value set to a predetermined future time after the first time interval, and with the third judgement value set to a future time after the second time interval. In the case where the output voltage sensitivity change value does not exceed the third judgement value even after the longer second time interval, a heater heating calibration of each detector 44 is carried out at the predetermined second longer time interval. With the elapse of in-core elapse time, in a state that the heater calibration time interval of the GT assembly of a longer, e.g., a third time interval is set, the nuclear instrumentation control process module 60M1 estimates and computes the output voltage sensitivity change curve from the latest two or more points time series data from the present point of time, and then, compares the output voltage sensitivity changes with predetermined future time points, that is, the third judgement value set with respect to a future time after the third time interval and a future time after the fourth time interval. Then, the nuclear instrumentation control process module 60M1 carries out the following output voltage calibration process for controlling the GT heater control unit 53 and the GT signal processing unit 48. More specifically, the processes includes: 1) in the case where the output voltage sensitivity change does not exceed the third judgement value even after the longer fourth time interval, a heater heating calibration of each GT detector 44 is carried out at the predetermined fourth longer time interval; 2) in the case where the output voltage sensitivity change does not exceed the third judgement value after the predetermined third time interval, but exceeds the third judgement value after the predetermined longer fourth time interval, a heater heating calibration of each detector 44 is carried out at the predetermined third longer time interval; and 3) in the case where the output voltage sensitivity change exceeds the third judgement value after the longer third time interval, a heater heating calibration of each detector 44 is carried out at the predetermined shorter second time interval, or the maximum value of the time intervals previously set for satisfying the third judgement value (in this case, either of the first or second time interval). Moreover, the nuclear instrumentation control process module 60M1 outputs a warning signal to the display unit 63 at the point of time when the heater calibration time interval of the GT assembly 35 is changed over from the previous time interval into a new time interval, and thus, informs the change-over of heater calibration time interval of the operator. [Third embodiment] The following is a description on a third embodiment of an in-core fixed nuclear instrumentation system, a power distribution simulating system and a power distribution monitoring system of the present invention. This third embodiment has the same construction and operation of the reactor power distribution monitoring system 29, the in-core fixed nuclear instrumentation system 30 and the power distribution simulating system 31 as the first and second embodiments of the present invention. Like reference numerals are used to designate the same constituent components as the first embodiment (see FIG. to FIG. 11), and the details are omitted. In the in-core fixed nuclear instrumentation system 30 and the power distribution simulating system 31 according to the third embodiment, discrimination processing of the GT assembly 35 for calibrating an output voltage sensitivity by heater heating of the fixed GT detector 44 is basically different from those shown in the first and second embodiments of the present invention. Namely, this third embodiment has been made in combination with the first and second embodiments of the present invention. The process control computer 31 is provided with the CPU 60 having the nuclear instrumentation control process module 60M1. The nuclear instrumentation control process module 60M1 has a function of computing a reactor operating time (in-core elapse time) after the GT assembly 35 is mounted in the core 3, and a function of storing the reactor operating time (in-core elapse time). Similarly to the first embodiment, a plurality of heater calibration time intervals {first heater calibration time interval (for example, 48 hours), second heater calibration time interval (for example, 165 hours), third heater calibration time interval (for example, 336 hours) and fourth heater calibration time interval (for example, 1000 hours)} are previously stored in the memory unit 61 so as to correspond to the in-core elapse times. In a state that there is no change in a normal operating parameter (the core state data including the reactor power level, the core cooling flow rate, the control rod pattern, etc.), in accordance with the elapse of time after being mounted in the core 3 (in-core elapse time in reactor operating mode), as described in the first embodiment, the nuclear instrumentation control process module 60M1, when executing the heater calibration, refers to the memory unit 61, thereby, automatically discriminating and selecting the GT assembly of heater calibration target from all GT assemblies 35 on the basis of the elapse time until the present from the previous calibration processing time of each GT assembly 35 and the heater calibration time interval corresponding to the computed present in-core elapse time of each GT assembly 35. Furthermore, the nuclear instrumentation control process module 60M1 executes a re-judgement process at each GT heater calibration process and registers the heater calibration time interval corresponding to in-core elapse time of the discriminated GT assembly 35 in the memory unit 61, and then, transmits the registered heater calibration time interval corresponding to the in-core elapse time corresponding to each GT assembly 35 to the display unit 63 so as to display the registered heater calibration time interval of each GT assembly 35 on the heater calibration time interval registration image of the display unit 63. Furthermore, the nuclear instrumentation control process module 60M1 automatically transmits a heater calibration processing execution instruction with respect to the GT assembly 35 selected as a heater calibration target, to the GT heater control unit 53 and the GT signal processing unit 48, or transmits a GT calibration instruction command transmission request with respect to the GT assembly 35 selected as a heater calibration target, to the display unit 63 so as to output and display the request onto the operator via the display unit 63. In the GT assembly 35 requiring the heater calibration process displayed on the display unit 63, heater calibration is carried out in a predetermined sequence by means of the GT signal processing unit 48 and the GT heater control unit 53 according to an (automatic or manual) instruction of the nuclear instrumentation control process module 60M1 of the CPU 60 of the process control computer 31. That is, the GT heater control unit 53 starts to supply a power (applies a voltage) to the built-in heater 71 of the GT assembly 35 with time delay corresponding to a positional address of the transmitted heater calibration processing execution instruction, and then, controls the heater voltage to be applied to the built-in heater 71 so that a current value flowing through the built-in heater 71 becomes a predetermined value. Next, the GT heater control unit 53 measures the applied voltage value and the current value flowing through the built-in heater 71, and then, transmits the measured values to the GT signal processing unit 48. On the other hand, the GT signal processing unit 48 receives the heater calibration processing execution instruction (positional address) transmitted from the nuclear instrumentation control process module 60M1, and then, in accordance with the receiving timing, concurrently measures the non-heated thermocouple output voltage signal (mV) of each GT detector 44 of the GT assembly 35 having the positional address. Further, the GT signal processing unit 48 receives the heater applied voltage and the current measured value transmitted from the GT heater control unit 53, and then, in accordance with the receiving timing, concurrently measures the heated thermocouple output voltage signal (mV) of each GT detector 44 of the GT assembly 35 along the axial direction. Then, the GT signal processing unit 48 stores the non-heated and heated output voltage signals in each GT detector 44 of the measured heater calibration target GT assembly 35 in the memory unit 48B of the GT signal processing unit 48 for each GT detector 44. The above heater calibration process according to the heater calibration process execution instruction is continued while changing the positional addresses of the target GT assemblies until the heater calibration process to all of the target GT assemblies which have need of the heater calibration is finished. Further, the GT signal processing unit 48 can compute the sensitivity S.sub.0 of all GT detectors 44, and voltage and current of heater power supply of the heater calibration target GT assembly 35 on the basis of the stored no-heated and heated output voltage signals and heater power of each GT detector 44 according to the same procedures as the first embodiment. Then, the GT signal processing unit 48 stores the computed sensitivity S.sub.0 of each GT detector 44 in the memory unit 48B of the processing unit 48, and transmits the computed sensitivity S.sub.0 of each GT detector 44 to the nuclear instrumentation control process module 60M1 of the process control computer 31. The aforesaid sensitivity measurement process by the GT heater control unit 53 and the GT signal processing unit 48 on the basis of the heater calibration process execution instruction, is repeatedly carried out until process with respect to all heater calibration target GT assemblies 35 is completed. The nuclear instrumentation control process module 60M1 has a function of storing an elapse time change data of the output voltage sensitivity S.sub.0 of each GT detector 44 transmitted from the GT signal processing unit 48 by the above sensitivity measurement process of the GT detector 44. Further, the nuclear instrumentation control process module 60M1 estimates (simulates) an output voltage sensitivity change curve by a linear extrapolation or quadratic curve extrapolation on the basis of the time series data points, or estimates (simulates) a future output voltage sensitivity change curve by approximating the data points to a linear approximation equation, a quadratic curve approximate equation or to the least square approximation equation of a curve represented by the above equation (4) "a+b.multidot.e.sup.-.lambda..sub.t ". Then, the nuclear instrumentation control process module 60M1 simulates an estimated sensitivity change after a predetermined time interval (i.e., future time) to be automatically selected from the in-core elapse time at the present point of time from the point of the last heater calibration time. In the case where the sensitivity change estimation value thus computed is more than the third judgement value (e.g., 1%), regardless of the next heater calibration time interval determined from the in-core elapse time stored in the memory unit 61, the nuclear instrumentation control process module 60M1 carries out update and registration with respect to the memory unit 61 at a predetermined time interval shorter one stage than the time interval selected at the present point of time, or at the maximum time interval of a plurality of the heater calibration time intervals previously stored in the memory unit 61 such that the sensitivity change is kept within a range of the third judgement value. Moreover, in the case where the sensitivity change estimation value thus computed is less than the third judgement value (e.g., 1%), the nuclear instrumentation control process module 60M1 registers the next heater calibration time interval as a predetermined time interval determined from the in-core elapse time (e. g., the next heater calibration time interval determined from the in-core elapse time stored in the memory unit 61). Then, the nuclear instrumentation control process module 60M1 researches the heater calibration time intervals of the axial GT detectors 44 of the GT assembly 35, and in the case where the heater calibration time interval of any one of detectors 44 (in this case, excluding detector which is determined as being synthetically failure, and is bypassed-registered to the nuclear instrumentation control process module 60M1), is registered in the memory unit 61 at a shorter time interval, the nuclear instrumentation control process module 60M1 transmits a heater calibration instruction of each GT detector 44 to the GT signal processing unit 48 and the GT heater control unit 53 on the basis of the heater calibration time interval having the shortest heater calibration time interval. According to this third embodiment, the sensitivity S.sub.0 of the GT detector 44 is calibrated at a predetermined shorter first time interval when the GT assembly 35 is initially mounted in the core 3, and then, when a predetermined in-core elapse time elapses, the heater calibration time interval is updated and registered as a predetermined longer second heater calibration time interval corresponding to the in-core elapse time. Thereafter, according to the above procedure, in accordance with the elapse of the in-core elapse time, a longer heater calibration time interval is automatically selected, and then, is successively updated and registered. However, in the case where the GT detector sensitivity S.sub.0 change after the elapse of future time (equivalent to a GT heater calibration time interval selected according to the present in-core elapse time) estimated from the data trend (time change data) of GT detector sensitivity S.sub.0 on the latest two or more points, is more than the third judgement value (e.g., 1%), the heater calibration time interval is temporarily changed and registered to the shorter time interval. Thereafter, automatic research process of each heater calibration time interval registered in the memory unit 61 is carried out in the GT detectors 44 discretely arranged in the axial direction of the identical GT assembly 35. Then, in the case where even one of GT heater calibration time intervals having a shorter time interval is registered in the GT detectors 44 which are not bypass-registered, the heater calibration time interval of all GT detectors 44 of the corresponding GT assembly 35 is automatically corrected and registered as the short time interval in the memory unit 61. In the finally registered GT heater calibration time interval, in the case where the time interval is changed over from the previous heater calibration time interval into a new heater calibration time interval, concerning the corresponding GT assembly 35, the nuclear instrumentation control process module 60M1 has a function of flashing and displaying the corresponding heater calibration time interval data on the heater calibration time interval registration image on the display unit 63 so as to give a caution to the operator. Thus, the operator refers to the trend graph of the output voltage sensitivity S.sub.0 of the GT detector 44 displayed on the display unit 63 while checking the flashed and displayed heater calibration time interval and comparing the previous output voltage sensitivity S.sub.0. In the manner as described above, the final registration result of the GT heater calibration time interval registered in the memory unit 61 is displayed on the display unit 63 by means of the nuclear instrumentation control process module 60M1 so that the operator can see the result. More specifically, when reaching a predetermined heater calibration time corresponding to the heater calibration time interval of the predetermined GT assembly 35, the nuclear instrumentation control process module 60M1 gives a warning for carrying out the GT heater heating calibration process to the operator via the display unit 63, and then, displays the target GT assembly 35 on the display unit 63 on the basis of the aforesaid registration result. In the GT assembly 35 changed into the next calibration time interval shorter than the heater calibration time interval selected from the in-core elapse time, the sensitivity change until the next heater calibration time determined by the in-core elapse time at the present time after the next heater calibration, will be estimated, and then, in the case where the sensitivity change satisfies the third judgement value, the time interval will be returned to the heater calibration time determined by the in-core elapse time. On the other hand, in the case where the sensitivity change does not satisfy the third judgement value, the next heater calibration will be again carried out at the shorter time interval. The above GT heater heating calibration process is carried out by the instruction of the process control computer 31 according to the same procedures as the first embodiment, and then, a newly obtained output voltage sensitivity S.sub.0 of the GT detector 44 is transmitted to the process control computer 31. The operator's procedures after that is the same as the first embodiment. As described above, according to this third embodiment, the sensitivity S.sub.0 change of the GT detector 44 in a predetermined future time is estimated from the latest S.sub.0 trend data with respect to the sensitivity S.sub.0 of the GT detector 44 at the present time, and in the case where the estimated change value is more than the third judgement value, a GT heater calibration time interval having a short time interval is registered. Thus, it is possible to flexibly cope with an unexpected output voltage sensitivity change of the GT assembly 35, and to prevent a deterioration of power distribution measurement accuracy by the GT assembly 35. Moreover, in the case where the heater calibration time interval is changed over from the previous time interval into a new time interval, the above change-over is informed of the operator via the display unit 63. Thus, it is possible to give a warning of inspection and caution for any failures to the operator, and to improve an efficiency for maintaining a reliability of reactor in-core instrumentation. According to this third embodiment, in the in-core fixed nuclear instrumentation system 30 of the reactor, the plurality of fixed neutron detectors (LPRM detectors) 37 for detecting the local power distribution of power range in the reactor core and fixed GT detectors 44 gamma-ray thermometer assembly 35 for detecting the .gamma.-ray heating value, are housed in the nuclear instrumentation tube 33. Further, the in-core fixed nuclear instrumentation system 30 includes: the in-core nuclear instrumentation assembly 32 in which the GT detectors 44 are arranged at least in the vicinity of the fixed LPRM detectors 37; the LPRM signal processing unit 40 for processing the LPRM signal S2 from the LPRM detector 37; the GT signal processing unit 48 for processing the output voltage signal (GT signal) S1 from the gamma-ray thermometer assembly 35; the GT heater control unit 53 for carrying out an electrically energizing control with respect to the heaters 71 built in the GT assembly 35; and the nuclear instrumentation control process module 60M1 for calculating (computing) and storing the in-core elapse time or in-core irradiation quantity (burn-up quantity) of the GT assembly 35. The nuclear instrumentation control process module 60M1 controls the operations of the GT heater control unit 53 and the GT signal processing unit 48. During a reactor operation, when the heater wire of the built-in heater 71 is electrically energized by means of the GT heater control unit 53, the thermocouple output voltage increasing sensitivity of the fixed GT detector 44 of the GT assembly 35 with respect to heater heating (additional heating value), is measured by means of the GT signal processing unit 48 on the basis of a heating voltage and current of the built-in heater 71 (heater wire). Then, an output voltage sensitivity per unit heating value (W/g) by a gamma-ray is calibrated by means of the GT signal processing unit 48 on the basis of the previously measured (already known) heater resistance value and the mass (heating conversion mass) of the fixed GT detector 44. In the case of calibrating the output voltage sensitivity of the GT assembly 35, in the nuclear instrumentation control process module 60M1, the in-core elapse time (or in-core irradiation quantity) of the GT assembly 35 is used as the parameter, and the heater calibration time interval is selected according to the plurality of heater calibration time intervals to be selected in accordance with the in-core elapse time previously stored in the memory unit 61, and thus, the heater calibration by heater heating is carried out via the GT signal processing unit 48 and the GT heater control unit 53. Further, the nuclear instrumentation control process module 60M1 stores the sensitivity time series data of each GT detector 44 computed by the GT signal processing unit 48, and estimates the sensitivity change curve from time series data of two or more latest points from the present point of time. In the case where the sensitivity change exceeds the third judgement value set with respect to a predetermined future time determined by the in-core elapse time, the time interval is changed into a calibration time interval having a short time interval, and the heater calibration is carried out by controlling the GT heater control unit 53 and the GT signal processing unit 48. [Fourth embodiment] The following is a description on a fourth embodiment of an in-core fixed nuclear instrumentation system, a power distribution simulating system and power distribution monitoring system of the present invention. This fourth embodiment has the same structure and operation of the reactor power distribution monitoring system 29, the in-core fixed nuclear instrumentation system 30 and the power distribution simulating system 31 as the first and second embodiments of the present invention. Like reference numerals are used to designate the same constituent components as the first embodiment (See FIG. 1 to FIG. 11), and the details are omitted. The reactor in-core fixed nuclear instrumentation system 30 and power distribution monitoring system 29 of this fourth embodiment are a modification of the first, second and third embodiments of the present invention. In the first, second and third embodiments of the present invention, a reactor operation elapse time (in-core elapse time) after the GT assembly 35 is mounted in the reactor core 3 has been used as the parameter in the GT heater calibration process of the sensitivity S.sub.0 of the GT detector 44. According to this fourth embodiment, the process control computer 31 has a function of computing a neutron irradiation quantity of the in-core nuclear instrumentation assembly 32. With the use of the function of computing the neutron irradiation quantity, the computed neutron irradiation quantity is used as the parameter in place of the in-core elapse time of the in-core nuclear instrumentation assembly 32. The computation of the in-core neutron irradiation quantity of each GT assembly 35 is carried out by means of the power distribution simulation process module 60M2 of the CPU 60 of the process control computer 31. More specifically, the power distribution simulation process module 60M2 carries out the computation of the in-core neutron irradiation quantity with the use of the three-dimensional nuclear thermal-hydraulics simulation code of the power distribution simulation program module PM2 which is a BWR three-dimensional simulator module. By using the power distribution simulation program module PM2, not only the in-core elapse time of the in-core nuclear instrumentation assembly 32 but also an elapse time change of thermocouple by a neutron irradiation is taken into consideration and is used as the parameter. In this case, in place of accurately carrying out the computation of the neutron irradiation quantity, an average burn-up increment of nodes surrounding the GT detector 44 is used as the parameter substantially proportional to the neutron irradiation quantity, and then, an accumulative burn-up increment of each GT detector 44 after being mounted in the core 3 may be used as the parameter. Thus, by the power distribution simulation module 60M2 of the process control computer 31, it is possible to more accurately reflect a sensitivity change of the GT detector 44 which varies by the in-core neutron irradiation. This fourth embodiment shows the modification of the first, second and third embodiments of the present invention. In the in-core fixed nuclear instrumentation system of this fourth embodiment, the nuclear instrumentation control process module 60M1 stores the in-core elapse time or in-core neutron irradiation quantity of each GT detector 44 of the each GT assembly 35, and then, evaluates the stored data so as to control the heater heating calibration of each GT detector 44. [Fifth embodiment] The following is a description on a fifth embodiment of an in-core fixed nuclear instrumentation system, a power distribution simulating system and a power distribution monitoring system of the present invention. This fifth embodiment is a modification of the first to fourth embodiments of the present invention. This fifth embodiment has the same construction and operation of the reactor power distribution monitoring system 29, the in-core fixed nuclear instrumentation system 30 and the power distribution simulating system 31 as the first and second embodiments of the present invention. Like reference numerals are used to designate the same constituent components as the first embodiment (See FIG. 1 to FIG. 11), and the details are omitted. In the first to fourth embodiments of the present invention, at the point of time of carrying out the GT heater calibration, for example, the operator confirms that reactor operating mode parameters (the core state data S3 including the core power distribution, the core coolant flow rate, the control rod pattern, etc.) are constant, and then, in the state that the above core state data is constant, the GT heater calibration has been carried out. However, in fact, the GT assembly 35 measures a gamma-ray heating value; for this reason, a response characteristic of the GT assembly 35 does not reach an accurate fission rate, that is, a GT signal level proportional to an in-core local power unless a nuclide decay chain of gamma-ray source is in an equilibrium state. Therefore, even if the reactor operating mode is a fixed state, unless the state is continued for a predetermined time, an accurate GT signal level is not obtained. According to this fifth embodiment, the reactor core state data measuring device 55 detects core state changes such as the change of the core power distribution, the core coolant flow rate and the control rod pattern, and then, the nuclear instrumentation control process module 60M1 of the process control computer 31 makes a decision whether a predetermined time elapses after detecting the core state change, on the basis of the transmitted core state data D3. More specifically, the nuclear instrumentation control process module 60M1 stores the change point of time when the core state (the parameter such as the core state data D3) varies, in the memory unit 61, and then, according to the same process as the first to fourth embodiments, when it is determined that a present time is reached to a predetermined time for executing the GT heater calibration process, the nuclear instrumentation control process module 60M1 automatically makes a decision whether a predetermined necessary time elapses after the change point at which the above core state (the parameter) varies, on the basis of the core state data D3. Further, the nuclear instrumentation control process module 60M1 displays the result on the display unit 63, and the operator confirms that the heater calibration process of the GT assembly 35 is suitably carried out, and manually starts up the GT heater calibration process, or automatically starts up the GT heater calibration process after the predetermined necessary time is elapsed from the change point at which the above core state (the parameter) varies. In this structure of this embodiment, in a state that the GT signal level is in a non-equilibrium transient state, it is possible to prevent the sensitivity S.sub.0 of the GT detector 44 from being calibrated by additional heating of the built-in heater 71, thereby, preventing an inaccurate conversion of unit heating value (W/g) of gamma-ray heating from the output voltage (mV) signal by the GT detector. [Sixth embodiment] The following is a description on a sixth embodiment of an in-core fixed nuclear instrumentation system, a power distribution simulating system and a power distribution monitoring system of the present invention. This sixth embodiment has the same construction and operation of the reactor power distribution monitoring system 29, the in-core fixed nuclear instrumentation system 30 and the power distribution simulating system 31 as the first embodiment of the present invention. Like reference numerals are used to designate the same constituent components as the first embodiment (See FIG. 1 to FIG. 11), and the details are omitted. The reactor power distribution monitoring system 29 of this sixth embodiment has the substantially same block configuration as shown in the reactor power distribution monitoring system of FIG. 1. More specifically, as shown in FIG. 1, the reactor pressure vessel 2 is housed in the primary containment vessel 1, and the reactor core 3 is housed in the reactor pressure vessel 2. As shown in FIG. 2 and FIG. 3, the reactor core 3 is constructed in a manner that a large number of fuel assemblies 4 and control rods 5 are mounted therein. In the reactor pressure vessel 2, the in-core nuclear instrumentation assembly 32 of the in-core fixed nuclear instrumentation system 30 is mounted in the fuel gap G between four fuel assemblies 4. The in-core fixed nuclear instrumentation assembly 32 includes the nuclear instrumentation tube 33, a plurality of (N) fixed neutron detectors (LPRM detectors) 37 constituting the neutron detector assembly 34 and a plurality (.gtoreq.N) of fixed .gamma.-ray heating detectors (GT detectors) 44 constituting the .gamma.-ray heating detector assembly 35, which are housed in the nuclear instrumentation tube 33. Meanwhile, the in-core fixed nuclear instrumentation system 30 has the power range neutron flux measuring system 41 and the gamma thermometer power distribution measuring system 50. The power range neutron flux measuring system 41 is composed of the plurality of fixed LPRM detectors 37 mounted in the core 3 and the signal processing unit 40. On the other hand, the gamma thermometer power distribution measuring system 50 is composed of the GT assemblies 35 each having a plurality (.gtoreq.N) of the GT detectors 44 and the GT signal processing unit 48. Thus, the in-core fixed nuclear instrumentation assembly 32 is provided with a detector group of the in-core fixed nuclear instrumentation system 30, and the in-core nuclear instrumentation assembly 32 including the fixed detector group measures neutron flux and the .gamma.-ray heating value at predetermined measurement points in the core 3. Further, the in-core fixed nuclear instrumentation system 30 includes the gamma-ray thermometer control unit 53 for supplying the power to the built-in heater 71 of the GT assembly 35. The gamma-ray thermometer control unit 53 carries out a control for electrically energizing the built-in heater 71 of the GT assembly 35 via the power cable 54 so as to adjust and control a heater heating value. Moreover, in the reactor pressure vessel 2 or a primary system piping (not shown), a core state data measuring device 55 is provided. The core state data measuring device 55 measures core state data S3 (process quantity) including a control rod pattern, a core coolant flow rate, an internal pressure of the reactor pressure vessel, flow of feed water, a temperature of feed water (a core inlet coolant temperature) and so on, which are used as various operating parameters indicative of a reactor operating mode (state) of the reactor. Further, one part of the core state data measuring devices 55, which is inside of the vessel 1, is connected to a core state data processing unit 58 via a signal cable 57 penetrating through a penetration portion 56, and other part thereof, which is outside of the vessel 1, is connected via the signal cable 57 to the core state data processing unit 58, so that a process data measuring system 59 is constructed. In addition, the power distribution monitoring system 29 of this embodiment is provided with a process control computer 31. The process control computer 31 inputs the following data: more specifically, the GT data D1 (W/g signal) obtained by signal processing of the GT signal processing unit 48 based on the GT signal S1 detected by the GT detector 44; the LPRM data D2 obtained by signal processing of the LPRM signal processing unit 40 based on the LPRM signal S2 detected by the LPRM detector 27; and the core state data D3 obtained by signal processing of the core state data signal processing unit 58 based on the core state data signal S3 measured by the core state data measuring device 55. Further, the process control computer 31 simulates the reactor power distribution, and thus monitoring and controlling the in-core nuclear instrumentation system 30. The process control computer 31 includes the CPU 60 having the power distribution simulation process module 60M2. The process module 60M2 is operative to input the core state data D3, and to simulate the neutron flux distribution in the core 3, the power distribution therein and the margin with respect to the operational thermal limit value by executing the three-dimensional nuclear thermal-hydraulics simulation according to the physical model (three-dimensional nuclear thermal-hydraulics simulation code) of the program module PM2 stored in the memory unit 61 of the process control computer 31. Furthermore, the CPU 60 includes: the power distribution simulation process module 60M2 which corrects the power distribution simulation result of the process module 60M1 by referring the GT data D1 (W/g signal) or the LPRM data D2, and thus, obtains the core power distribution reflecting the actually measured data in the core 3 and having a high reliability; and the nuclear instrumentation control process module 60M1 for monitoring and controlling the in-core nuclear instrumentation system 30. In addition, the process control computer 31 has the memory unit 61, the input console and the display unit 63. The GT assembly 35 incorporated into the in-core fixed nuclear instrumentation system 30 of this sixth embodiment has the same structure as the GT assembly shown in FIG. 2, FIG. 3, FIG. 5 and FIG. 6 described in the first embodiment. By the way, the in-core power distribution simulation of the reactor, in which the large number of fuels assemblies 4 closely stand in the core 3, is carried out by the power distribution simulation process module 60M2 of the process control computer 31 according to the so-called three-dimensional nuclear thermal-hydraulics simulation. Then, the power distribution simulation process module 60M2 displays the in-core power distribution, the operational limit value {the MLHGR(kW/m) and the MCPR} of core fuel, the margin with respect to the operational limit value, to the operator via the display unit 63. According to this sixth embodiment, the core state data signal S3 indicative of the present core state obtained from the core state data measuring device 55 in the core 3 is collected by the core state data processing unit 58 (there is the case where it is carried out by the process control computer 31), and then, the reactor thermal output, the core inlet coolant temperature and so on are simulated. The core state data D3 including the simulation result of the reactor thermal output and so on is transmitted to the power distribution simulation module 60M2 of the CPU 60 of the process control computer 31 via the signal interface function of the nuclear instrumentation control process module 60M1 of the CPU 60. The process control computer 31, which monitors the reactor operating mode and the core power distribution, always continuously receives the core state data D3 (the parameters showing the state in the core 3), and then, the power distribution simulation process module 60M2 periodically (e.g., one time per hour) or always caries out the core power distribution simulation (three-dimensional nuclear thermal-hydraulics simulation) on the basis of the latest operating parameters of the reactor (the core state data D3) and the three-dimensional nuclear thermal-hydraulics simulation code of the program module PM2 in accordance with the simulation request command inputted by operating the input console 62 by the operator. Then, the process control computer 31 (the power distribution simulation process module 60M2) corrects the core power distribution obtained by the above three-dimensional nuclear thermal-hydraulics simulation of the process module 60M2 on the basis of the GT data D1 (W/g signal) or the LPRM data D2 at the point of time of the core power distribution simulation, thereby calculating the in-core power distribution and the margin with respect to the operational thermal limit value, which have high accuracy and high reliability. In addition, in this sixth embodiment, the axial in-core power distribution is adapted so as to be corrected with the GT data (w/g) converted by the GT signal processing unit 48 on the basis of the GT signals S1 detected by the GT detectors 44 having the number (e.g., the same number as LPRM detectors 37, that is, four or more) less than 24 nodes (there is the case of 12 nodes or 26 nodes) and the three-dimensional nuclear thermal-hydraulics simulation code. Related to some nodes in the axial direction at which the GT detector 44 is provided, difference between each simulation value of .gamma.-ray heating values of each axial node of each GT assembly 35 and each actual measured value (GT data D1 value) thereof is obtained by ratio. Then, by the power distribution simulation process module 60M2 of the CPU 60, data indicative of differences (ratios) between the respected actual .gamma.-ray heating values (GT data D1 values) of the GT detectors 44 having the limited number in the core axial direction and the respected simulation values of the .gamma.-ray heating values corresponding to the GT detectors 44 are interpolated and extrapolated in other (remained) nodes in the axial direction, respectively, wherein the GT detector 44 is not provided at the other nodes in the axial direction, thereby obtaining the correction data of the .gamma.-ray heating value differences with respect to the whole axial nodes. Incidentally, in addition to interpolation and extrapolation in the axial direction, it is possible to interpolate and extrapolate the .gamma.-ray heating value difference corrections (correction ratios; correction factors) with respect to radial positions at which the GT assemblies are not provided along a core radial direction. In the case of carrying out the above power distribution learning correction, when capturing (gathering) the GT data D1 (W/g) computed by the GT signal processing unit 48 on the basis of the GT signal S1 for the power distribution simulation process module 60M2, the nuclear instrumentation control process module 60M1 of the CPU 60 makes a decision whether or not a predetermined time, e.g., one hour or more elapses after the parameter (core state data S3) indicative of the core state varies, on the basis of the core state data D3. As a result, in the case where the predetermined time does not elapse, the nuclear instrumentation control process module 60M1 of the CPU 60 outputs a warning indicative that the predetermined time does not elapsed to the display unit 63, and thus, informs the result of the operator via the display unit 63. On the other hand, regardless of the aforesaid state that the predetermined time elapses or the state that the predetermined time does not elapse, the power distribution simulation process module 60M2 of the process control computer 31 carries out the three-dimensional nuclear thermal-hydraulics simulation so as to simulate the in-core power distribution, and then, the process module 60M2 learns and corrects the simulated in-core power distribution with the use of the GT data D1 based on the GT signal S1. At this time, in the case where the in-core power distribution is obtained by adapting and correcting it by the GT data D1 based on the GT signal S1 measured in a state that the predetermined time does not elapse, that is, the GT data D1 based on the GT signal S1 of non-equilibrium state, the nuclear instrumentation control process module 60M1 of the CPU 60 outputs a warning indicative of the power distribution adapting correction result based on the GT signal of non-equilibrium state to the display unit 63 so as to be displayed thereon thereby informing the warning to the operator. The power distribution adapting process (method) by the power distribution simulation process module 60M2 of the process control computer 31 is substantially the same as the contents described in the specification and drawings in the U.S. patent application No. 09/271,350. Thus, in FIG. 13, there is shown a schematically flow chart of the core power distribution simulating process and the power distribution adapting correction process of the CPU 60. More specifically, as shown in FIG. 13, the power distribution simulation process module 60M2 of the CPU 60 of the process control computer 31 carries out the three-dimensional nuclear thermal-hydraulics simulation based on the core state data D3 and the three-dimensional nuclear thermal-hydraulics simulation code of the program module PM2, and thus, an in-core power distribution Pn (I, J, K) is simulated (step S51). In this case, an additional character (I, J, K) denotes a position of each node of the fuel assembly, and n denotes iteration number during a core power distribution simulation at the present iteration. Next, the power distribution simulation process module 60M2 makes a decision whether or not a difference between a node core power distribution Pn (I, J, K) at the present iteration (n) and node core power distribution Pn-1 (I, J, K) at the previous iteration (n-1) is less than a predetermined (fixed) value (step S52). If the decision of step S52 is YES, the process module 60M2 calculate an operational thermal limit value (minimum critical power ratio: MCPR, and maximum linear heat generating ratio: MLHGR) and a margin based on the operational thermal limit value so as to output the operational thermal limit value and so on to the display unit 63 so as to be displayed thereon (step S53). If the decision of step S52 is NO, that is, when the simulation process of the core power distribution is not sufficiently repeated, the power distribution simulation process module 60M2 determines a simulation value {Wc (I, J, K)} of .gamma.-ray heating value on the basis of the simulated core power distribution (step S54). Meanwhile, as described above, the GT signal processing unit 48 reads the thermocouple output voltage signal U.sub..gamma. (I, J, K) detected from the GT detector 44 (step S55), and then, the read thermocouple output voltage signal U.sub..gamma. (I, J, K) is converted into the gamma-ray heating value W.sub.m (I, J, K) (corresponding to GT data D1)(step S56). At this time, the power distribution simulation process module 60M2 determines difference data (.gamma.-ray heating value difference correction data) between the calculated simulation value {Wc (I, J, K)} of .gamma.-ray heating value and the gamma-ray heating value Wm (I, J, K). Then, the difference data is interpolated and extrapolated in each node of the core axial direction, and thus, .gamma.-ray heating value difference correction data BCF.sub.IJK with respect to all axial nodes are determined (step S57). In the power distribution simulation process module 60M2, the simulated reactor power distribution on the way of final converged value Pn (I, J, K) is corrected {Pn (I, J, K).fwdarw.P'n (I, J, K)} so that the .gamma.-ray heating value difference correction data (corrective coefficient) BCF.sub.IJK with respect to all axial nodes are "1.0", that is, the simulation value {Wc (I, J, K)} of the .gamma.-ray heating value in each node and the gamma-ray heating value Wm (I, J, K) are coincident with each other. A correction ratio {power distribution correction quantity (adapting correction quantity)} for each fuel assembly node at this iteration is stored in the memory unit 61 of the process control computer 31 (step S58). Then, in the process module 60M2, an adjustment factor of the three-dimensional nuclear thermal-hydraulics code (physical model) is guessed in accordance with P'n (I, J, K) corrected by the gamma-ray heating value Wm (I, J, K) (GT data value) based on the actually measured GT signals (step S59), after the step S59, the process of the process module 60 returns to step S51. Then, the above iteration processes of the step S51 to step S59 of the process module 60M2 are repeated until the judgement of step S52 is YES. Finally, when the judgement of step S52 is YES, the corrected results including the corrected core power distribution, the operational thermal limit value (MCPR, MLHGR) and so on are obtained (step S53). As described above, the amendment of the adjustment factor of the repetition simulation is executed according to the three-dimensional nuclear thermal-hydraulics simulation code (physical model) so that the power distribution simulation process module 60M2 carries out the power distribution simulation after the next iteration (n+1 iteration) (see step S51), and thus, when the step S52 is converged so that the judgement of the step S52 is YES, a core power distribution having a high accuracy is obtained. Incidentally, as a another adapting correction process of the CPU 60, FIG. 14 shows a modification of the flow chart of FIG. 13. That is, as shown in FIG. 14, the power distribution simulation process module 60M2 of the CPU 60 of the process control computer 31 carries out the three-dimensional nuclear thermal-hydraulics simulation with iteration method based on the core state data D3 and the three-dimensional nuclear thermal-hydraulics simulation code of the program module PM2 so as to simulate the in-core power distribution Pn (I, J, K) (step S51 in FIG. 14). Next, the power distribution simulation process module 60M2 makes a decision whether or not a difference between a node core power distribution Pn (I, J, K) at the present iteration (n) and node core power distribution Pn-1 (I, J, K) at the previous iteration (n-1) is less than a predetermined (fixed) value (step S52). When the decision of step S52 is NO, in a case where the simulation process of the core power distribution is not sufficiently repeated, the process of the process module 60 returns to step S51, and the power distribution simulation process module 60M2 carries out next (n+1) iteration three-dimensional nuclear thermal-hydraulics simulation so as to simulate (n+1) iteration in-core power distribution Pn+1(I, J, K). On the other hand, when the decision of step S52 is YES, the process module 60M2 determines a simulation value {Wc (I, J, K)} of .gamma.-ray heating value on the basis of the simulated core power distribution (step S54). Meanwhile, as described above, the GT signal processing unit 48 reads the thermocouple output voltage signal U.sub..gamma. (I, J, K) detected from the GT detector 44 (step S55), and then, the read thermocouple output voltage signal U.sub..gamma. (I, J, K) is converted into the gamma-ray heating value W.sub.m (I, J, K) (corresponding to GT data D1)(step S56). At this time, the process module 60M2 determines difference data (.gamma.-ray heating value difference correction data) between the determined simulation value {Wc (I, J, K)} of .gamma.-ray heating value and the gamma-ray heating value Wm (I, J, K). Then, the difference data is interpolated and extrapolated in each node of the core axial direction, and thus, .gamma.-ray heating value difference correction data BCF.sub.IJK with respect to all axial nodes are determined (step S57). In the process module 60M2, the simulated reactor power distribution Pn (I, J, K) is corrected {Pn (I, J, K).fwdarw.P'n (I, J, K)} so that the .gamma.-ray heating value difference correction data (corrective coefficient) BCF.sub.IJK with respect to all axial nodes are "1.0", that is, the simulation value {Wc (I, J, K)} of the .gamma.-ray heating value in each node and the gamma-ray heating value Wm (I, J, K) are coincident with each other. A correction ratio {power distribution correction quantity (adapting correction quantity)} for each fuel assembly node at this time is stored in the memory unit 61 of the process control computer 31 (step S58), the process of the CPU 60 returns to step S53. The power distribution simulation process module 60M2 obtains the operational thermal limit value (MCPR, MLHGR) and so on according to the basis of the corrected core power distribution (step S53). As a result of that, when the step S52 is converged so that the judgement of the step S52 is YES, a core power distribution having a high accuracy is obtained. Incidentally, in this modification, the corrected core power distribution is contradicted to a neutron flux distribution, but the adapting correction process shown in the modification is one process in a large number of adapting correction processes. Still furthermore, FIG. 15 is an explanatory view of interpolating and extrapolating the ratios (symbol E in FIG. 15) of the simulated GT signal levels on the positions where the GT detectors 44 are provided to the actually measured GT signal levels in the axial 24 nodes. In this case, a linear interpolation is carried out, and both ends are extrapolated as ratios of GT detectors 44 on upper and lower ends being kept constant. That is, in FIG. 15, a symbol F represents a result of the approximation line of the linear interpolation. Moreover, the interpolation and extrapolation may be a quadratic curve. On the other hand, in this embodiment, the nuclear instrumentation control process module 60M1 of the CPU 60 initializes elapse time counter (step S70A), and counts the elapse time (step S70B). Next, the CPU 60 judges whether or not the core state change is detected by the core state data (step S70C). In the case where the core state change is detected, that is, the judgement of step S70C is YES, the CPU 60 returns to a process of step S70A. In the case where the core state change is not detected, that is, the judgement of step S70C is NO, the CPU 60 judges whether or not the LPRM detector gain adjustment instruction is requested by the input console 62 (step S70D). In the case where the LPRM detector gain adjustment instruction is not requested, that is, the judgement of step S70D is NO, the CPU 60 returns to a process of step S70B. On the other hand, in the case where the LPRM detector gain adjustment instruction is requested, that is, the judgement of step S70D is YES, the process module 60M1 makes a decision whether or not a predetermined time, e.g., one hour or more elapses after the parameter (the core state data S3) varies (step S71). In the case where the predetermined time, for example, one hour, elapses, that is, the judgement of step S71 is YES, the process module 60M1 confirms that the predetermined time elapses so as to transmit an adjustment execution instruction of at least one of a sensitivity and a gain of the LPRM detector 37 including the gamma-ray heating value Wm (I, J, K) (GT data value) of each node to the LPRM signal processing unit 40 via the nuclear instrumentation control process module 60M1, periodically or in accordance with the operator's operating instruction via the input console 61 (step S72). The LPRM signal processing unit 40 sets at least one of the sensitivity and the gain of each LPRM detector 44 to a value corresponding to the gamma-ray heating value Wm (I, J, K) (Unit: W/g) on the identical node position, or to a value proportional thereto, in accordance with the transmitted adjustment execution instruction. On the other hand, in the case where the predetermined time does not elapse after detecting the core state change, that is, the judgement of step S71 is NO, the process module 60M1 of the CPU 60 does not carry out the LPRM detector sensitivity and gain adjustment process, that is, does not transmit the adjustment execution instruction to the LPRM signal processing unit 40 at this point of time, and waits until the next predetermined time (period) adjustment (or the next adjustment instruction is transmitted by the operator) or until the above predetermined time, (e.g., one or more hour) elapses (step S73). In the case of waiting, the process module 60M1 outputs an information representing a waiting state as a warning to the operator via the display unit 63. As described above, according to this sixth embodiment, the operator is prevented from unnoticeably using an error caused by power distribution adapting by the GT detection signal obtained in a state (non-equilibrium state) which does not reach an equilibrium state of gamma decay chain in which the output signal level of the GT detector 44 of the GT assembly 35 accurately corresponding to the core power distribution. Thus, it is possible to improve a reliability of the reactor power distribution monitoring system 29, the in-core fixed nuclear instrumentation system 30 and the power distribution simulating system 31. Furthermore, the adjustment of at least one of the sensitivity and the gain of the LPRM detector 37 is prevented from being carried out with the use of the GT signal of the aforesaid non-equilibrium state; therefore, it is possible to further improve a reliability of the reactor power distribution monitoring system 29, the in-core fixed nuclear instrumentation system 30 and the power distribution simulating system 31. According to this sixth embodiment, in the in-core fixed nuclear instrumentation system 30, a plurality of fixed neutron detectors (LPRM detectors) 37 for detecting a local power distribution of power range in the reactor core 3 and fixed GT detectors 44 gamma-ray thermometer assembly 35 for detecting a .gamma.-ray heating value, are housed in the nuclear instrumentation tube 33. Further, the in-core fixed nuclear instrumentation system 30 includes: the in-core nuclear instrumentation assembly 32 in which the GT detectors 44 are arranged at least in the vicinity of the fixed LPRM detectors 37; the LPRM signal processing unit 40 for processing the LPRM signal S2 from the LPRM detector 37; the GT signal processing unit 48 for processing the output voltage signal S1 from the gamma-ray thermometer assembly 35; the GT heater control unit 53 for carrying out an electrically energizing control with respect to the heaters 71 built in the GT assembly 35; the core state data measuring device 55 for detecting core state data indicative of the core states such as a reactor power level, core coolant flow rate, control rod pattern or the like; and the nuclear instrumentation control process module 60M1 of the CPU 60 for calculating and storing the in-core elapse time or in-core irradiation burn-up of the GT assembly 35 in the memory unit 61. The nuclear instrumentation control process module 60M1 makes a decision whether or not the predetermined time elapses after detecting the core state change in accordance with the core state data while outputting the result to the display unit 63 so as to inform it of the operator as a warning. Further, the nuclear instrumentation control process module 60M1 has an interface function of the process control computer 31, and gathers the GT data D1 (W/g signal) based on the GT signal S1 outputted from the GT assembly 35. Then, the power distribution simulation process module 60M2 simulates a reactor power distribution with the use of the gathered GT data D1, and corrects the power distribution result based on the physical model so as to obtain a core power distribution reflecting the actually measured data and having a high reliability. Furthermore, the nuclear instrumentation control process module 60M1 transmits the adjustment instruction of at least one of the sensitivity and the gain of the LPRM detector 37 to the LPRM signal processing unit 40 so as to execute the adjustment of at least one of the sensitivity and the gain of the LPRM detector 37 with the use of the GT data D1 (.gamma.-ray heating value; W/g signal) based on the GT signal S1 from the GT detector 44 of the same core axial position as the LPRM detector 37 in the identical in-core nuclear instrumentation tube 33. [Seventh embodiment] The following is a description on a seventh embodiment of an in-core fixed nuclear instrumentation system, a power distribution simulating system and a power distribution monitoring system of the present invention. This seventh embodiment shows a modification of the reactor power distribution monitoring system 29, the in-core fixed nuclear instrumentation system 30 and the power distribution simulating system 31, which have been described in the sixth embodiment. In the sixth embodiment, the adjustment of at least one of the sensitivity and the gain of the LPRM detector 37 has been carried out so as to coincide with the GT data (W/g detection signal) of the GT detector 44 located on the same position as the LPRM detector 37. On the contrary, in this seventh embodiment, the memory unit 61 of the process control computer 31 stores a predetermined correlation equation data showing a relationship between a reading value (simulation value) of the LPRM detector 37 and nodal power of the fuel nodes around the LPRM detector, wherein the correlation equation data has a correlation parameter including a fuel type (enrichment distribution in a horizontal cross section, Gd distribution design type), control rod inserted state, node burn-up, historical node void fraction, instantaneous node void fraction, etc.}. Then, in the process control computer 31, the CPU 60 (the power distribution simulation process module 60M2), as shown in FIG. 17, receives the request for executing the LPRM data gain adjustment transmitted from the input console S80. Then, in the case where it is determined that the GT signal is in an equilibrium state on the basis of the core state data (the judgement of step S81 is YES), the CPU 60 simulates and corrects the three-dimensional power distribution in the core 3 by using the GT data D1 (W/g signal) based on the GT signal S1 detected by each GT detector 44 of the GT assembly 35 (step S82). In accordance with the simulated result of the power distribution and the correlation equation data, the CPU 30 of the process computer 31 simulates a response value of the LPRM detector 37 on the basis of fuel nodal power around the LPRM detector 37 obtained by the simulated result of the power distribution (step S83). The process module 60M1 of the CPU 60 compares the simulated response value of the LPRM detector 37 and the actually measured LPRM data D2 value (step S84) so as to simulate a gain adjustment factor (RGAF.sup.L ; where L represents an address of the LPRM detector 37) related to at least one of the sensitivity and the gain of the LPRM detector 37 so that the actual detection value (actually measured LPRM data D2 value) of the LPRM detector 44 coincides with the transmitted simulated response value thereof (step S84). Next, the process module 60M1 transmits the gain adjustment factor RGAF.sup.L and an adjustment instruction of at least one of the sensitivity and the gain of the LPRM detector 37 to the LPRM signal processing unit 40, respectively. The LPRM signal processing unit 40 adjusts the actually measured LPRM data D2 so as to multiply the actually measured LPRM data D2 by the gain power factor RGAF.sup.L or by an integration gain adjustment factor GAF.sup.L.sub.n. The integration gain adjustment factor GAF.sup.L.sub.n by which the actually measured LPRM signal S2 is multiplied is obtained described above. EQU GAF.sup.L.sub.n =GAF.sup.L.sub.(n-1).multidot.RGAF.sup.L (5) The first gain adjustment factor is stored in the memory unit of the LPRM signal processing unit 40, and further, may be transmitted to the process control computer 31. The details of the correlation parameter used in the response simulation of the LPRM detector 37 are omitted because the correlation parameter is an already-known technique. In this sixth and seventh embodiments, in the case where it is determined that the GT detection signal S2 (GT data D2) does not reach an equilibrium signal level; namely, in the case where the nuclear instrumentation control process module 60M1 confirms that the predetermined time (e.g., one hour or more) does not elapse after the core state parameter (core state data D3) varies (the judgement of step S81 is NO), the process module 60M1 of the CPU 60 outputs a non-equilibrium warning of the GT detection signal to the display unit 63 so as to be displayed thereon and executes the process described in step S73 in FIG. 16, and, in the case where the judgement of step S81 is YES, the process module 60M2 carries out a power distribution adaptive simulation is carried out by the process module 60M2 on the basis of the detected GT signals. At this time, the process module 60M2 of the CPU 60 carries out the adaptive simulation on the basis of the GT data D1 of non-equilibrium state so as to correct the simulated core power distribution. In addition, the process module 60M2 of the CPU 60 may execute the adaptive simulation with the use of LPRM signals which will be described later in ninth and tenth embodiments, so as to correct the core power distribution. Further, it is possible to carry out the adaptive simulation with the use of GT signals predicted by a GT signal prediction function which will be described later in an eleventh embodiment so as to correct the core power distribution. The process module 60M2 of the CPU 60 may select either of the above correcting simulation processes. In the adjustment of at least one of the sensitivity and the gain of the LPRM signal, in the case where the adjustment is periodically and high frequently carried out (e.g., one time per hour in a state that the core state does not vary more than one hour), the data D1 to D3 captured via the nuclear instrumentation control process module 60M1 of the CPU 60 of the process control computer 31 is transmitted to the power distribution simulation module 60M2 by interface process of the process module 60M1. At this time, the power distribution simulation process module 60M2 simulates the second gain adjustment factor RGAF.sup.L by which the LPRM data D2 transmitted from the LPRM signal processing unit 40 is multiplied, so as to store the second gain adjustment factor RGAF.sup.L in the memory unit 61 without transmitting the second gain adjustment factor RGAF.sup.L to the LPRM signal processing unit 40. Therefore, in the LPRM signal processing unit 40, no sensitivity or gain adjustment process is carried out with the use of the newly first gain adjustment factor; on the other hand, in the power distribution simulation process module 60M2, the adjustment of at least one of the sensitivity and the gain of the LPRM detector 37 is carried out on the basis of the second gain adjustment factor RGAF.sup.L. Then, after the core power distribution greatly varies, and for example, one hour or more elapses, and thereafter, in the case where the response simulated value of the LPRM signal by the three-dimensional nuclear thermal-hydraulics simulation is shifted from the actually measured value of the LPRM data D2 processed by the LPRM signal processing unit 40 more than a predetermined ratio (e.g., 20%), or only in the case where a predetermined time, e.g., 1000 hours or more elapses, the second gain adjustment factor RGAF.sup.L is transmitted from the CPU 60 of the process control computer 31 to the LPRM signal processing unit 40 so that newly first gain adjustment factor GAF.sup.L.sub.n is obtained by multiplying the last first gain adjustment factor GAF.sup.L.sub.n-1 by the transmitted second gain adjustment factor RGAF.sup.L. Therefore, the adjustment of at least one of the sensitivity and the gain of the LPRM data D2 is carried out by the LPRM signal processing unit 40 with the use of the newly first gain adjustment factor GAF.sup.L.sub.n. Then, the following method is considered; more specifically, at the point of time when the adjustment of at least one of the sensitivity and the gain is carried out with the use of the first gain adjustment factor GAF.sup.L.sub.n the second gain adjustment factor RGAF.sup.L is zero-cleared to 1.0 (means that the read LPRM data D2 is used as it is). By doing so, it is possible to reduce the adjustment frequency of at least one of the sensitivity and the gain of the LPRM data D2 in the LPRM signal processing unit 40, to reduce a bypass time of the LPRM assembly 34 which is a part of the safety protection system with respect to the reactor, and to execute an adjustment of the safety protection system under surveillance and control by the operator. According to this seventh embodiment, in the in-core fixed nuclear instrumentation system 30, a plurality of fixed neutron detectors (LPRM detectors) 37 for detecting a local power distribution of power range in a reactor core and fixed GT detectors 44 gamma-ray thermometer assembly 35 for detecting a .gamma.-ray heating value, are housed in the nuclear instrumentation tube 33. Further, the in-core fixed nuclear instrumentation system 30 includes: the in-core nuclear instrumentation assembly 32 in which the GT detectors 44 are arranged at least in the vicinity of the fixed LPRM detectors 37; the LPRM signal processing unit 40 for processing the LPRM signal S2 from the LPRM detector 37; the GT signal processing unit 48 for processing the output voltage signal S1 from the gamma-ray thermometer assembly 35; the GT heater control unit 53 for carrying out an electrically energizing control with respect to the heaters 71 built in the GT assembly 35; the core state data measuring device 55 for detecting core state data indicative of the core states such as a reactor in-core power level, core coolant flow rate, control rod pattern or the like; and the nuclear instrumentation control process module 60M1 for simulating and storing the in-core elapse time or in-core irradiation burn-up of the GT assembly 35. The nuclear instrumentation control process module 60M1 receives the core states such as a reactor in-core power level, core coolant flow rate, control rod pattern, etc., which are outputted from the core state data measuring device 55 and are processed by the core state data processing unit 58, and then, makes a decision whether or not the predetermined time elapses after detecting the core state change in accordance with the core state data while outputting the result to the display unit 63 so as to inform it of the operator as a warning. Further, the nuclear instrumentation control process module 60M1 outputs the result to the display unit 63 so as to inform it of the operator. Moreover, the power distribution simulation process module 60M2 learns and simulates a core power distribution by the GT data (gamma heating value) computed based on the GT signal detected by the fixed GT detector 44 and the three-dimensional nuclear thermal-hydraulics simulation model, and simulates a reading value (simulated value) of each LPRM detector 37 from the core power distribution. Further, the power distribution simulation process module 60M2 compares the reading value with the actually reading value at present (actual detected value), and thus, carries out an adjustment of at least one of the sensitivity and the gain of each LPRM detector 37. In the case where the conventional nuclear instrumentation system uses the TIP, when using the LPRM detector signal D2 as auxiliary means of the power distribution monitoring system, there is a difference between a correlation of fuel nodal power around the GT detection portion of the GT signal D1 and a correlation of fuel nodal power around the LPRM detection portion of the LPRM signal D2. In particular, it is found that the LPRM detection signal D2 strongly depends upon an output of a corner fuel rod on the nuclear instrumentation tube 33 side of the fuel assembly 4, as compared with the GT signal D1. Therefore, by operating the fuel rod 5, in the case where a power distribution on a cross section of the fuel assembly 4 adjacent to the nuclear instrumentation tube 33 is greatly different between the control rod side and the nuclear instrumentation tube side, the GT signal change and the LPRM signal change have no proportional relationship. However, in the sixth and seventh embodiments, a behavior of the LPRM detection signal is proportional to a thermal neutron flux in the actual LPRM detection portion; therefore, very fast response is noticeable. For this reason, in the power distribution after operating the control rod 5, even if the GT signal becomes an equilibrium state, the LPRM signal always coincides with a thermal neutron level at a position of the nuclear instrumentation tube. Thus, a power distribution learning simulation is carried out on the basis of the LPRM signals, and thereby, there is a merit such that no delay of response is caused with respect to a change of local power. This embodiment will be described hereinafter in ninth and tenth embodiments detailedly. [Eighth embodiment] The following is a description on an eighth embodiment of an in-core fixed nuclear instrumentation system, a power distribution simulating system and a power distribution monitoring system of the present invention. This eighth embodiment shows a modification of the reactor power distribution monitoring system 29, the in-core fixed nuclear instrumentation system 30 and the power distribution simulating system 31, which have been described in the sixth and seventh embodiments. In the eighth embodiment, a method for adjusting at least one of the sensitivity and the gain of the LPRM detector 37 is different from the above sixth and seventh embodiments, and other construction is the substantially same as those; therefore, the details are omitted. The LPRM signal processing unit 40 has a function of adjusting (calibrating) at least one of the sensitivity and the gain of each LPRM detector 37. The adjusting method based on the adjusting function of the LPRM detector 37 includes tow kinds, that is, a first adjusting method and a second adjusting method. In the case of adjusting at least one of the sensitivity and the gain of each LPRM detector 37, the first adjusting method by the LPRM signal processing unit 40 comprises the following steps of: dividing the large number of LPRM detectors 37 into a plurality of APRM channels or LPRM groups (there is an LPRM detector which is not captured in the APRM channels depending upon a design); automatically selecting a predetermined LPRM detector 37 from each APRM channel or each LPRM group by an instruction signal from the nuclear instrumentation control process module 60M1 according to the maximum bypass number condition of the LPRM detector allowable in the operation each LPRM detector 37 belonging to each APRM channel or each LPRM group so as to change into a bypass state (bypass mode); adjusting at least one of the sensitivity and the gain of the selected LPRM detector 37 of the bypass mode; and returning the LPRM detector 37 of the bypass mode after being adjusted to a normal mode. Therefore, the adjustment of at least one of the sensitivity and the gain of each LPRM detector 37 is almost simultaneously carried out in the LPRM signal processing unit 40 in accordance with the instruction from the nuclear instrumentation control process module 60M1 by the number of "{number of APRM channels (or number of LPRM groups).times.(number of maximum allowable LPRM detector bypasses)}", and is carried out without the bypass of each APRM channel or each LPRM group itself. The LPRM adjustment is carried out with respect to all LPRM detectors 37, and thereafter, the gain adjustment of each APRM channel need to be carried out by way of precaution. In order to make a confirmation, the CPU 60 of the process control computer 31 automatically carries out a simulation for making a comparison between the APRM signal instruction with a thermal output computed from a heat balance of an atomic power plant. In the case where a difference by the comparative computation is more than a preset value, the CPU 60 of the process control computer 31 outputs a warning to the display unit 63 so as to inform it of the operator. As a result, there is no bypass of the APRM channels or the LPRM groups in which the LPRM gain or sensitivity adjustment is executed, and the adjustment of at least one of the sensitivity and the gain of each LPRM detector is carried out for a short time. In the case of adjusting at least one of the sensitivity and the gain of each LPRM detector 37, the second adjusting method by the LPRM signal processing unit 40 comprises the following steps of: selecting and bypassing one APRM channel (one LPRM group) from each APRM channel or each LPRM group (there is an LPRM detector which is not captured in the APRM channel depending upon a design) according to the instruction from the nuclear instrumentation control process module 60M1; changing all LPRM detectors 37 belonging to the one bypassed APRM channel (LPRM group) into a bypass state (bypass mode); and adjusting at least one of the sensitivity and the gain of each LPRM detector 37 of the bypass mode. Therefore, at least one of the sensitivity and the gain of the LPRM detector 37 is almost simultaneously adjusted by the LPRM signal processing unit 40 according to the instruction from the nuclear instrumentation control process module 60M1 by the number of LPRM detectors included in one or the bypass allowable maximum number of APRM channels or LPRM groups. When the adjustment of at least one of the sensitivity and the gain of the LPRM detector 37 is completed, the LPRM detector 37 and the APRM channel (LPRM group) of the bypass mode is returned to a normal mode. When the LPRM detector 37 and the APRM channel (LPRM group of the bypass mode is returned to the normal mode, almost simultaneously, the adjustment of at least one of the sensitivity and the gain of each APRM channel need to be carried out by way of precaution. In order to make a confirmation, the process control computer 31 automatically carries out a simulation for making a comparison between the APRM signal instruction with and a thermal output computed from a heat balance of an atomic power plant. In the case where a difference by the comparative computation is more than a preset value, the process control computer 31 outputs a warning to the display system 63 so as to inform it of the operator. Then, when the adjustment of at least one of the sensitivity and the gain of all LPRM detectors constituting one APRM channel or one LPRM group is completed, the LPRM signal processing unit 40 starts an adjustment of another APRM channel (or another LPRM group) according to the instruction from the nuclear instrumentation control process module 60M1. As a result, during the adjustment of at least one of the sensitivity and the gain of the LPRM detector, one or the maximum number of bypass allowable APRM channels (or LPRM groups) is simultaneously bypassed. However, even if a fault is caused in the nuclear instrumentation control process module 60M1, the bypass mode of the APRM of the safety protection system and the bypass mode of the LPRM detectors 37 are only specified bypass allowable APRM channels (or specified LPRM groups). Thus, a reliability of the safety protection system is superior to the first adjusting method. The LPRM detectors 37 belonging to one APRM have different core ordinates, and disperse in the core axial direction. Four LPRM detectors which exist in the specified nuclear instrumentation tube 33 belong to substantially different APRM channels, respectively. Therefore, the LPRM detector 37 belonging to one APRM channel is automatically selected and adjusted from a plurality of LPRM detectors and not manually selected by the operator, whereby there is an advantage such that a mistake in the manual selection can be prevented. By using the gamma-ray thermometer, at least one of the sensitivity and the gain of the LPRM detector is high frequently carried out, and in the case of carrying out the adjustment, the aforesaid atomization and consideration to a safety are required. [Ninth embodiment] The following is a description on a ninth embodiment of an in-core fixed nuclear instrumentation system, a power distribution simulating system and a power distribution monitoring system of the present invention. This ninth embodiment has the basically same structure as the reactor power distribution monitoring system 29, the in-core fixed nuclear instrumentation system 30 and the power distribution simulating system 31, which have been described in the first embodiment (see FIG. 1 to FIG. 11); therefore, the details are omitted. In the reactor power distribution monitoring system 29, the in-core fixed nuclear instrumentation system 30 and the reactor power distribution simulating system 31 of this ninth embodiment, the core state data S3 representing the parameter of the core state (operating mode) such as a core power level, core coolant flow rate and control rod pattern, detected by the reactor core state data measuring device 55, is inputted as the digital core state data D3 to the nuclear instrumentation control process module 60M1 of the CPU 60 of the process control computer 31 via the core state data processing unit 58. The nuclear instrumentation control process module 60M1 detects a change of the core state in accordance with the inputted core state data D3 so as to make a decision whether or not a predetermined time elapses, thereby displaying the result on the display unit 63 so as to inform the result of the operator. Then, on the basis of the result displayed on the display unit 63, in the case where the operator makes a decision that the operating mode does not satisfy a required condition, the operator outputs an simulation instruction to the power distribution simulation process module 60M2 of the CPU 60 via the nuclear instrumentation control process module 60M1 thereof by operating the input console 62. At this time, the power distribution simulation process module 60M2 of the CPU 60 adapts and corrects a three-dimensional power distribution on the basis of the latest nuclear instrumentation information (core state data; core state, operating mode) obtained from the core 3. The above embodiments described so far have described the case where the core state is a steady state, and when the steady state continues for a sufficiently long period, the core power distribution simulating processing is carried out by adaptive simulation using the three-dimensional simulation model and the GT data of the GT assembly. In this ninth embodiment, the following is a description on a process in place of the processing means of simulating the core power distribution just after the parameter representing the operating state such as power distribution or power level or the like changes. Just after the parameter representing the operating state such as power distribution or power level or the like changes, it is determined that the GT signal of the GT assembly 35 does not reach an equilibrium state. In such a case, by using the LPRM detector assembly 34 having a fast response in the GT assembly 35 and the LPRM detector assembly 34 constituting the in-core nuclear instrumentation assembly 32, there is employed a first adaptive correction process of correcting a power distribution obtained by the simulation based on the three-dimensional simulation model. More specifically, according to the first adaptive correction process, the memory unit 61 of the process control computer 31 stores interpolation and extrapolation approximation data (data set) on the basis of correlation parameters for showing a correlation between a fuel assembly nodal power value and an LPRM data reading value (simulated value corresponding to the actually measured LPRM data D2) {e.g., a fuel type, a node burn-up, presence of control rod, a historical relative water density (historical void fraction), an instantaneous relative water density (instantaneous void fraction)}, or an interpolation and extrapolation look up table data (data set) based on the above correlation parameter. According to the first adaptive correction process, for example, in the case where it is determined by the GT signal processing unit 48 or the nuclear instrumentation control process module 60M1 of the process control computer 31 that the GT signal of the GT assembly 35 does not reach an equilibrium state, when carrying out the adaptive correction, the power distribution simulation process module 60M2 simulates a core power distribution at the present point of time on the basis of the following data. More specifically, the data includes: core power distribution data stored in the memory obtained, in a state (equilibrium state) that the latest steady state retroactive from the point of time of adaptive correction execution (present point of time) requests, by the adaptive correction based on the GT signal S1 outputted from the GT assembly 35 at that time; adaptive correction data calculated and stored value for each fuel assembly node; and change data of the operating parameter (core state data D3) representing the core state (operating mode) such as an increase of core average burn-up until the present point of time (the point of time of adaptive correction executive processing) from the power distribution simulation in the above equilibrium state, control rod pattern from the latest point of time, core coolant flow rate, core power, core inlet enthalpy, core pressure or the like. Then, the power distribution simulation process module 60M2 determines LPRM prediction values corresponding to the simulated power distribution at the present point according to the values of the correlation parameters {e.g., a fuel type, a node burn-up, presence of control rod, a historical relative water density (historical void fraction), an instantaneous relative water density (instantaneous void fraction)} and the approximation equation (look up table) stored in the memory unit 61. On the other hand, the LPRM data D2 adjusted by the LPRM signal processing unit 48 according to the process described in the above seventh embodiment, is gathered in the power distribution simulation process module 60M2 of the CPU 60. Then, the power distribution simulation process module 60M2, as shown in FIG. 18, makes a comparison between the prediction value of the LPRM data based on the simulation result at the present point of time and the actually measured LPRM data D2 so as to obtain a correction ratio representing the difference between the LPRM data prediction value and the actually measured LPRM data D2 (step S90), thereby interpolating and extrapolating the correction ratio in the core axial direction, so that it is possible to obtain a correction ratio (additional adaptive correction quantity; relative adaptive correction quantity) with respect to all axial nodes (step S91). Next, the power distribution simulation module 60M2 corrects the in-core power distribution for each axial node which is simulated at the present point of time on the basis of the determined additional adaptive correction quantity of all axial node (step S92). Subsequently, the power distribution simulation module 60M2 evaluates the maximum linear heat generating ratio (MLHGR) and the minimum critical power ratio (MCPR) at the present point of time (at the point of adaptive correction execution process time) on the basis of the corrected in-core power distribution of each axial node (step S93). Next, the power distribution simulation process module 60M2 stores the additional adaptive correction quantity of all axial nodes based on the aforesaid LPRM data D2 in the memory unit 61, independently from the adaptive correction quantity of all axial nodes based on the GT data D1 stored therein (step S94). In the above manner, in the case of carrying out power distribution adaptive correction in the GT signal non-equilibrium state, the in-core power distribution simulated at the point of power distribution adaptive correction process time is corrected on the basis of the adaptive correction in the GT data D1 equilibrium state and the additional adaptive correction quantity based on the LPRM data, whereby it is possible to generate a core power distribution having a high accuracy. Then, in the case where it is determined by the GT signal processing unit 48 or the nuclear instrumentation control process module 60M1 of the process control computer 31 that the GT data D1 of the GT assembly 35 reaches the equilibrium state, the power distribution simulation process module 60M2 zero-clears the additional adaptive correction quantity based on the LPRM data D2 stored in the memory unit 61 {means that the additionally relative adaptive correction ratio (correction coefficient) based on the LPRM data D2 is returned to 1.0}. As shown in FIG. 13 and FIG. 14, again an adaptive correction quantity based on only the GT data D1 in an equilibrium state is obtained and stored in the memory unit 61, and then, a power distribution adaptive correction process is carried out on the basis of the obtained adaptive correction quantity. Thereby, even in the case where the GT data from the GT assembly 35 varies in a non-equilibrium state, the power distribution simulation process module 60M2 can execute a power distribution adaptive correction simulation based on in-core nuclear instrumentation. Therefore, it is possible to carry out, while the GT data lies in a non-equilibrium state, the in-core power distribution adaptive correction without waiting until the GT data D1 reaches the equilibrium state, thereby periodically or always executing the in-core power distribution adaptive correction process. According to this ninth embodiment, in the point of power distribution adaptive correction time of non-equilibrium, the power distribution adaptive correction simulation has been carried out with the use of the interpolated and extrapolated additional adaptive correction obtained according to the LPRM data D2 detected by four LPRM detectors 37 arranged along the core axial direction. When the GT data reaches the equilibrium state, the additionally relative adaptive correction quantity having the possibility including an error in interpolation and extrapolation is zero-cleared at a timing of power distribution adaptive correction process based on the GT data, and thereafter, the power distribution adaptive correction process is carried out on the basis of the GT data D1 of the equilibrium state detected by the large number of GT detectors 35 arranged along the axial direction. Therefore, it is possible to solve the problem that an error of the LPRM assembly 34 gives an influence to the in-core power distribution simulation for a long period. Thus, since the simulation depends upon only error of the GT assembly 35, it is possible to considerably improve an accuracy of evaluation of the maximum linear heat generating ratio (MLHGR) and the minimum critical power ratio (MCPR). [Tenth embodiment] The following is a description on a tenth embodiment of an in-core fixed nuclear instrumentation system, a power distribution simulating system and a power distribution monitoring system of the present invention. In the in-core fixed nuclear instrumentation system 29, the in-core nuclear instrumentation system 30 and the power distribution simulating system 31 of this tenth embodiment, similarly to the structure described in the ninth embodiment, it is determined that the GT data of the GT assembly 35 does not reach an equilibrium state. In such a case, of the GT assembly 35 constituting the in-core nuclear instrumentation assembly 32 and the LPRM detector assembly 34, with the use of the LPRM detector assembly 34 having a fast response, there is employed a second adaptive correction process of adapting and correcting a power distribution obtained by the simulation based on the three-dimensional simulation model. The second adaptive correction process is based on the process of directly adjusting the LPRM signal S2 detected by the LPRM detector 37 with the use of the GT data D1 (W/g) converted into a gamma-ray heating value based on the GT signal S1 of the GT assembly 35. According to the second adaptive correction process, at a given point of time after the directly adjustment process of at least one of the sensitivity and the gain of the LPRM signal S2 is completed with the use of the GT data by the LPRM signal processing unit 40 of the LPRM signal S2 of the LPRM detector 37, the power distribution simulation process module 60M2 simulates a core power distribution at the present given point of time on the basis of the following data. That is, the data includes: core power distribution data stored in the memory unit 61 obtained by the adaptive correction process based on the GT data D1 outputted from the GT assembly 35 in the latest equilibrium state before the present given point of time; adaptive correction data from a simulation prediction power value for each fuel assembly node by the GT data D1; and change data of the operating parameter (core state data D3) representing the core state (operating mode) until the present point of time from the power distribution simulation in the above equilibrium state. In this case, the simulation process of the power distribution simulation process module 60M2 described in the ninth embodiment is different from this tenth embodiment in that, when the GT signal lies in the transient state, the LPRM signal is considered as varying the same proportional quantity as the GT signal. That is, when the LPRM signal is adjusted at a relatively high frequency, for example, one time per day or per hour so as to coincide with the W/g signal level of the GT data D1, even in the case where the power change is caused in the local in the core 3 or the whole thereof, the change of each fuel assembly node by burn-up is very small. As a result, it is possible to make very small a power peaking change by the burn-up of the fuel rod 5 on the corner portion of the nuclear instrumentation tube side strongly given the influence to the detection signal level of the LPRM detector 37; therefore, the change can be disregarded. Accordingly, an average power change of the fuel assembly node is reflected in the GT signal level detected by the GT detector 44 and the LPRM signal level detected by the LPRM detector 37 at the same change rate. However, the control rod 5 is inserted or drawn out; for this reason, a change is generated in the control rod state or the fuel void fraction in the fuel channel. In this case, a great change is generated between a response change of the GT signal and a response change of the LPRM signal. As described in the first embodiment, in the memory unit 61, at least one of approximate expression data (data set) according to correlation parameters representing the correlation between the output values of the nodes of the fuel assembly 4 and the GT data values D1 based on the GT signals S1 and interpolation and extrapolation lookup table data (data set) according to the above correlation parameters is stored, wherein the correlation parameters includes, for example, a fuel type, a node burn-up, presence of control rod, a historical relative water density (historical void fraction), an instantaneous relative water density (instantaneous void fraction). In view of the aforesaid backgrounds, in this tenth embodiment, the memory unit 61 of the process control computer 31 stores first interpolation and extrapolation approximation equation data (look up table data) based on a correlation parameter for representing a correlation between the fuel assembly nodal power value and the LPRM data based on the LPRM signal, or only a second interpolation and extrapolation approximation data (look up table data) for the LPRM data, which is described hereinafter. In the case where the LPRM signal S2 of the LPRM detector 37 is high frequently adjusted by means of the LPRM signal processing unit 40 with the use of the GT data D1, when the control rod state or a void fraction of the channel changes so that an instantaneous or rapid transient phenomenon of the LPRM data D2 is generated, the second approximation data (look up table data), in order to simulate a different rate between an LPRM response change quantity of the LPRM data D2 and a GT response change quantity of the GT data D1 (change quantity in the case where the GT data D1 instantaneously shows a value of equilibrium state, is stored in the memory unit 61 of the process control computer 31 as interpolation and extrapolation approximation data (data set) or look up table data (data set) each of which is based on a correlation parameter for representing a correlation between the LPRM response change quantity of the LPRM data and the GT response change quantity of the GT data {e.g., a fuel type, a node burn-up, presence of control rod, a historical relative water density (historical void fraction), an instantaneous relative water density (instantaneous void fraction)}. Namely, the second approximation data (look up table data) for the LPRM detector is used in the process of the power distribution simulation process module 60M2, as described hereinafter. That is, as shown in FIG. 19, in a state that it is determined by the GT signal processing unit 48 or the nuclear instrumentation control process module 60M1 that the GT data of the GT assembly 35 is changed into a transient state, in response to a selection instruction command of a power distribution adapting mode outputted from the input console by the operator, or a change instruction command of the power distribution adaptive mode automatically outputted from the nuclear instrumentation control process module 60M1, the power distribution simulation process module 60M2 executes the core power distribution calculation on the basis of the operational parameters (core state data) D3 at the present point of time in which the GT data D1 lies in the non-equilibrium state but the power distribution adaptive simulation is instructed with the LPRM data being calibrated, by using the core power distribution data (first data) in a state (equilibrium state) that the latest steady state retroactive from the present point of time, the adaptive correction data (second data) and the change data (third data) of the operating parameter (core state data D3) representing the core state (operating mode) until the present point of time (the point of time of adaptive correction executive processing) from the power distribution simulation in the above equilibrium state (step S100). Next, the process module 60M2 receives the LPRM data D2 at the present point of time and obtains the parameters related to the second approximation data (look up data) according to the power distribution at the present point of time so as to convert the first response change quantity of the LPRM data D2 into the pseudo GT response of the second response change quantity thereof corresponding to the equilibrium value of the GT data (step S101). Then, the power distribution simulation process module 60M2 replaces the four LPRM data D2 of each LPRM detector with the converted second response change quantity in the axial direction as pseudo GT data reaching the equilibrium state at the present point of time (step S102). The power distribution simulation process module 60M2 makes a comparison between the above pseudo GT data value of the predetermined nodes at which the LPRM detectors are positioned and equilibrium GT data of the predetermined nodes in equilibrium GT data value (simulated value) of 24 nodes obtained in step S100 (step S103). Furthermore, the power distribution simulation module 60M2 interpolates and extrapolates a correction ratio, which is obtained from the comparison process, referred to step S103, showing a difference between the pseudo GT data value and the equilibrium GT data value (simulated value) in the core axial direction so as to obtain a correction ratio (additional adaptive correction quantity; relative adaptive correction quantity) with respect to all axial nodes (24 nodes) (step S104). The power distribution simulation process module 60M2 corrects an in-core power distribution of each fuel assembly node of the simulation result at the present point of time on the basis of the obtained additional adaptive correction quantity (correction ratio) of all axial nodes (step S105) so as to evaluate the maximum linear heat generating ratio (MLHGR) and the minimum critical power ratio (MCPR) at the present point of time (step S106). Moreover, the power distribution simulation process module 60M2 stores the additional adaptive correction quantity of all axial nodes based on the aforesaid pseudo GT data in the memory unit 61, independently from the adaptive correction based on the GT data D1 of all axial nodes stored therein (step S107). In the above manner, when the GT data is in non-equilibrium state and the signal level of them reaches in the transient state, in the case of carrying out power distribution adaptive correction process, the in-core power distribution simulated at the point of power distribution adaptive correction process execution time is learn and corrected on the basis of the adaptive correction in the GT data D1 which is in the equilibrium state and the additional adaptive correction based on the pseudo GT data estimated as the equilibrium value of the transient state obtained from the LPRM data, so that it is possible to generate a power distribution having a high accuracy. Then, in the case where it is determined by the GT signal processing unit 48 or the nuclear instrumentation control process module 60M1 of the process control computer 31 that the GT data D1 of the GT assembly 35 reaches the equilibrium state, the power distribution simulation process module 60M2 zero-clears the additional adaptive correction quantity based on the LPRM data D2 stored in the memory unit 61 {means that the additionally relative adaptive correction ratio (correction coefficient) based on the LPRM data D2 is returned to 1.0}. As shown in FIG. 13 and FIG. 14, again an adaptive correction quantity based on only the GT data D1 in an equilibrium state is obtained and stored in the memory unit 61, and then, a power distribution adaptive correction process is carried out on the basis of the obtained adaptive correction quantity. Thereby, even in the case where the GT data from the GT assembly 35 varies in a non-equilibrium state, the power distribution simulation process module 60M2 can execute a power distribution adaptive correction simulation based on in-core nuclear instrumentation. Therefore, it is possible to carry out, while the GT data lies in a non-equilibrium state, the in-core power distribution adaptive correction without waiting until the GT data D1 reaches the equilibrium state, thereby periodically or always executing the in-core power distribution adaptive correction process. According to this tenth embodiment, in the point of power distribution adaptive correction time of non-equilibrium, the power distribution adaptive correction simulation has been carried out with the use of the interpolated and extrapolated additional adaptive correction obtained according to the LPRM data D2 detected by four LPRM detectors 37 arranged along the core axial direction. When the GT data reaches the equilibrium state, the additionally relative adaptive correction quantity having the possibility including an error in interpolation and extrapolation is zero-cleared at a timing of power distribution adaptive correction process based on the GT data, and thereafter, the power distribution adaptive correction process is carried out on the basis of the GT data D1 of the equilibrium state detected by the large number of GT detectors 35 arranged along the axial direction. Therefore, it is possible to solve the problem that an error of the LPRM assembly 34 gives an influence to the in-core power distribution simulation for a long period. Thus, since the simulation depends upon only error of the GT assembly 35, it is possible to considerably improve an accuracy of evaluation of the maximum linear heat generating ratio (MLHGR) and the minimum critical power ratio (MCPR). In the ninth and tenth embodiments, in the case where the GT data is in a non-equilibrium state, the in-core power distribution has been simulated with the use of the LPRM data. In the non-equilibrium state, in the case where the LPRM detector 37 is bypassed due to a failure, it is possible to disregard the LPRM data of the bypassed LPRM detector 37 so as to preferentially gather the simulation value, or to use the LPRM data on a nuclear instrumentation tube position having a symmetry based on core fuel loading control rod pattern, in place of the bypassed LPRM data. [Eleventh embodiment] The following is a description on an eleventh embodiment of an in-core fixed nuclear instrumentation system, a power distribution simulating system and a power distribution monitoring system of the present invention. This eleventh embodiment has the basically same structure as the reactor power distribution monitoring system 29, the in-core fixed nuclear instrumentation system 30 and the power distribution simulating system 31, which have been described in the first embodiment (see FIG. 1 to FIG. 11); therefore, the details are omitted. In the power distribution monitoring system 29, the in-core fixed nuclear instrumentation system 30 and the power distribution simulating system of this eleventh embodiment, the number of GT detectors 44 in the axial direction is larger than that of LPRM detectors 37. By using this advantage, it is possible to carry out an adaptive correction of power distribution with the use of only GT data D1 even in the case where the GT data D1 outputted from the GT detector 44 is in a transient state. More specifically, in the reactor power distribution monitoring system 29 described in the ninth and tenth embodiments, the GT data D1 outputted from the GT detector 44 is in a state of not reaching an equilibrium level of gamma decay, that is, after the operating parameter (core state data D3) including a core power, core coolant flow rate, control rod pattern varies, for examples within a short time such as an hour, a level of the GT data D1 (W/g) based on the GT signal S1 varies in minutes. Then, when the power distribution simulation process module 60M2 corrects the power distribution on the basis of the GT data D1, due to the non-equilibrium state of the GT data D1 is non-equilibrium state, the local power is over-estimated (in the case where the local power lowers) or is underestimated (in the case where the local power increases). For this reason, even if the power distribution simulation process module 60M2 corrects the power distribution, the correction result of the power distribution includes an error. Thus, the response of the LPRM detector 37 having a fast response has been used as an auxiliary means (see ninth and tenth embodiments). However, four LPRM detectors 37 are only arranged in the axial direction; for this reason, there is the possibility that an accuracy is deteriorated in view of the adaptive correction of the axial power distribution. To solve the above problem, the in the reactor power distribution monitoring system 29, the in-core fixed nuclear instrumentation system 30 and the power distribution simulating system 31, the core state parameter signal (core state data D3) including a core power level, a core coolant flow rate, a control rod pattern and so on, which is outputted from the reactor core state data processing unit 58, is inputted to the power distribution simulation process module 60M2 via the nuclear instrumentation control process module 60M1 of the CPU 60 of the process control computer 31. Then, the power distribution simulation process module 60M2 periodically (every predetermined time) or always simulates an in-core power distribution on the basis of the inputted core state data D3 and the GT data D1. Namely, as shown in FIG. 20, the nuclear instrumentation control process module 60M1 of the CPU 60 initializes elapse time counter (step S110A), and counts the elapse time (step S110B). Next, the process module 60M1 judges whether or not the core state change is detected by the core state data (step S110C). In the case where the core state change is detected, that is, the judgement of step S110C is YES, the process module 60M1 returns to a process of step S110A. In the case where the core state change is not detected, that is, the judgement of step S110C is NO, the process module 60M1 judges whether or not the LPRM detector gain adjustment instruction is requested by the input console 62 (step In the case where the LPRM detector gain adjustment instruction is not requested, that is, the judgement of step S110D is NO, the process module 60M1 returns to a process of step S110B. On the other hand, in the case where the LPRM detector gain adjustment instruction is requested, that is, the judgement of step S110D is YES, the process module 60M1 Namely, as shown in FIG. 20, the process module 60M1 detects a change of the core state on the basis of the core state data D3 so as to make a judgement whether or not a predetermined time elapses after the change of the core state is detected (step S110E). Then, in the case where it is determined that the predetermined time elapses after detecting the change of the core state (the judgement of step S110E is YES), the process module 60M1 transmits the judgement result to the display unit 63 so as to inform it of the operator via the display unit 63 (step S111). Then, the operator confirms the judgement result displayed on the display unit 63 so as to transmit a confirmation instruction to the nuclear instrumentation control process module 60M1 of the CPU 60 of the process control computer 31 by operating the input console 62 (step 112). The nuclear instrumentation control process module 60M1 automatically makes a judgement whether or not the GT data D1 lies in an equilibrium state (step S113). When the judgement of step S113 is YES, that is, the GT data D1 lies in the equilibrium state, the power distribution simulation process 60M2 of the CPU 60 corrects the simulated core power distribution according to the GT data D1 in the equilibrium state (step S114). On the other hand, when the judgement of step S113 is NO, that is, the GT data D1 does not lie in the equilibrium state, the process module 60M1 transmits a simulation instruction for predicting the GT data D1 to the GT signal processing unit 48 (step S115). In accordance with the transmitted GT signal prediction simulation instruction, the CPU 48A of the processing unit 48 gathers the predetermined number of GT data D1 (W/g signal level) generated according to the varying GT signal S1 at each predetermined times, for example, at several tens of seconds (20 to 30 seconds) or at each minute, that is, captures several or 10 points of the GT data D1 so as to carry out least square approximation to the following equation (6) (step S116). EQU (a+.SIGMA.b.sub.i.multidot.e.sup.-.lambda.it) (6) The CPU 48A of the GT signal processing unit 48 approximates the GT data D1 to the above equation (6) so as to determine the constants (coefficients), and then, estimates (predicts) an equilibrium level of the GT data based on the GT signal S1 actually measured by the GT detector 44 after a required time, e.g., one hour or more elapses according to the above equation (6) on the basis of the determined coefficients, thereby transmitting the estimated first GT data (equivalent to the equilibrium value) to the power distribution simulation process module 60M2 of the CPU 60 of the process control computer 31 (step S117). In addition, the above estimation (prediction) process of the equilibrium level of the GT data according to the above least square approximation may be carried out by the process module 60M1 of the CPU 60 of the process control computer 31. On the other hand, the process module 60M2 simulates an in-core power distribution corresponding to the reactor operating mode on the basis of the core state data D3 representing the operating mode at the present point of time so as to obtain second GT data from the simulation result (step S118). Next, the process module 60M2 obtains data showing a difference (ratio) between the second GT data (equivalent to the equilibrium value) simulated by the above simulation and the first GT data estimated from the actually measured GT signal S1 (step S119). Subsequently, the process module 60M2 interpolates and extrapolates the difference data in each node arranged in the core axial direction so as to generate correction data (correction ratio data) with respect to all axial nodes, making it possible to obtain an adaptive correction quantity based on the GT signal of all axial nodes (step S120). Then, the process module 60M2 corrects an in-core power distribution for each fuel assembly node of the simulation result at the present point of time on the basis of the determined adaptive correction quantity of all axial nodes and then calculates the maximum linear heat generating ratio (MLHGR) and the minimum critical power ratio (MCPR) at the present point of time (at the point of adaptive correction execution processing time) on the basis of the corrected in-core power distribution (step S121). As a result, even in the case where the GT signal outputted from the GT detector 44 varies, a so-called reactor transient state, it is possible to monitor and evaluate a reactor power distribution. At this time, the GT signal processing unit 48 (or the nuclear instrumentation control process module 60M1) carries out the prediction simulation for estimating (predicting) the equilibrium signal level of the GT detector 44 after a required time, e.g., one hour or more elapses. The prediction simulation is automatically successively carried out for each new time (e.g., for each 20 to 30 seconds or for each minute) for the duration that an information on prediction function mode (power distribution adaptive mode by GT data estimation) selective instruction is transmitted via the nuclear instrumentation control process module 60M1 from the input console 62. More specifically, when the GT signal processing unit 48 (the nuclear instrumentation control process module 60M1) gathers the GT data D1 (W/g level signal) based on the new GT signal S1, the GT signal processing unit 48 cancels (deletes) the oldest data in time series, and updates the predictive GT data value (equilibrium value) from a time series GT data group having several or 10 data including the new GT data D1 according to the least square approximation based on the above equation (6), and thus, transmits the updated GT data value (equilibrium value) to the power distribution simulation process module 60M2. Then, the process module 60M2 carries out the in-core power distribution adaptive correction on the basis of the updated GT predictive simulation value. Furthermore, the process module 60M2 outputs an information indicative of a mode (power distribution adaptive correction by GT data estimation) of carrying out the power distribution adaptive correction with the use of the GT data equilibrium value to the display unit 63 so as to inform it of the operator. The GT detector 44 has a thermal time constant of an order of second unit, and a gamma-ray source contributing to heating of the GT detector 44 has a time constant distribution of a wide range from a time constant of emitting a gamma-ray substantially simultaneous with a nuclear fission or a short time such as an order of second to a time constant of an order of minute, time and day. A weight of component of each time constant depends upon a gamma source included in the fuel; however, the gamma source is different depending upon a fission nuclide (e.g., U235, Pu239, etc.) in a nuclear fission, and is different depending upon an elapse time after the nuclear fission. A nuclide enrichment is strictly treated in the three-dimensional BWR simulator; for this reason, a component of nuclear time constant is determined for each node of the fuel assembly, and as a result, it is not practical use because the nuclear time constant data library stored in the memory unit becomes large. So, in this eleventh embodiment, the time constant of the gamma source is limited to a time constant subjected to the gamma source at the point of time after a required time, e.g., one hour or a minute order elapses, and the number of data is limited to 10 or less or about 10. Then, the least square approximation is carried out so as to meet the following equation (7). EQU (a+.SIGMA.b.sub.i.multidot.e.sup.-.lambda.it) (7) Coefficients of a, bi (i=1 to about 10 at the maximum) is obtained from the above equation (7), and then, an equilibrium GT data value after a required time, e.g., one hour is estimated from the time series data of the GT data. A time constant (sec.sup.-1) of JNDC approximation equation shown in the table 1 obtained from the nuclear time constant library is selected as a value of .lambda.i in the above equation (7). This is one example, other time constants may be selected depending upon edit, and the number of data is reduced to 10 or less, and thereby, the GT detector has a further longer time constant (half life), and there is a method of omitting a low gamma source strength. For example, it is considered that a time constant of 10.sup.-5 (sec.sup.-1) is disregarded. TABLE 1 group .lambda.(sec.sup.-1) 1 1.330E-02 2 3.488E-02 3 1.357E-03 4 3.591E-03 5 5.004E-03 6 1.850E-04 7 5.645E-04 8 1.922E-05 9 4.918E-05 10 5.435E-05 According to this process, the GT signal processing unit 48 (the nuclear instrumentation control process module 60M1) selects the GT data D1 (W/g signal) based on the GT signal S1 outputted from the GT detector 44 for predetermined time, e.g., 30 seconds or one minute, and then, stores ten and several data in the memory unit 61 in time series. Further, the GT signal processing unit 48 (the nuclear instrumentation control process module 60M1) erases (deletes) the old time GT data, and successively updates and stores the new GT data in the memory unit 61. For example, the least square approximation of the following equation (8) is repeatedly carried out every 30 seconds or one minute. EQU (a+.SIGMA.b.sub.i.multidot.e.sup.-.lambda.it) (8) The GT prediction value after a predetermined time, e.g., about one hour is updated every least square approximation, and thereby, even if the GT signal S1 is in the non-equilibrium state, it is possible to generate the pseudo GT data value of the equilibrium state based on the GT signal outputted from the GT detector 44 every predetermined time, e.g., 30 second or one minute. Then, after about 5 to 15 minutes just after power change, it is possible to obtain the prediction equilibrium GT data value based on the GT signal outputted from the GT detector 44. A sampling interval of the time series data is practically selected according to an operation (processing) speed of the GT signal processing unit 48 or the process control computer 31; therefore, the sampling interval is not limited to 30 seconds or one minute. In this embodiment, a clearly limited time interval is selected so that the sampling interval is determined to 30 seconds or one minute. However, it is considered that several number of GT data or 10 data may be successively sampled at substantially each 1 second per 1 sampled data for filtering noise for example fluctuation noise included in the GT data. The sampling process is able to obtain the least square approximation having high accuracy. In addition, the number of data, which is stored in time series in the memory unit 61 and is fitted to the least square approximation, is not limited to about 10, but may be reduced to the number of data, that is, about 5 in view of a balance of a predictive accuracy and a time required for estimation. Namely, it is important to obtain a predictive value as fast as possible without decreasing the predictive accuracy. The following is a description on an operation of the in-core fixed nuclear instrumentation system 30 and the power distribution simulating system 31 with reference to FIG. 21 and FIG. 22. FIG. 21 is a graph showing the actually measured GT data value and the GT data prediction value with respect to the elapse time (minute) in the case where the power of fuel node around the detection portion of the GT detector 44 increases. FIG. 22 is a graph showing the actually measured GT data value and the GT data prediction value with respect to the elapse time (minute) in the case where the power of fuel node around the detection portion of the GT detector 44 conversely decreases. In FIG. 21 and FIG. 22, a solid line P shows the actual change of the core local power, a dotted line O shows the actually measured GT data (W/g signal) value, and a broken line Q shows the GT data prediction value. In any cases, the GT signal having the number of data after change, e.g., 10 is captured in time series, and then, is fitted to the least square approximation. By using the sum of the polynomial and the constant term of the obtained exponential function (see equations (6) to (8)), the equilibrium value can be accurately simulated. In particular, a decay time constant of gamma source contributing for the duration from a minute to hour is selected so as to reduce the number of polynomial, and thereby, it is possible to very effectively shorten the simulation time. In calculation of the least square approximation, in order to make convergence fast, a set of a and bi of the initial guess is prepared separately from the case of core power increase and the case of core power decrease. As a result, according to the aforesaid simple process, it is possible to readily estimate the gamma-ray heating value (GT equilibrium data value) in the equilibrium state of gamma decay without simulating a history before the point of time at which the transient change happens, which includes the power change of local power distribution and the whole of core, and a distribution of gamma source at the point of time, which are studied conventionally, so as to execute a complicate process for determining a gamma heating value after core power change according to the simulated result. Moreover, the power distribution monitoring system 29 is operated at a mode (power distribution adaptive mode by GT data prediction) for determining the gamma heating value (GT data value) by a predictive calculation, and then, the above information is displayed on the display unit 63 so as to inform it of the operator. Whereby it is possible to give the following caution to the operator, that is, a caution of including an error by the predictive calculation in LPRM adjustment and the power distribution simulation result already corrected. In this eleventh embodiment, the equilibrium state prediction simulation function of the GT data value used in the case where the GT signal is in the non-equilibrium state, and it is normal not to always use the function. The present invention is not limited to this. The following is a description on a modification of this eleventh embodiment. More specifically, the GT signal processing unit 48 always transmits the GT data (W/g signal using the prediction simulation function to the process control computer 31 (power distribution simulation process module 60M2) except heater calibration of the GT detector 44. Then, the CPU 60 of the process control computer 31 always captures the transmitted predicted GT data value, and uses the predicted GT data value for adapting correction of the power distribution, or for adjusting at least one of the LPRM sensitivity and gain in the LPRM signal processing unit 48. In this modification, the process of adjusting at least one of the LPRM sensitivity and the gain includes the following two processes: more specifically, (1) a process for adjusting at least one of the sensitivity and the gain of the LPRM detector so as to directly coincide with the GT equilibrium data simulation value; and (2) a process for adapting and simulating the in-core power distribution by the power distribution simulation process module 60M2 with the use of the GT equilibrium data simulation value, and adjusting at least one of the sensitivity and the gain of the LPRM detector so as to directly coincide with the LPRM signal (LPRM data) simulated with the use of the simulated result. At least one of the sensitivity and the gain of the LPRM detector 37 is carried out according to an operation command from the operator from the display unit 63 via the nuclear instrumentation control process module 60M1. Whereby the operator does not need to be anxious about whether the GT data level is in the non-equilibrium state or the equilibrium state, and power distribution adaptive simulation and LPRM gain and sensitivity adjustment are automatically carried out on the basis of the GT data of equilibrium level. Therefore, it is possible to readily use the in-core power distribution monitoring system 29, and to reduce a load work onto the operator. In this case, however, the GT signal processing unit 48 or the CPU 60 of the process control computer 31 always monitors an accuracy by the least square approximation, and in the case where the accuracy of the least square approximation is less than a predetermined accuracy, the CPU 60 outputs a warning to the display unit 63 so as to inform it of the operator. Then, the operator manually controls the power distribution adaptive simulation or at least one of the adjustment of the sensitivity and the gain of the LPRM using the above GT equilibrium data value via the input console 62. With the above construction, in the case where the accuracy of the least square approximation is less than the predetermined accuracy, the power distribution adaptive simulation or the adjustment of at least one of the sensitivity and the gain of the LPRM using the above GT equilibrium data value is stopped, and then, is changed into the power distribution adaptive simulation or the adjustment of at least one of the sensitivity and the gain of the LPRM based on the operator's manual instruction. Thus, it is possible to maintain a high reliability of the power distribution monitoring system. As described above, in this eleventh embodiment, even in a state that the GT signal detected by the GT detector does not reach the signal level of equilibrium state of the gamma decay chain, the response of the GT detector 44 is readily corrected so as to carry out the three-dimensional power distribution adaptive process. Therefore, with the use of only simply GT detection signal of the in-core fixed nuclear instrumentation system 30, it is possible to monitor the operational thermal limit value such as the maximum linear heat generating ratio (MLHGR) and the minimum critical power ratio (MCPR). The monitoring can be carried out within a practical time delay at an arbitrary point of time. While there has been described what is at present considered to be the preferred embodiments and modifications of the present invention. It will be understood that various modifications which are not described yet may be made therein, and it is intended to cover in the appended claims all such modifications as fall within the true spirit and scope of the invention.
	abstract	In a lithographic apparatus using exposure radiation of a relatively short wavelength, e.g. 157 or 126 nm, a laminar flow of N2 is provided across parts of the beam path in or adjacent to moving components of the apparatus. The laminar flow is faster than the maximum speed of the moving components and the diffusion rate of air thereby minimizing the contamination of the N2 by mixing with air. Laminar flow may be ensured by providing partitions to divide the beam path into separate spaces, by covering rough or non-planar surfaces in components on or adjacent to the laminar flow and by providing aerodynamic members.
	description	This application claims the benefit of, and priority to, U.S. Provisional Application No. 62/770,125, filed on Nov. 20, 2018, which is incorporated herein by reference in its entirety. This disclosure relates generally to mobile radiation oncology coach system and more specifically to a mobile radiation oncology coach system with internal and/or external shielding to a mobile unit. A medical linear particle accelerator (LINAC) is widely used to treat cancer by using customized high energy x-rays or electrons to conform to a tumor's shape of a patient and destroy cancer cells while sparing surrounding normal tissue of the patient. Like all expensive equipment, a LINAC with normal usage (e.g., 25 treatments per day) would require regular maintenances in addition to daily or weekly-based calibrations. Typically, a regular maintenance would require the LINAC to be shut down for a period of time that may takes weeks or months. In addition, when an upgrade for renovation or when new equipment installation is required, the LINAC is typically shut down. The shutdown of a single vault LINAC facility could cost a million dollars of revenue lost during a multi-month shutdown. The inventors here have recognized that there is a need for mobile and/or interim (e.g., portable, substantially portable/movable, leasing) radiation oncology service solutions that are capable of overcoming the foregoing shortcomings and maintain high-quality care, referrals and revenue while the fixed site equipment is temporarily unavailable or where fixed site equipment are not possible. Disclosed here are numerous aspects of a unique and advantageous mobile radiation oncology coach equipped with state-of-the-art LINAC facility that is able to provide the same or equivalent technology, such as accelerated treatment times, a six-point safety system, ergonomic operator controls and many patient-friendly features, that are typically offered to patients in leading cancer centers. While patients receive excellent clinical care, the user and/or owners of the mobile radiation oncology coach experience no disruption in referrals, revenue or staffing during equipment upgrades or construction projects. In some embodiments, a mobile radiation oncology coach system comprises a trailer configured to include a control console area, a treatment area, and a vestibule area located between the control console area and the treatment area, the treatment area being equipped with a medical treatment facility that can emit radiation; a first internal shielding provided between the vestibule area and the treatment area; a first door configured and providing access between the treatment area and the vestibule area, the first door including a first supplemental shielding; and a second door configured and providing access between the vestibule area and the control console area, the second door including second supplemental shielding and further configured to be constructed near an opposite side of said trailer, preventing a direct line of sight between the treatment area and the control console area. In some embodiments, the first internal shielding comprises interlocked lead bricks. In some embodiments, the interlocked lead bricks comprises a predetermined thickness to provide substantially effective shielding between the control console area and the treatment area. In some embodiments, the mobile radiation oncology coach system further comprises a second internal shielding provided between the vestibule area and the control console area. In some embodiments, the second internal shielding comprises additional interlocked lead bricks comprising a second thickness to provide substantially effective shielding between the control console area and the vestibule area. In some embodiments, the mobile radiation oncology coach system further comprises an alternating door containing interlocked lead bricks to shield direct line of sight of the medical treatment facility and people located in the control console area. In some embodiments, the medical treatment facility includes medical linear particle accelerator (LINAC). In some embodiments, the mobile radiation oncology coach system further comprises an external shielding, wherein the external shielding comprising a plurality of barriers. In some embodiments, the plurality of barriers are made of concrete. In some embodiments, the mobile radiation oncology coach system further comprises a support pad dimensioned to support the trailer, and wherein the support pad comprises concrete. In some embodiments, the mobile radiation oncology coach system further comprises a tractor, wherein said tractor and said trailer are arranged in tandem. In some embodiments, the first door is a pocket door that is driven by a motor which in turn is controlled by door switches. In some embodiments, the mobile radiation oncology coach system further comprises a lever for manually disengaging the pocket door and the motor. In some embodiments, a mobile radiation oncology coach system comprises a trailer configured to include a control console area and a treatment area, the treatment area being equipped with a medical treatment facility that can emit radiation; a first internal shielding disposed between the control console area and the treatment area; a first door configured and providing access between the treatment area and the control console area, the first door including a first supplemental shielding, wherein the first door is further configured to be constructed and positioned to prevent a direct line of sight between the treatment area and the control console area; and a swing door including second supplemental shielding, and constructed and positioned to shield radiation that may be emitted in an area associated with the first door between the treatment area and the control console area. In some embodiments, the first internal shielding comprises interlocked lead bricks. In some embodiments, the interlocked lead bricks comprises a predetermined thickness to provide substantially effective shielding between the control console area and the treatment area. In some embodiments, the mobile radiation oncology coach system further comprises a vestibule area located between the control console area and the treatment room; and a second internal shielding provided between the vestibule area and the control console area. In some embodiments, the second internal shielding comprises additional interlocked lead bricks comprising a second thickness to provide substantially effective shielding between the control console area and the vestibule area. In some embodiments, the medical treatment facility includes medical linear particle accelerator (LINAC). In some embodiments, the mobile radiation oncology coach system further comprises an external shielding, wherein the external shielding comprising a plurality of barriers. In some embodiments, the plurality of barriers are made of concrete. In some embodiments, the mobile radiation oncology coach system further comprises a support pad dimensioned to support the trailer, and wherein the support pad comprises concrete. In some embodiments, a mobile radiation oncology coach system comprises a trailer configured to include a control console area and a treatment area, the treatment area being equipped with a medical treatment facility that can emit radiation; internal shielding disposed between the control console area and the treatment area; and external shielding provided at a predetermined location outside of the trailer. In some embodiments, the internal shielding comprises interlocked lead bricks. In some embodiments, the interlocked lead bricks comprises a predetermined thickness to provide substantially effective shielding between the control console area and the treatment area. In some embodiments, the mobile radiation oncology coach system further comprises a vestibule area located between the control console area and the treatment room; and a second internal shielding provided between the vestibule area and the control console area. In some embodiments, the second internal shielding comprises additional interlocked lead bricks comprising a second thickness to provide substantially effective shielding between the control console area and the vestibule area. In some embodiments, the mobile radiation oncology coach system further comprises an alternating door between the treatment room and the control console area, wherein the alternating door contains interlocked lead bricks to shield direct line of sight of the medical treatment facility and people located in the control console area. In some embodiments, the mobile radiation oncology coach system further comprises a first door configured and providing access between the treatment area and the vestibule area, the first door including first shielding; and a second door configured and providing access between the vestibule area and the control console area, the second door is further configured to be constructed near an opposite side of said trailer, preventing a direct line of sight between the treatment area and the control console area. In some embodiments, the mobile radiation oncology coach system further comprises a first door configured and providing access between the treatment area and the control console area, the first door including first supplemental shielding. In some embodiments, the first door is further configured to be constructed and positioned to prevent a direct line of sight between the treatment area and the control console area. In some embodiments, the mobile radiation oncology coach system further comprises a swing door including a second supplemental shielding, and constructed and positioned to shield radiation that may be emitted in an area associated with the first door between the treatment area and the control console area. In some embodiments, the medical treatment facility includes medical linear particle accelerator (LINAC). In some embodiments, the external shielding comprising a plurality of barriers. In some embodiments, the plurality of barriers are made of concrete. In some embodiments, the mobile radiation oncology coach system further comprises a support pad dimensioned to support the trailer, and wherein the support pad comprises concrete. In some embodiments, the mobile radiation oncology coach system further comprises a tractor, and wherein said tractor and said trailer are arranged in tandem. In some embodiments, a method for providing a mobile radiation oncology services, the method comprises moving a trailer to a designated site, the trailer having a control console area and a treatment area being equipped with a medical treatment facility that can emit radiation; providing an internal shielding disposed between the control console area and the treatment area; and providing an external shielding at a predetermined location outside of the trailer. In some embodiments, the internal shielding comprising interlocked lead bricks. In some embodiments, the method further comprises providing an alternating door positioned between the treatment area and the control console area, wherein the alternating door contains interlocked lead bricks to take away direct line of sight of the medical treatment facility and people located in the control console area. In some embodiments, the medical treatment facility is a LINAC. In some embodiments, the external shielding comprising a plurality of barriers. In some embodiments, the plurality of barriers is made of concrete. In some embodiments, the method further comprises providing a support pad dimensioned to support the trailer, wherein the support pad is made of concrete. In some embodiments, the method further comprises providing a tractor, wherein the tractor and the trailer are arranged in tandem. In some embodiments, the method further comprises securing the trailer after the trailer is moved to the designated site. In some embodiments, the method further comprises removing the external shielding after the services is complete. It should be understood that each of the foregoing and various aspects, together with those set forth in the claims and summarized above and/or otherwise disclosed herein, including the drawings, may be combined to support claims for a device, apparatus, system, method of manufacture, and/or use without limitation. As summarized above and illustrated in the drawings, disclosed herein are various aspects and embodiments of a mobile radiation oncology coach system. According to some embodiments, the exemplary mobile radiation oncology coach system 10 described herein comprises a mobile unit 100, an external shielding 201-211 and an optional support pad 300. Referring to FIG. 1, the mobile unit 100 has a tractor 110 and a trailer or relocatable coach 120 arranged in tandem. The trailer 120 can be attached to the tractor 110 during relocation (travel mode). The trailer 120 is configured to house a LINAC facility. When the trailer 120 is in the treatment configuration or clinical mode, the trailer 120 can be detached from the tractor 110 so that the tractor 110 can be separated from the trailer 120 for other duties. In some embodiments, the mobile unit 100 can be a single motorized vehicle (e.g., a bus or a motorhome) instead of a separate tractor and trailer combination. In the exemplary embodiment, the mobile trailer 120 is about 58 feet in length and about 13 feet 6 inches in height. When the mobile trailer 120 is in the travel mode, the mobile trailer 120 is about 10 feet in width. In the exemplary embodiment, the mobile trailer 120 has slide-out sections that allow the width of the mobile trailer 120 to be extended to a wingspan of about 18 feet in width when the mobile unit 100 is in the clinical mode. In some embodiments, the dimensions of the mobile trailer 120 are varied, including the dimensions of the slide-out sections. Preferably, the mobile trailer 120 is provided with sufficient area to be maneuvered and positioned for setup and takedown. The mobile trailer 120 can be provided with external storage compartments and service doors that require access during processes or operation. The slide-out sections, patient lift, entry stair and any optional platform may require additional space on the passenger side of the mobile unit 100. In some embodiments, storage compartments, service doors, slide out sections, patient lift and/or platforms are provided in alternative configurations. Referring also to FIG. 2, proper and safe operation of the LINAC system can be obtained when the mobile trailer 120 is located on a substantially level area or firm pad 300. Hydraulic support legs 180 can be provided to assist in the leveling and stabilization of the mobile trailer 120. In some embodiments, load bearing screw jacks and support legs can ultimately replace the hydraulic support legs as long-term support. In some embodiments, the recommended mobile unit support pad 300 is a concrete pad with a dimension of, for example, 20 feet wide and 60 feet long. In some embodiments, the minimum support pad 300 could be split into two or more separate pads, rather than one large pad if properly configured. For example, a support pad 300 at the front 122 of the mobile trailer 120 can provide support for the landing gear, leveling legs and king pin support. A support pad at the rear 124 of the mobile trailer 120 can support tandem-axles (two sets of axles) 190, the hydraulic leveling legs and the load bearing screw jacks. Referring also to FIG. 12, in some embodiments, the thickness of support pad 300 is to be determined at the local level, based on, for example, soil conditions. In some embodiments, the levelness of support pad 300 is preferably not to exceed ¼ inches per 10 feet. In some embodiments, the overall length of the mobile unit 100 of the tractor and trailer tandem is generally 75 feet. The travel weight can be approximately 80,000 pounds. In some embodiments, an area, for example, one hundred sixty feet by 60 feet immediately adjacent to the support pad 300 on both sides is blocked off and reserved to allow for assembly, set-up, and upon conclusion of its use, dismantling of this unit. Access to this area from the adjacent roadway infrastructure preferably be available in all weather conditions, while taking into consideration the weight of the trailer and supporting vehicles. Exemplary electrical options are provided below. The electrical power for the mobile unit 100 can be 480 volt AC, 3-phase Wye system with neutral and ground, at 200 Amperes. The frequency can be 60+/−2 Hertz. The maximum voltage variance can be +11%/−4% from nominal voltage. The maximum line regulation can be 2.5%. The maximum line-to-line imbalance can be 3%. For power cord/plug, a Russell Stoll 200 Amp plug, can be supplied with the 50 feet power cord for connection to facility power. The cord connection point can be on the roadside of the trailer at around mid-point. For electrical support requited at the facility, a 200 Amp, 480 Volt, 5 wire dedicated service including, for example, a Russell Stoll 200 Amp receptacle, can be mounted in a NEMA 3R rated enclosure to meet local codes requirements. An auxiliary earth ground connection point may be required in addition to the ground circuit within the pin and sleeve connector. An easily accessible NEMA 3R service disconnect in the immediate area is preferable. The ground for the mobile unit 100 can be, for example, originated at the system power source, e.g., transformer or first access point of power into a facility, and be continuous to the system power disconnect on the mobile trailer 120. This ground can be spliced with high compression fittings and can be terminated at each distribution panel it passes through. When it is broken for a connection to a panel, it can be connected into an approved grounding block with the incoming and outgoing ground in this same grounding block, which then can be connected to the steel panel. The connection at the power source can be at the grounding point of the neutral—ground if a Wye transformer is used. In the case of an external facility, it can be bonded to the facility ground point at the service entrance. In some embodiments, the ground wire can generally be copper wire with a minimum AWG 1/0 or the same size as the power feeders, whichever is larger. This means that if there is a primary feeder to a distribution panel of 500 MCM with a secondary feeder to this system of AWG 1/0 wire, the ground to the distribution panel can be 500 MCM with an AWG 1/0 to the system. The ground wire impedance from the system disconnect, including the ground rod, preferably not have an impedance greater than 2 ohms to earth as measured by one of the applicable techniques, for example, ANSI/IEEE Standard 142-1982. In some embodiments, a 15 feet ground cable can be pre-installed and can be found in the forward most, entry door side of the mobile trailer. In some embodiments, a grounding rod is provided and installed as part of the system installation. When the mobile unit 100 is generating radiation for either imaging or treatment, an exclusion zone is generally required to prevent exposure to either radiation workers or members of the public. This exclusion zone is generally determined based on the level of radiation exposure and local, state and federal requirements. It is possible to add shielding that allows a building to be closer, but the distances allowed may be determined by the customer's physicist and local, state and federal requirements. FIGS. 13-18 depict diagrams of exemplary radiation scatter and leakage measurements conducted on the mobile unit 100 of FIG. 1. In some embodiments, measurements can be acquired by using a PTW Unidos E electrometer and a PTW TK-30 large volume chamber. Measurements can be engaged at different field sizes (FS) and gantry angles (G). Leakage measurements can be engaged with the multileaf collimator completely closed (or field size of 0 cm×0 cm). Full scatter measurements can be produced with the collimator in the full open position (or 28 cm×28 cm). All measurements can be displayed as a percentage of the delivered dose at isocenter. All displayed data are at the level of isocenter and radiate out at 1 m increments at various angles from zero degrees to 360 degrees. Based on the exemplary measurements and calculations, expected radiation exposure results for the mobile unit 100 are provided in this disclosure and the above-mentioned distances are exemplary recommendations. The final site plan may also be determined based on distances to adjacent buildings and structures. In addition, as with the installation of any ionizing radiation device, the appropriate site radiation survey should be conducted to verify compliance with these recommendations. Failure to correctly calculate and construct the radiation barriers and shielding as required may result in radiation exposure levels that are in excess of allowable limits, and may present hazards to radiation workers and members of the public. In some embodiments, 6-10 anchoring points embedded in the support pad 300 are provided. In some embodiments, it is preferable that a minimum of 6 anchor points be used. See FIG. 12 for a typical pad layout and typical tie down. Actual thickness of the support pad 300 can be based on, for example, site conditions, coach weight distribution, and other factors. More or less anchoring points can optionally be used. A typical LINAC system uses a 6 MV FFF beam. The maximum dose rate can be 800 cGy/min. The maximum treatment field can be 28 cm×28 cm. The isocenter can be 100 cm. The unit can employ a beam stop so that the primary consideration for shielding is leakage and scatter. A typical LINAC system can deliver 3D, IMRT, and VMAT treatments. Shielding considerations for the mobile radiation oncology coach system 10 can have the following exemplary assumptions: workload, use factors (U=1), occupancy times, design goal (permissible limits), distances, and utilization rate (beam on time). These considerations allow many variants to the external shielding design of the mobile radiation oncology coach system 10. Design goals for unrestricted areas can be set as 1 mSv/yr (0.02 mSv/wk). Design goals for restricted areas can be set as 5 mSv/yr (0.10 mSv/wk). The conventional exposure rate in any one hour of 2 mR/hr guideline can be used. In addition, occupancy factors and utilization rate can be considered. In addition, actual instantaneous dose rates may optionally be considered as well. The following lists an exemplary series of iso-scatter/leakage measurements (see FIGS. 13-18) for calculating shielding requirements. As a guideline, the following parameters were used to calculate the required shielding. ParametersValueWorkload1,200,000 mGy/wk (based on a workload of 40 patientsper day)Use Factor1.0 (leakage and scatter)OccupancyBased on locationDistancesMeasured in meters with 0.3 meter from barrierDesign Goal1 mSv/year (unrestricted), 5 mSv/year (restricted)IMRT4 (used for workload leakage) Referring to FIGS. 2-3, the mobile trailer 120 has a front end 122 and a rear end 124. Between the front end 122 and rear end 124, the interior space of the coach 120 can be partitioned into multiple rooms. In some embodiments, as shown in FIGS. 2-3, the coach 120 includes a control console room or area 130, a vestibule 140, and a treatment room or area 150. As shown in FIGS. 2-3, the treatment room 150 is closer to the rear end 124 than the vestibule 140 and the control console room 130 while the vestibule 140 is closer to the rear end 124 than the control console room 130. In some embodiments, other partitions or arrangements for the rooms/areas 130, 140, 150 can be provided. For example, in some embodiments, only the treatment room 150 and control console area 130 are partitioned. In other embodiments, only the treatment room 150 is provided in the coach 120 and the functions of the control console room can be provided outside of the coach 120 via, for example, wired or wireless communications. In some embodiments, the control console room 130 can contain the operator's station and the planning station. The control console room 130 can also be an entry room and has a front door 132 for entering and exiting the coach 120. A stair 134 can be provided to facilitate the access. In some embodiments, the control console area 130 has an access door 136 for connecting the control console area 130 and the vestibule 140. In some embodiments, a door 142, for example a sliding door, is provided for connecting the vestibule 140 and the treatment room 150. The sliding door 142 can slide into wall 148. The sliding door 142 can be loaded with lead bricks to form a pocket door. In one embodiment, the pocket door 142 can weight about 5000 lbs. Referring also to FIGS. 8A-8C, in some embodiments, through an engagement mechanism 151, the sliding door 142 can be mechanically driven by a drive chain 147 that is engaged with a motor 141. The motor 141 can be controlled by door switches 145 from the treatment room 150 and switches (not shown) from the vestibule 140. A track or rail 149 can be installed at an upper location of the sliding door 142 to accept one or more wheels 155 that are coupled to the sliding door 142 through a supporting mechanism 153. Additional wheels can be provided at the bottom of the sliding door 142 to facilitate the sliding of the door 142 on a bottom track or rail provided on the floor between the vestibule 140 and the treatment room 150. In one embodiment, the track or rail is provided as recessed in the floor about ½ inches. A lever 143 can be provided for manually disengaging the sliding door 142 and the motor 141 so that and the sliding door 142 can be open/close manually from the treatment room 150, for example, in case of an emergency. Referring also to FIG. 9, in some embodiments, the vestibule 140 has a swing door 144 which is preferably closed for the treatment to be delivered. One or more interlock switches 157 can be installed for redundancy and fail safe to insure the swing door 144 is shut during treatment delivery. The vestibule 140 can be provided with a door 146 that can be used as the primary entrance or for use with the wheelchair lift. A patient under treatment can also enter the coach 120 from the front door 132. After confirmed by a representative of the control console room 130, the patient can enter the vestibule 140 through the access door 136. Then the patient can be guided to the treatment room 150 for treatment through the pocket door 142. In some embodiments, the treatment room 150 can be designed for installation of a LINAC system 152 or other treatment or diagnostic instrument(s). A LINAC system 152 generally uses microwave technology to accelerate electrons in a wave guide and enable these electrons to collide with a heavy metal target to produce high-energy x-rays. These high energy x-rays can be shaped as they exit the machine to conform to the shape of the patient's tumor, enabling the customized beam can be directed to the patient's tumor. The x-ray beam comes out of a part of the accelerator called a gantry 154, which can be rotated around the patient. Radiation can be delivered to the tumor from many angles by rotating the gantry 154 and moving the treatment couch 156. Because radiation may scatter or leak from the treatment room 150 during a patient's treatment, protection to people outside of the treatment room 150 is desired. In some embodiments, the coach 120 can have shielding for protecting the control console area 130 during treatment operations. In some embodiments, lead (Pb) shielding 160, for example interlocked lead bricks, is used in walls 138, 148 and doors 142, 144 to block line of site of leakage and scatter. In other embodiments, standard shielding materials can be used. The wall 138 between the vestibule 140 and the control console room 130 can provide secondary or additional shielding for protecting the control console area 130. Referring to FIGS. 5-7, internal lead shielding formed by interlocked lead bricks 160 is applied to the wall 148 and doors 142, 144. The actual thickness of the interior walls and doors are determined based on standard techniques. In some embodiments, using standard lead (Pb) shielding materials, the thickness of lead bricks on the interior wall 138 between the control console room 130 and the vestibule 140 is 2 inches, the thickness of lead bricks on the interior wall 148 between the vestibule 140 and the treatment room 150 is 4 inches, the thickness of lead bricks on the pocket door is 2 inches, and the thickness of lead bricks on the swing door is 2 inches. In some embodiments, the access door 136 between the control console room 130 and the vestibule 140 can be made without shielding and is there for privacy purposes only. Alternatively, the access door 136 can be provided with shielding if necessary. In some embodiments, different thicknesses of the shielding can be used based on different shielding materials and/or radiation scatter and/or leakage. In some embodiments, the pocket door 142 between the vestibule 140 and the treatment room 150 can be a steel door with 2 inches of lead bricks. The steel plates holding the lead in place can be ¼ inches steel or ½ inches steel. In some embodiments, the coach 120 advantageously incorporates alternating doors between the treatment room 150, vestibule 140 and control console room 130 to provide effective shielding. For example, a manual swing door 144 can be added. The manual swing door 144 can contain 2 inches of interlocked lead bricks 160 to take away any potentially not blocked direct line of sight of the machine 154 and those located in the console area. In some embodiments, non-manual or automated doors may optionally be used. In some embodiments, the internal shielding 160 is installed after the trailer 120 has arrived at a designated site. In this configuration, the mobile unit 100 can meet the highway weight limitations set forth by the state authorities. A forklift (not shown) may be utilized to unload the lead shielding 160 (and other accessories) from a secondary vehicle. If a forklift is needed, a site survey may be conducted to determine the size of the forklift prior to the trailer's arrival. Alternatively, the internal shielding 160 is installed before the trailer 120 arrives at the designated site. In this configuration, the mobile unit 100 can be ready for used in a timely manner. In some embodiments, coach 120 is equipped with limited yet adequate internal shielding 160 so as to achieve energy efficiency for relocation of the coach 120 to a designated site. The protection of the public external to the coach 120 is achieved using external shielding provided on the outside of the coach 120. In some embodiments, concrete or high-density concrete for the external shielding is used surrounding the coach 120. The amount of external shielding required can be dependent on, for example, one or more or all of the following factors: workload, distance to surrounding areas, occupancy of surrounding areas, height of surrounding buildings, density of concrete used for shielding, and/or barrier location. The external shielding may be customized and does not need to be symmetric depending on the above listed parameters. The closer the coach 120 is placed to an existing occupied structure, the more shielding will generally be required. In some typical examples, the mobile radiation oncology coach system workload stays fairly constant (35-40 patients per day). The radiation decreases by 1/R2, where R is the distance from the radiation source. In other words, when the distance from the radiation source is doubled, the radiation exposure decreases by approximately ¼. Occupancy Rate (T) can be determined by how often someone will be in a certain area. If people are in an area 100% of the time (T=1) the machine is on, then that area must be shielded as needed/appropriately. If a person is in an area where there is little to no occupancy and no direct line of sight to the particle accelerator, in theory, that area would not need as much shielding. If there are multi-story structures, then that can be taken into account as well in determining the appropriate shielding. In some embodiments, the following external shielding recommendation is based on the following: ParametersValueDesign Goal0.02 mGy/wk (unrestricted)Workload1,200,000 mGy/wkLeakage0.05%Occupancy RateT = 0.5 and T = 1.0ShieldingHigh-Density Concrete (240 pcf) On leakage parameter, Federal regulations require that radiation producing machines cannot exceed 0.1% of the output at 1 meter from the radiation source. Many existing machines are able to achieve 0.05% of the calibrated output at 1 meter. From the measurements depicted in FIGS. 13-18, the leakage parameter may be even less than 0.05%. For the shielding parameter, in some embodiments, concrete of around 147 pcf is used to achieve more cost effective result for adequate shielding. FIG. 10 depicts a top view of the trailer 120 of the mobile unit 100 illustrated in FIG. 1, with external shielding barrier 201-211 positioned according to some embodiments. The trailer or coach 120 illustrated in FIG. 10 is in a clinic mode. As shown in FIG. 10, each of the external shielding barriers or walls 201-211 is placed at a calculated and/or predetermined distance from the nearest surface of coach 120. In some embodiments, the calculated distance is 3 meters. In some embodiments, at least 1 meter is disposed between the furthest out wall and the shielding to account for room to maneuver around the unit when parked. FIG. 11 is table showing exemplary recommended thickness (in inches, when Occupancy Rate T=0.5 and T=1.0, respectively) for each external shielding barrier 201-211 identified in FIG. 10. As depicted in FIG. 11, barrier 210 is thicker than barrier 211 although both barrier 210 and barrier 211 are placed on relatively symmetrical positions. In consideration of the position of the pocket door 142, it is preferable to have the barrier 210 thick enough to prevent or reduce any potential radiation leaked through the pocket door 142 or the surrounding of the pocket door 142. In other embodiments, barrier 210 and barrier 211 can have the same thickness or other thicknesses. The height of the external shielding walls 201-211 is dependent on location of nearby structures. This will be different on each location the unit is placed. The height of the external barriers 201-211 is configured to block a direct line of sight of the leakage coming from any gantry position. Alternative thicknesses, heights and/or materials may optionally be used to accomplish similar shielding results. In the present embodiment, the ceiling of coach 120 is not shielded. Alternatively, the coach 120 can also include shielding in the ceiling. This may help to reduce the shielding height and thickness of the external shielding barriers 201-211. Skyshine can be evaluated as needed. In the present embodiment, shielding below the trailer 120 is not required for lateral barriers and for the rear of the trailer 120. The external barriers 201-211 will block any ground scatter. Alternatively, shielding below the trailer between the tractor 110 and the control console area 130 can be provided. In some embodiments, sand is used to provide such shielding. The amount of sand can be range from 30″ to 36″. In some embodiments, external shielding 201-211 may require using L-block shields where the external shield abuts with the trailer 120. This is to ensure there are no areas of leakage. The x-ray radiation generated by the mobile unit 100 is typically 6 MV. It is typically when about 10 MV that pair production is achieved and elements become radioactive. As concrete is a low Z-element, even at high energy levels, advantageously no radioactive material is created. Accordingly, all external shielding barriers 201-211 will not be contaminated after used. In some embodiments, external shielding barriers 201-211 can be removed from site after use and can be reused. A full radiation survey can be conducted around the trailer 120 after installation. Any areas that exceed limits for unrestricted areas will be marked as restricted areas. These areas may not be occupied by members of the general public, and any professionals working in these areas may be permitted based on training and being equipped with the appropriate monitoring badge (personal monitor). Areas that have no occupancy may require little to no shielding. These areas can preferably be treated as restricted and access to these areas may be limited. Fences and appropriate signage may be required. These areas are preferably monitored closely by the staff and security. It is to be understood that the invention is not limited in its application to the details of construction and to the arrangements of the components set forth in the following description or illustrated in the drawings. The invention is capable of other embodiments and of being practiced and carried out in various ways. Also, it is to be understood that the phraseology and terminology employed herein are for the purpose of description and should not be regarded as limiting. As such, those skilled in the art will appreciate that the conception, upon which this disclosure is based, may readily be utilized as a basis for the designing of other structures, methods and systems for carrying out the several purposes of the present invention. It is important, therefore, that the invention be regarded as including equivalent constructions/processes to those described herein insofar as they do not depart from the spirit and scope of the present invention. For example, the specific sequence of any described component and/or process may be altered. For example, certain processes are conducted in parallel or independent, with other processes, to the extent that the processes are not dependent upon each other. Other alterations or modifications of the above components and/or processes are also contemplated. For example, further insubstantial changes to the components, systems and/or processes are also considered within the scope of the processes described herein. In addition, features illustrated or described as part of one embodiment can be used on other embodiments to yield a still further embodiment. Additionally, certain features may be interchanged with similar devices or features which perform the same or similar functions. It is therefore intended that such modifications and variations are included within the totality of the present invention. The disclosures of all articles and references, including patent applications and publications, are incorporated by reference for all purposes. The term “consisting essentially of” to describe a combination shall include the elements, ingredients, components or steps identified, and such other elements ingredients, components or steps that do not materially affect the basic and novel characteristics of the combination. The use of the terms “comprising” or “including” to describe combinations of elements, ingredients, components or steps herein also contemplates embodiments that consist essentially of the elements, ingredients, components or steps. By use of the term “may” herein, it is intended that any described attributes that “may” be included are optional. By use of the term “at least one of A and B” herein, it is intended to mean “one or more of X and/or Y.” Plural elements, ingredients, components or steps can be provided by a single integrated element, ingredient, component or step. Alternatively, a single integrated element, ingredient, component or step might be divided into separate plural elements, ingredients, components or steps. The disclosure of “a” or “one” to describe an element, ingredient, component or step is not intended to foreclose additional elements, ingredients, components or steps. The detailed description, for purposes of explanation, used specific nomenclature to provide a thorough understanding of the invention. However, it will be apparent to one skilled in the art that specific details are not required in order to practice the invention. Thus, the detailed descriptions of specific embodiments of the invention are presented for purposes of illustration and description. They are not intended to be exhaustive or to limit the invention to the precise forms disclosed. Modifications and variations of the above detailed description are considered within the scope of the described invention. The embodiments were chosen and described in order to best explain the principles of the invention and its practical applications, they thereby enable others skilled in the art to best utilize the invention and various embodiments with various modifications as are suited to the particular use contemplated. It is intended that the following claims and their equivalents define the scope of the invention.
054426685	abstract	A nuclear reactor has a containment vessel and a voided calandria space having a plurality of fuel channels therein. A chamber surrounding the calandria space is receptive of a light water moderator and a solid reflector in the chamber surrounds the calandria space. Each fuel channel has a solid fuel matrix having a plurality of coolant holes extending longitudinally therethrough and receptive of light water coolant, a pressure tube surrounding the fuel matrix and a calandria tube surrounding the pressure tube and forming a gap therebetween.
	abstract	A control rod drive mechanism (CRDM) comprises a lead screw, a motor threadedly coupled with the lead screw to linearly drive the lead screw in an insertion direction or an opposite withdrawal direction, a latch assembly secured with the lead screw and configured to (i) latch to a connecting rod and to (ii) unlatch from the connecting rod, the connecting rod being free to move in the insertion direction when unlatched, and a release mechanism configured to selectively unlatch the latch assembly from the connecting rod.
	claims	1. An accident tolerant fuel combination for light water and lead fast reactors comprising:a multi-layered ceramic cladding; andfuel in pellet form positioned within the cladding such that a gap sized to prevent pellet-cladding mechanical interaction and centerline melt during high power excursions is defined between the cladding and the pellet, the fuel comprising U15N intermixed with a boron-containing integral fuel burnable absorber selected from the group consisting of UB2 and ZrB2 and having a 1310 content at 19% to 80% of the boron. 2. The fuel combination recited in claim 1 wherein the ceramic cladding comprises at least one monolith layer and at least one composite layer. 3. The fuel combination recited in claim 2 wherein the ceramic cladding is made of a SiC monolith layer and a SiC ceramic composite layer. 4. The fuel combination recited in claim 1 wherein the ceramic cladding has a total wall thickness between 0.4 mm and 1.4 mm. 5. The fuel combination recited in claim 1 wherein the U15N fuel has a N15 isotope content between 75% and 99.9%. 6. The fuel combination recited in claim 1 wherein the U15N fuel has a UN purity greater than 90%. 7. The fuel combination recited in claim 1 wherein the boron-containing integral fuel burnable absorber is UB2. 8. The fuel combination recited in claim 1 wherein the boron-containing integral fuel burnable absorber content in the U15N pellet is between 100 ppm and 10000 ppm. 9. The fuel combination recited in claim 1 wherein the U15N fuel has a density between 80% and 99% of theoretical density. 10. An accident tolerant nuclear fuel combination comprising:a ceramic cladding comprised of at least a monolith layer and a composite layer;a plurality of fuel pellets stacked in the cladding such that a gap sized to prevent pellet-cladding mechanical interaction and centerline melt during high power excursions is defined between the cladding and the fuel pellets, the fuel pellets formed from U15N mixed with from 100 to 10000 ppm of a boron-containing integral fuel burnable absorber selected from the group consisting of UB2 and ZrB2, having a B10 content at 19% to 80% of the boron. 11. The nuclear fuel combination recited in claim 10 wherein the ceramic cladding is formed from a SiC monolith wrapped in SiC composite fibers. 12. The nuclear fuel combination recited in claim 10 wherein the U15N fuel has a N15 isotope content between 75% and 99.9% and a UN purity greater than 90%. 13. The fuel combination recited in claim 1 wherein the gap is between about 0.01 to 0.3 mm. 14. The nuclear fuel combination recited in claim 10 wherein the gap is between about 0.01 to 0.3 mm.
	summary
	abstract	Process for simulation of the response of a detector of radiation emitted by radioactive objects and process for inspection of nuclear fuel elements using this simulation.
047479965	abstract	An apparatus and method for adjusting the level of nuclear fuel rods in a fuel assembly is disclosed, and which comprises a tool composed of an expander having a pair of parallel, laterally spaced apart and flexible arms which have first ends which are interconnected to each other, and opposite free ends. The tool further includes an actuator disposed between the free ends of the arms, and the free ends of the arms and the actuator mount cooperating wedge surface means, whereby relative longitudinal movement between the expander and actuator causes the arms to separate a predetermined maximum distance. In use, the tool is inserted into the fuel assembly with the free ends of the arms located between the end of a misaligned fuel rod and the nozzle, and the actuator is then longitudinally advanced to separate the arms to the predetermined maximum distance, which serves to move the misaligned rod to its proper level. During the separation of the arms, the arms may be manually pressed toward each other so that the free ends are maintained in a parallel relationship and so as to avoid lateral forces being imparted to the misaligned fuel rod.
062663895	claims	1. A method for manufacturing an exposure apparatus which transcribes an image of a projection master on a substrate by a projection system having a plurality of reflection mirrors, comprising the steps of: a) measuring an image property of said projection system having said plurality of reflection mirrors; b) selecting from said image property an image property component to be adjusted; c) selecting at least one of said reflection mirrors which can adjust said image property component; d) calculating a surface shape of reflection mirror which can adjust substantially said image property component based on image property measurement results from step a); e) giving said reflection mirror selected in step c) the surface shape as calculated in step d); f) embedding said reflection mirror having said surface shape into said projection system. g) selecting positions for the plurality of said reflection mirrors in said projection system, wherein the positions of said reflection mirrors are selected so that to adjust substantially said image property. depositing a photosensitive material on said substrate; projecting an image of the pattern on said mask onto said substrate through said projection system; developing said photosensitive material on said substrate; transforming a predetermined circuit pattern onto said substrate by using said developed photosensitive material as a mask. depositing a photosensitive material on the substrate; projecting an image of the pattern on said mask onto said substrate through said projection system; developing said photosensitive material on said substrate; forming a predetermined circuit pattern onto said substrate by using said developed photosensitive material as a mask. depositing a photosensitive material on the substrate; projecting an image of the pattern on said mask onto said substrate through said projection system; developing said photosensitive material on said substrate; forming a predetermined circuit pattern onto said substrate by using said developed photosensitive material as a mask. 2. A method for manufacturing an exposure apparatus according to claim 1, wherein said image property component includes an image magnification error, an image skew, a curvature of an image plane, a gradient of an image plane, a displacement of a focal point caused by a direction in an aperture, a displacement of a focal point caused by a numerical aperture, and a telecentric error. 3. A method for manufacturing an exposure apparatus according to claim 1, further comprising step of: 4. A method for manufacturing an exposure apparatus according to claim 3, wherein said image property is measured using a light having same wave length as in said projection system in said step a). 5. A method for manufacturing an exposure apparatus according to claim 3, wherein said image property is measured by using a light having a wave length different from that used in said projection system in step a). 6. A method for manufacturing an exposure apparatus according to claim 3, wherein a reflection mirror is given the surface shape as calculated in step d), and said reflection mirror in the projection system which is selected in step c) is interchanged with the reflection mirror having said surface shape. 7. A method for manufacturing an exposure apparatus according to claim 6, wherein said reflection mirror selected in step c) is the reflection mirror placed near said projection master or said substrate. 8. A method for manufacturing an exposure apparatus according to claim 6, wherein said reflection mirror selected in step c) is the reflection mirror placed near an aperture stop of said projection system. 9. An exposure apparatus comprising an X-ray source, an illumination system for guiding said X-ray from said X-ray source to a mask, a projection system for projecting a pattern on said mask by guiding said X-ray to an exposed plane through said mask, wherein said projection system is manufactured by the method according to claim 1. 10. An exposure apparatus comprising an X-ray source, an illumination system for guiding said X-ray from said X-ray source to a mask, a projection system for projecting a pattern on said mask by guiding said X-ray to an exposed plane through said mask, wherein said projection system is manufactured by the method according to claim 6. 11. An exposure apparatus comprising an X-ray source, an illumination system for guiding said X-ray from said X-ray source to a mask, a projection system for guiding said X-ray to an exposed plane through said mask and projecting a pattern on said mask, wherein said projection system comprises a plurality of reflection mirrors, and at least one reflection mirror of the plurality of reflection mirrors is interchangeable with a reflection mirror having a surface shape different from said one reflection mirror. 12. An exposure apparatus according to claim 11, wherein certain aberration components among a plurality of aberration components in said projection system may be changed by interchanging said one reflection mirror with the reflection mirror having different surface shape without substantially influencing other aberration components. 13. An exposure apparatus according to claim 12, wherein said one reflection mirror is a reflection mirror placed near said projection master or said substrate. 14. An exposure apparatus according to claim 13, wherein said one reflection mirror has insignificant power. 15. An exposure apparatus according to claim 13, wherein said one reflection mirror is a reflection mirror placed near said aperture stop of said projection system. 16. An exposure apparatus according to claim 15, wherein said one reflection mirror has insignificant power. 17. An exposure apparatus according to claim 15, wherein said aperture stop is placed on at least one reflection mirror of the plurality of said reflection mirrors constituting said projection system. 18. A method for manufacturing a device using the exposure apparatus according to claim 12, comprising: 19. A method for manufacturing a device using the exposure apparatus according to claim 13, comprising: 20. A method for manufacturing a device using the exposure apparatus according to claim 15, comprising:
	abstract	A method and monitoring system for tracking operational parameters of a drive train of a monitored system includes a sensor unit affixed to the drive train. The sensor unit includes a sensor for generating an analog signal in response to an operating condition of the monitored system and an integrated local processor for processing the analog signal generated by the sensor. Actions of a sensor unit associated that is mounted from movement of a component of the drive system may be initiated via wireless receipt of instructions from another processing unit.
	abstract	A method of forming a protection layer on a specimen for TEM inspection and a method of forming a specimen for TEM inspection are provided. The method of forming a protection layer on a specimen for TEM inspection generally comprises coating a wafer slice comprising an inspection point with a protection material and compressing the protection material to the wafer slice. The method of forming a specimen for TEM inspection generally comprises cutting a wafer slice comprising an inspection point from a wafer, forming a protection layer on the wafer slice, forming a first preliminary specimen by cutting the wafer slice, forming a second preliminary specimen by grinding the first preliminary specimen, and forming a TEM specimen by etching portions of the second preliminary specimen.
050376047	claims	1. A structure for shielding stored internals' radiation from an area where a reactor vessel is housed in a containment building, comprising: a plurality of segments introducible through an equipment hatch of the containment building, said segments being connectable to each other into a coffer dam and being connectable to an upper flange of the reactor vessel; first connecting means for connecting said segments to each other to form the coffer dam; first sealing means positionable between said segments for forming a seal therebetween; second connecting means for connecting the resulting coffer dam to the upper flange of the reactor vessel; and second sealing means for forming a seal between the coffer dam and the upper flange of the reactor vessel. a plurality of bolt and nut combinations connecting adjacent segments. a strip seal between adjacent segments. a plurality of bolts connecting a bottom flange of the coffer dam to the upper flange of the reactor vessel. one of an O-ring and gasket between a bottom flange of the coffer dam and the upper flange of the reactor vessel. an equal vertical section of a cylinder. a horizontal section of a cylinder. one of a vertical and horizontal section of a cylinder. a two-piece, circular member connected by fourth connecting means. (a) introducing a plurality of segments through an equipment hatch of the containment building; and (b) assembling the plurality of segments to each other in sealing relation with sealing means between the segments to form a coffer dam; and (c) fixedly connecting the coffer dam to an upper flange of a reactor vessel in sealing relation. (a) pre-fabricating a plurality of segments at a location outside the containment building; (b) transporting the plurality of segments to the containment building; (c) introducing each of the plurality of segments into the containment building; (d) connecting the plurality of segments to each other in sealing relation with sealing means between the segments to form a coffer dam; (e) fixedly connecting the coffer dam to an upper flange of the reactor vessel in sealing relation. (a) removing nuclear fuel from the reactor vessel; (b) removing and storing upper and lower core internals underwater in a refueling canal; (c) introducing a plurality of segments individually through a hatch in the containment building; (d) connecting the plurality of segments to each other in sealing relation with sealing means between the segments to form a coffer dam; (e) fixedly connecting the coffer dam to an upper flange of the reactor vessel in sealing relation; (f) placing an annealing apparatus in the reactor vessel; (g) pumping water from the reactor vessel and coffer dam; (h) annealing; (i) filling the reactor vessel with water; (j) removing the annealing apparatus. (k) disconnecting the coffer dam from the reactor vessel upper flange; (1) disassembling the coffer dam; and (m) removing the plurality of segments from the containment building. 2. The structure as recited in claim 1, wherein the first connecting means comprises: 3. The structure as recited in claim 1, wherein the first sealing means comprises: 4. The structure as recited in claim 1, wherein the second connecting means comprises: 5. The structure as recited in claim 1, wherein the second sealing means comprises: 6. The structure as recited in claim 1, wherein each segment comprises: 7. The structure as recited in claim 1, wherein each segment comprises: 8. The structure as recited in claim 1, wherein each segment comprises: 9. The structure as recited in claim 1, wherein at least one segment comprises a side port. 10. The structure as recited in claim 1, further comprising an access cover plate connected to an upper flange of the coffer dam by third connecting means and third sealing means. 11. The structure as recited in claim 10, wherein the third connecting means comprises a plurality of bolts. 12. The structure as recited in claim 10, wherein the third sealing means comprises an O-ring. 13. The structure as recited in claim 10, wherein the cover comprises: 14. A method for assembling a coffer dam relative to a reactor vessel housed in a containment building, comprising the steps of: 15. A method for assembling a coffer dam used in an operation requiring human intervention in a containment building, comprising the steps of: 16. A method for shielding stored internals' radiation in preparation for annealing a reactor vessel in a containment building, comprising the steps of: 17. The method as recited in claim 16, further comprising the steps of:
050948040	summary	The invention relates to nuclear fuel element construction and composition, and to a method of manufacture for making fuel elements that are particularly useful in high temperature gas reactors. More particularly, the invention relates to fuel elements that are useful in high temperature reactors in which the normal operating temperature of the reactor is in excess of 2,000 degrees centigrade and exceeds the melting temperature of fissionable material that is localized and stabilized within the reactor fuel elements by a combination of capillary and surface tension forces that result from the unique structure and composition of the elements. In recent years attempts have been made to design small nuclear power reactors that can be used as the source of driving power on space ships or other vehicles that impose stringent space and weight limitations on the reactor design. In such designs, it has been proposed that particle bed reactors be used, in which relatively small diameter fuel particles in the range of 500 microns in diameter are provided to supply high power densities within the reactor. It has also been proposed that the fuel charge in such reactors be changed or reloaded on a relatively short-cycle basis. For example, such reloading could occur every several hours or perhaps once per day. The normal sustained operating temperature of the moderating gas in such prior art high temperature gas reactors is in a range that does not exceed about 1,000 degrees centigrade and is not in excess of the melting point temperature of the nuclear fuel compositions used in the fuel elements for the reactors. In order to prevent the nuclear fuels from being diffused or evaporated from the reactor prematurely, that is before the fuels serve their intended function of transmitting a major proportion of their power content to the hot gases within the reactor, it has been necessary to design the fuel elements with a multi-layered structure that enables its outer layers to prevent premature diffusion of the fissionable material from the fuel elements into the reactor moderating gases. So far as the inventor knows, prior to the present invention there was not known or available any nuclear fuel element or reactor concept in which nuclear fuels could be used at high temperatures that substantially exceed the melting point temperature of the fissionable materials in such fuels by several hundred degrees. By permitting the use of molten fissionable fuels, the present invention increases the operating temperature and resultant performance of high temperature gas reactors by many hundreds of degrees above the known maximum sustained operating temperatures heretofore achieved by gas moderated reactors. An example of the type of composite nuclear fuel elements that were developed for use in prior art high temperature gas reactors is shown in U.S. Pat. No. 3,212,989. It describes the use of nuclear fuel elements that include at least two independent sealing zones around a core material for retaining fission products within a sealing jacket. Such prior art nuclear fuel elements and related systems have limitations that are imposed by their inherent graphite reactions, and they have two classes of temperature limitations. First, the outlet gas temperature from the associated reactor is limited to about 1,000 degrees centigrade by their primary circuit metal properties and, second, the maximum fuel temperature is limited by melting, with resultant diffusion or evaporation of the fuel before its power has been used to heat the gases in the reactor. As noted in that prior art patent, the type of "high temperature" reactors contemplated for use with the disclosed, jacketed fuel elements has an operating temperature in the range of 700.degree. C. to about 1,000.degree. C. It should also be noted that in such prior art "high temperature" gas reactors there exists geometry problems that arise from the need to separate the reactor structural components from the nuclear fuel. Such limitations have, prior to the present invention, prevented the development of high temperature gas cooled reactors that have an outlet temperature substantially above 1000.degree. C., and have maximum short-term operating fuel temperatures of only about 2000.degree. C. In general, the operating temperatures in present day nuclear fuel reactors are limited by the melting points of the nuclear fuels used to power the reactors. The present invention discloses a process that is used to form predetermined microscopically localized liquid nuclear fuel concentrations within a confining graphite or carbon fuel element, in a manner such that the fuels are capable of performing at temperatures up to the sublimation temperatures of the confining graphite or carbon, i.e. is in the range of about 3300.degree. C. The invention also discloses novel structures of a variety of useful nuclear fuel elements that are made to include predetermined and controllable porosity configurations for localizing and confining fissionable nuclear fuel within the fuel elements. Common causes of failure of known prior art nuclear fuel elements at the elevated operating temperatures of high temperature gas reactors result from either the reaction and decomposition of the carbide coating of the element surrounding the fissionable fuel, or from the migration of molten fuel and fission products through the pores of the surrounding carbon or graphite element into the moderating gases before the fuel has discharged its energy to heat the gases. SUMMARY OF THE INVENTION A primary object of the present invention is to provide a nuclear fuel element having a distribution of fissionable material within the pores of a confining graphite or carbon member so that the fissionable material is thermodynamically stable with respect to its migration beyond the pores of the confining graphite or carbon member. A further object of the invention is to provide a method for making nuclear fuel elements in a commercially feasible manner such that the porosity of carbon or graphite fuel members within such elements is controlled to a predetermined degree, so that fissionable material can be effectively deposited within the pores of the carbon or graphite members, which are then sealed, and the members are heat treated to melt the fissionable materials causing them to react with the carbon walls of the members to increase the fuel localizing and stabilizing porosity of the members. Yet another object of the invention is to provide a nuclear fuel element having fissionable material confined within pores of a carbon or graphite member in a manner such that when the nuclear fuel is heated to above the melting point of the fissionable material the molten fuel material is held by capillary forces and surface tension forces within the pores of the surrounding graphite or carbon member. A still further object of the invention is to provide a nuclear fuel element in which a porous graphite or carbon member is impregnated with fissionable material, which is localized within the pores of the member, even after the fissionable material melts during operation of the reactor. Both capillary and surface tension forces hold the molten fuel within the pores and the fuel is also stabilized within the pores by coating the member with one or more layers of pyrolytic carbon or diamond. Still another object of the invention is to provide nuclear fuel elements in the form of flexible filaments having predetermined and controlled porosity such that fissionable material disposed within the pores of the element is localized therein by capillary and surface tension forces when the fissionable material is melted during its use within a reactor. To assure further localization of the molten fissionable material within the flexible filament fuel elements, one or more layers of pyrolytic carbon or diamond are formed over the exterior surface of the carbon or graphite member to afford additional surface barriers to the migration or diffusion of fissionable material from the pores of the carbon or graphite member. Additional objects and advantages of the invention will become apparent to those skilled in the art from the description presented herein, considered in conjunction with the accompanying drawings. In the preferred practice of the method of the invention conventional high temperature gas reactor type carbon or graphite fuel particle members are impregnated with oxidants and subjected to a high temperature reaction to develop a predetermined and controlled degree of porosity within the members. The porous members are then impregnated with a solution of fissionable material, such as uranyl nitrate in a suitable solvent, such as water, methanol, etc., to fill the pores of the members with the solution. The solvent is then removed by heat treatment of the members and further heat treatment at higher temperatures reacts the fissionable material with the graphite of the members thereby to further increase the porosity of the members. Additional fuel impregnations may be made to achieve desired, controllable levels of fuel loading of the porous members. The matrix of carbon or graphite members is then soaked in a solution of sugar or other organics and is again heated, thereby to form a layer of carbon that plugs the pores at the surface of the graphite members. The fuel elements are then heated above the melting point temperature of the fissionable material in order to react the fissionable material with the walls of the pores. The molten fuel is held by capillary and surface tension forces within the pores of the graphite members. In a most preferred embodiment of the invention, additional assurance for localization of the molten fissionable material within the pores of the fuel element is afforded by forming one or more layers of pyrolytic carbon or diamond over the outer surface of the porous graphite members, using conventional fluidized bed or controlled vapor deposition techniques for depositing such layers. The novel nuclear fuel elements formed by the method of the invention may be made as generally spherically shaped pellets, elongated filaments, or other arbitrary configurations. In one preferred form, the fuel elements are formed as flexible filaments of carbon or graphite material having predetermined and controllable porosity and having fissionable material localized within the pores of the carbon members by the capillary and surface tension forces existing between the surface of the pores and the molten fissionable material, when the molten state is achieved by heating the fuel above its melting temperature. Alternate layers of pyrolytic carbon and/or diamond are deposit on the outer surface of the porous carbon or graphite members to provide further assurance against migration or diffusion of molten fissionable material from the pores of the graphite or carbon members.
	summary
053012127	summary	FIELD OF THE INVENTION The invention relates to a process and an apparatus for dismantling an irradiated component of a nuclear reactor, particularly a vessel of a nuclear reactor cooled by pressurized water. BACKGROUND OF THE INVENTION Water-cooled nuclear reactors, particularly pressurized-water nuclear reactors, comprise a vessel which is intended for containing the core of the nuclear reactor and which is connected to the reactor cooling circuit in which the cooling water circulates. The wall of the reactor vessel which is in contact with the cooling fluid and which is exposed to the radiation emitted by the reactor core can be activated and contaminated after the reactor has been in operation for some time. In the case of nuclear power stations which have reached the end of their life and which require a complete shutdown, the solution adopted in the past has been to leave these power stations in their existing state and to allow the activity of the constituent materials of their components to decrease, in order subsequently to dismantle them under more satisfactory conditions than at the time of the shutdown, without the need to employ complex, remotely controlled equipment. The number of power stations put out of industrial operation will increase appreciably in the future, and it is therefore necessary to consider dismantling these power stations in order to restore the site where they are installed to its original state. The dismantling of the conventional part of the power station presents no particular problem, but, in contrast, the dismantling of the part of the power station constituting the actual nuclear reactor poses problems which are difficult to solve in view of the radioactive emissions of the constituent materials of the reactor components. In particular, the vessel of water-cooled nuclear reactors, which contains the fuel assemblies and which is in contact with the cooling water of the reactor during its operation, is activated and contaminated where reactors which have reached the end of their life are concerned. As regards pressurized-water nuclear reactors in operation at the present time, the reactor vessel takes the form of a body of generally cylindrical shape closed by domed bottoms, of large size and having a considerable wall thickness. The vessel, which has a very high mass, is arranged within a vessel well made in a concrete structure which also delimits one or more pools located above the upper level of the vessel. The vessel which contains not only the fuel assemblies but also various internal structures, is connected by means of connection pieces to pipelines of the primary circuit of the reactor. The core assemblies and some components of the internal structures can be dismantled and removed from the vessel, in order to ensure their disposal and, if appropriate, their elimination at the time when the reactor is put out of operation. Some components of the highly activated internal structures of the reactor, such as the shroud of the core, may need to be kept inside the vessel so as to be cut under water (radiological protection). Their dismantling has to be carried out within the vessel and during the operations of dismantling the vessel itself. To date, no process and apparatus is known which enables the vessel of a pressurized-water nuclear reactor to be dismantled under very good safety conditions without the risk of radioactive contamination in the work zone, while at the same time using machining and handling means of relatively simple design in order to carry out the fragmentary disposal and elimination of the material of the vessel. SUMMARY OF THE INVENTION The object of the invention is, therefore, to provide a process for dismantling an irradiated component of a nuclear reactor, comprising at least one wall of tubular shape arranged with its axis in the vertical direction and fastened inside a well made in a concrete structure, this process making it possible under very good safety conditions and in a simple way to carry out the fragmentation of the wall of the component and the disposal and elimination of the fragments obtained. To achieve this object: the connecting elements between the concrete structure and the component are destroyed, the component is displaced some distance in the vertical direction along its axis on the inside of the well and in successive steps, the wall of the component is cut over a height corresponding substantially to the vertical displacement distance, so as to obtain blocks of the irradiated material of the wall, at the upper level of the well of the concrete structure after each displacement of the component, the cut blocks are disposed of for the purpose of effecting their elimination or storage, and the cutting of the component is carried out in successive steps separated by a vertical displacement. Advantageously, and in order to increase the safety of the process, to carry out the displacement of the component in the vertical direction in successive steps: first means for raising the component are placed under a lower part of the component and so as to bear on a stationary support resting on the concrete structure of the reactor, in the vicinity of the bottom of the vessel well, the component is lifted by a push of the first raising means on the lower part of the component, a first modular supporting element is introduced between the lower part of the component and the stationary support on which the modular element comes to bear, the first raising means are actuated oppositely to the lifting direction, in order to bring the lower part of the component to bear on the first modular element, and for each of the subsequent successive displacement steps of the component: a unit lift of the component over a specific vertical distance is executed by second raising means bearing on the support and in engagement with a modular supporting element interposed between the component and the stationary support and resting on the stationary support before the unit lift of the component, a modular supporting element, the height of which is smaller than the vertical distance of unit lift of the component, is introduced between the modular element with which the second raising means interact and the stationary support, and the second raising means are actuated oppositely to the lifting direction, in order to bring the component to bear on the support by means of the superposed modular elements.
046831061	claims	1. A wiring installation located in a nuclear reactor building containing a nuclear reactor vessel having a vertical axis provided with a cover for connection of electric devices of the nuclear reactor, comprising: (a) a plurality of layers of fluid tight conduits located substantially above said cover and disposed perpendicularly to the said axis of the reactor vessel, the conduits in each layer being substantially parallel to each other and the conduits in one layer being disposed at an angle to the conduits in another one of said layers, (b) a plurality of first cables each extending from one of said devices and terminated with a first fixed connector located above said cover, those of the first fixed connectors which correspond to similar devices having similar positions, (c) a plurality of second cables each associated with one of the first cables and each having (d) means for supporting said second cables between the first fixed connectors and the second fixed connectors, and (e) wherein each one of said first fixed connectors is associated with one of said lead-in connectors which opens into an adjacent one of said conduits, so that all the first sections each connecting one of the lead-in connectors to one of the first fixed connectors, and corresponding to similar devices, are of the same length, while the first sections corresponding to different devices are of different lengths. 2. An installation as claimed in claim 1, wherein a plurality of said devices are for redundancy of a given measurement, and the first fixed connectors relative to said redundant devices are associated with lead-in connectors carried by separate conduits. 3. An installation as claimed in claim 1, wherein all those third sections extending in the same direction from mutually parallel conduits are supported by the same bridge up to the same plate at a distance from the reactor. 4. An installation as claimed in claim 1, wherein the lead-in connectors associated with one of the conduits are disposed on lateral surfaces of the conduit and the lead-out connectors are at one of the ends of the conduit. 5. An installation as claimed in claim 4, wherein each of said conduits is divided longitudinally into two separate compartments each comprising one of the lateral surfaces and one of the ends of the conduit.
050283820	abstract	The disclosed invention comprises a method for assembling fuel bundles for service in nuclear reactors which minimizes damage to the assembled components of a nature that renders the components susceptible to destructive corrosion in service, and the enhanced product of the improved method. The invention includes the utilization of a temporary protective barrier, such as water soluble sodium silicate or gelatin, intermediate the components during assembly.
055984509	claims	1. In a fuel bundle for a boiling water nuclear reactor comprising a plurality of fuel rods secured within an array and extending between upper and lower tie plates, and including at least one additional fuel rod extending from said lower tie plate but terminating short of said upper tie plate, the improvement comprising a removable extension rod secured to said at least one additional fuel rod and extending substantially to said upper tie plate. 2. The fuel bundle of claim 1 wherein a plurality of spacers are provided at axially spaced locations along said bundle and wherein said at least one additional rod terminates adjacent and above an upper edge of one of said spacers. 3. The fuel bundle of claim 1 wherein said extension rod is secured to the top of said at least one additional rod such that said extension rod and said at least one additional rod are substantially co-linear. 4. The fuel bundle of claim 1 wherein said extension rod is hollow and filled substantially with single phase water to increase reactivity. 5. The fuel bundle of claim 1 wherein said extension rod is substantially devoid of any hydrogen bearing material. 6. The fuel bundle of claim 5 wherein said extension rod is formed of solid Zircaloy material. 7. The fuel bundle of claim 5 wherein said extension rod is formed of Zircaloy material in tubular form, and wherein gadolinium is added to further reduce reactivity. 8. In a fuel bundle for a nuclear reactor having fuel rods secured with an array and extending generally axially between upper and lower tie plates, and a plurality of spacers at axially spaced locations along said bundle holding said fuel rods in the array, said fuel rods including a first plurality of full length rods secured at lower ends thereof to the lower tie plate and at upper ends thereof to the upper tie plate, and a second plurality of partial length rods secured at lower ends thereof to the lower tie plate, with upper ends thereof terminating above and adjacent one or more of said spacers, the improvement comprising fuel rod extension members removably attached to at least some of the plurality of partial length fuel rods such that said at least some of the plurality of partial length fuel rods extend substantially to said upper tie plate, and such that said fuel rod extension members can be removed from the bundle independently of the partial length fuel rods to which they are attached. 9. The fuel bundle of claim 8 wherein said fuel rod extension members are formed of solid Zircaloy material. 10. The fuel bundle of claim 8 wherein said fuel rod extension members are formed as hollow tubes. 11. The fuel bundle of claim 10 wherein said fuel rod extension members contain single phase water. 12. The fuel bundle of claim 10 wherein said fuel rod extension members contain gadolinium. 13. In a fuel bundle for a nuclear reactor having fuel rods secured within an array and extending generally axially between upper and lower tie plates, and a plurality of spacers at axially spaced locations along said bundle holding said fuel rods in the array, said fuel rods including a first plurality of full length rods secured at lower ends thereof to the lower tie plate and at upper ends thereof to the upper tie plate, and a second plurality of partial length rods secured at lower ends thereof to the lower tie plate, with upper ends thereof terminating above and adjacent one of said spacers, the improvement comprising means for causing all of said fuel rods to behave similarly relative to coolant flow pressure drop in the channel, and associated means for varying reactivity of the second plurality of partial length rods.
	abstract	A tungsten sheet includes a tungsten layer. The tungsten layer includes a binder resin and a plurality of tungsten particles included in the binder resin. A tungsten composition amount of the tungsten layer is at least 70% by mass (wt %). An average particle diameter of the plurality of tungsten particles is more than 1 μm and less than 15 μm.
046866949	claims	1. In an apparatus for analyzing a sample of material by the X-ray fluorescence method having a radiation source for radiating the sample, a radiation detector for receiving X-rays generated from the sample, and a detector housing for enclosing the radiation detector, the detector housing having an aperture formed in a side wall thereof, a source housing comprising: a source housing body formed as a hollow, generally right triangular prism tapering from an open base to a tip having an aperture formed therein; means for attaching said base of said source housing body to the detector housing over the aperture formed in the detector housing; and means for mounting the radiation source adjacent said aperture in said tip of said source housing body whereby, when said tip of said source housing body is positioned adjacent the material sample, radiation from the radiation source passes through said aperture in said tip to radiate the material sample and X-rays generated from the material sample pass through said aperture in said tip and said aperture in the detector housing to the radiation detector in said detector housing. 2. The source housing according to claim 1 including shutter means mounted in the source housing between the radiation source and said aperture in said tip, and means for selectively moving said shutter means between an "off" position blocking radiation and an "on" position passing radiation from the source to the aperture. 3. The source housing according to claim 2 wherein the radiation source includes radioactive means enclosed in a generally tubular collimator having a radiation exit port formed therein facing said aperture in said tip, and wherein said shutter means has a generally tubular body enclosing said collimator and having a port formed therein in axial alignment with said collimator port when said shutter means is in the "on" position. 4. The source housing according to claim 1 including radiation pervious means attached to the interior of said source housing body and covering said aperture in said tip. 5. The source housing according to claim 4 wherein said radiation pervious means is formed of a sheet of Mylar material facing a sheet of polypropylene material. 6. The source housing according to claim 1 wherein the radiation source includes a first radioactive means positioned on one side of said aperture and a second radioactive means positioned on another side of said aperture. 7. In an apparatus for analyzing a sample of material by the X-ray fluorescence method having a radiation source for radiating the sample and a radiation detector for receiving X-rays generated from the material sample, a hand-holdable probe comprising: a detector housing enclosing the radiation detector and having an aperture formed in a side wall thereof adjacent the radiation detector, and a source housing having a body formed as a hollow, generally right triangular prism for enclosing the radiation source, said body tapering from an open base attached to said detector housing over said aperture to a tip having an aperture formed therein, the radiation source being positioned adjacent said aperture in said tip. 8. The probe according to claim 7 including shutter means mounted in said source housing body between the radiation source and said aperture in said tip, and means for selectively moving said shutter means between an "off" position blocking radiation and an "on" position passing radiation from the radiation source through said aperture in said tip. 9. The probe according to claim 8 including means responsive to said means for moving for indicating when said shutter means is in said "on" and "off" positions. 10. The probe according to claim 9 wherein said means for indicating includes a shutter position indicating aperture formed in a wall of said detector housing and a shutter position tag positioned in said detector housing and visible through said indicating aperture, said tag being coupled to said means for moving for indicating said "on" and "off" positions of said shutter means. 11. The probe according to claim 8 wherein the radiation source includes radioactive means enclosed in a generally tubular collimator having a radiation exit port formed therein facing said aperture in said tip, and wherein said shutter means has a generally tubular body enclosing said collimator and having a port formed therein in axial alignment with said collimator port when said shutter means is in the "on" position. 12. The probe according to claim 7 including radiation pervious means attached to the interior of said source housing body covering said aperture in said tip. 13. The probe according to claim 12 wherein said radiation pervious means is formed of a sheet of Mylar material facing a sheet of polypropylene material. 14. The probe according to claim 7 wherein radiation source includes a first radioactive means positioned on one side and a second radioactive means positioned on another side of said aperture in said tip. 15. The probe according to claim 14 wherein said first radioactive means includes Fe 55 material and said second radioactive means includes Cd 109 material.
047325275	summary	In a nuclear plant, areas containing hazardous materials are separated by thick walls and slabs from areas where personnel may be present. The hazardous area is generally termed the active region of the plant. The radioactive products which may be present in this zone include gamma-emitters, which justify the thick containment walls, and/or alpha and beta emitters, considered to be contaminants. All equipment leaving an active region is considered contaminated and must be placed behind a thick wall (to protect against gamma rays) and within a sealed enclosure (to avoid spreading the contamination). When equipment in the active region must be replaced or taken out and transferred to a maintenance cell, or conversely, when it is desired to return same to its working location, full confinement must be maintained throughout the transfer procedure to avoid spreading the contamination. It is also crucial to maintain a sufficient wall thickness as a biological shield between said equipment and the plant personnel as the equipment is transferred to the maintenance or repair cell. This invention concerns such a procedure for transferring an "object" such as a piece of equipment, a machine component, a waste drum and the like from the active region or containment to the clean region, for transporting or conveying said object within the clean region and transferring it back from the clean region to the active region without ever departing from containment and biological safety requirements. A first feature of the invention consists in using, for the purpose of maintaining confinement, a two-part device including a stationary pre-enclosure and a movable "transfer" enclosure. The pre-enclosure is kept in place throughout the time required to extract an object, remove it to a cell and either bring back another object or return the same object after repair to the active region or containment. The invention will be described in terms of an active cell accessible through its top, with transfer operations carried out vertically (over-the-wall handling). However, the same procedure applies to all geometrical arrangements such as through-the-wall transfer or through-the-floor. This description will thus concern an active cell the ceiling whereof is a horizontal slab with a plug therein. The top surface of the plug is clean and must be kept clean. The inside face of the plug is within the active region and is considered contaminated and contaminating. The object, at least at the time of transfer, is suspended from the plug. The object's dimensions are assumed to be such that, on lifting the plug, said object can be removed through the unplugged opening.
	abstract	The present invention provides for a composition comprising an inorganic scintillator comprising an optionally lanthanide-doped barium mixed halide, useful for detecting nuclear material.
045049645	summary	Technical Field The invention relates to systems for generating X-rays from plasma, and usable in fine line lithography for semiconductors. BACKGROUND The invention arose from efforts to develop an X-ray system for use in manufacturing microelectronic circuits. With the ever increasing miniaturization of semiconductor integrated circuitry, optical lithography does not afford the necessary resolution due to diffraction effects along mask lines. X-ray lithography provides greater resolution due to the shorter wave length of X-rays. Various types of X-ray sources are known for use in X-ray lithography. These systems are costly, and have not yet achieved a consistently high level of performance and intensity necessary for high production rate lithography. Various of these systems are cumbersome, and are not amenable to repetitive manufacturing sequences. In the conventional type of X-ray system, a metal target is continuously bombarded by a stream of high energy electrons. Most of the energy is dissipated in the target in the form of heat, while a very small fraction is emitted in the form of relatively high energy X-rays. This type of source system has low intensity and low production rates. The high heat generation requires complicated mechanical designs to dissipate the heat, such as rotating anodes or high velocity water cooling. In another type of X-ray system, commonly called the gas puff type, a neutral non-ionized gas is pumped in cylindrical form between a pair of electrodes. High current is then passed between the electrodes, which heats and ionizes the gas, thus creating plasma. The high current also causes magnetic field pinching of the plasma to a smaller constricted area, i.e. parallel lines of current create magnetic fields which cause attraction of the current lines towards each other. The magnetic field pinching and compression of the plasma further heats the plasma and causes X-ray emission. In an alternate gas puff type X-ray system, the cold, neutral gas is pre-ionized, for example electrically, or by radio frequency radiation setting up a standing wave which ionizes the gas. This alternate gas puff system affords better performance, but is extremely costly. Mechanical valving or the like is needed for introduction of the gas, as in the original gas puff system, and there is required the additional equipment for the intermediate pre-ionization stage. Also as in the original gas puff system, the X-ray generating material selection is limited by the requirement that the material be a gas. In another known X-ray system, called the exploding wire type, high current explodes and vaporizes a circumferential array of wires to a vapor plasma. The high current also causes magnetic field pinching of the vapor plasma, generating X-rays. The same current which generates the plasma also generates the X-rays. The plasma and the X-rays are generated in the same area and at the same time. A drawback of the exploding wire type X-ray system is its one shot nature. The wires must be replaced after each firing, and the system is thus not amenable to cost effective use in manufacturing sequences. In another type of X-ray system, as shown in McCorkle U.S. Pat. No. 4,201,921, plasma is generated by passing a high current along the inner capillary wall of a hollow tubular insulator. X-rays are generated by directing an electron beam on the plasma. In another known X-ray system, as shown in Nagel et al U.S. Pat. No. 4,184,078, laser beam bombardment of a target generates plasma which emits X-rays. SUMMARY The present invention provides a simple, low cost X-ray system. Though not limited thereto, the invention is particularly suitable for use in X-ray lithography, including cost effective repetitive manufacturing use. The invention provides the combination of plasma generation directly from a solid material target by laser beam impingement, and X-ray generation by passing high current axially through the plasma causing magnetic field radial inward plasma pinching. Both the plasma generation and the X-ray generation are rapidly repeatable and well suited to manufacturing efficiency. This system further enables accurate control of both the plasma generation and the X-ray generation.
	summary
	summary
	description	The present invention claims priority from Japanese application JP 2004-357539 filed on Dec. 10, 2004, the content of which is hereby incorporated by reference on to this application. The present invention relates to an electron beam apparatus, such as an electron microscope, which measures an electromagnetic field in a matter or vacuum using interference of electron beams. The electron holography, or the electron interference microscopy, is a technique of quantitatively measuring an electromagnetic field in a matter or vacuum by measuring a phase shift of an electron beam caused by a specimen, and specifically a technique in which an electron beam generated in an electron source is splitted into a plurality of electron beams by an electron biprism, a splitted electron beam is made to enter the specimen, and the electron beam having transmitted through the specimen is detected, whereby an interference image is acquired. Such a scanning interference electron microscope is disclosed in, for example, Japanese Patent Application Laid-Open No. 8-45465 and Japanese Patent Application Laid-Open No. 9-134687. The electron beam holography method is classified in terms of its system into an interference electron microscopy of the scanning transmission electron microscope (STEM; Scanning Transmission Electron Microscope) type and an interference electron microscopy of the transmission electron microscope (TEM; TranSmission Electron Microscope) type. The interference electron microscopy of the STEM type has the following merits as compared with the interference electron microscopy of the TEM type: (1) The STEM type interference electron microscopy can display a phase image on-line and real-time; (2) It can display simultaneously an analytical image, such as detection of a characteristic X-ray etc. generated by scanning illumination of an electron beam, and an interference image; and (3) Since a spatial resolution is determined by a spot size of a focused electron beam, controllability of spatial resolution is excellent; and the like. An electromagnetic field in the specimen can be estimated by measuring the amount of phase shift of interference fringes by image analysis of a detected interference image, namely the amount of positional shift between positions of constructive interference and of destructive interference. As a technique of measuring the amount of phase shift of interference fringes, for example, there is the method of Leuthner et al. In addition, in the invention disclosed in Japanese Patent Application Laid-Open No. 9-134687, the amount of phase shift is calculated with the method of Leuthner et al. In the Leuthner' method (Th. Leuthner, H. Lichte, and K-H. Herrmann: “STEM-Holography Using the Electron Biprism” Phys. Stat. Sol. A 116, 113. (1989)), a phase image of the specimen is acquired by detecting an electron beam having passed through a grating-type slit with an electron beam intensity detector, and converting an intensity signal of the detected electron beam into phase information. Hereafter, the Leuthner's method will be explained in detail using FIG. 2A and FIG. 2B. FIGS. 2A and 2B are schematic diagrams each showing a comparative relation among interference fringes of electron beams, a slit, and an electron beam intensity detector. In FIGS. 2A and 2B, the reference numeral 46 denotes a slit and 50 denotes an electron beam intensity detector. The numerals 48 and 49 each denote interference fringes of the electron beams which reach the slit. FIG. 2A corresponds to a case where an aperture of the slit coincides with a position of constructive interference and FIG. 2B corresponds to a case where the aperture of the slit coincides with a position of destructive interference. The vertical axis of the interference fringes 48 and 49 corresponds to the intensity of the electron beams. When performing the method of Leuthner et al., first a direction of the interference fringes and a direction of the slit are set in the same direction. Usually, the apparatus user observes the image of interference fringes by visual inspection, and manually adjusts the direction of the interference fringes obtained, the direction of the slit apertures, and a position of the grating-type slit itself. When the interference fringes are detected in a state where the direction of the interference fringes agrees with the direction of the slit, the intensity of the detected electron beam varies depending on positions of constructive interference and of destructive interference relative to the slit. In the case of FIG. 2A, the amount of the electron beams passing through the slit 46 becomes a maximum, and in the case of FIG. 2B, the amount of the electron beams passing through the slit 46 becomes a minimum. Therefore, if the amount of the electron beams detected with the detector 50 is normalized using its maximum and minimum, the amount of the detected electron beams could be converted to a cosine of the amount of phase shift. That is, when the amount of the electron beams of the interference fringes passing through the slit 46 assumes a maximum, the phase shift by the specimen is 0□}2π□Λn, and when the amount of the electron beams of the interference fringes passing through the slit 46 assumes a minimum, the phase shift by the specimen is π□}2□Λn. Generally, a direction of the apertures of the slit 46 and a position of the slit 46 are so adjusted that detected constructive interference and destructive interference assume detection intensities of those formed under the condition that there is no specimen or both of the splitted electron beams pass through a vacuum. Therefore, it becomes possible to display an image having phase information of the specimen by displaying the amount of the electron beams having passed through the slit 46 which is normalized to be a value between a maximum and a minimum as a cosine of the amount of phase shift or further converting the value so obtained into the amount of phase shift between zero and π. An S/N ratio of a scanning phase information image obtained with the scanning interference electron microscope becomes higher with increasing intensity of the detected electron beam intensity. Therefore, it is essential to make the electron interference fringes enter a detector effectively in order to achieve a clear scan image. The conventional scanning interference electron microscope using the method of Leuthner et al. has the following problems. (1) Setting and adjustment are complex and difficult to do. (2) Simultaneous display of a phase image and an amplitude image cannot be performed. (3) The detection efficiency of the electron beams is low. The above (1) problem arises from a fact that a relative direction between the slit and the interference fringes and positions thereof are adjusted manually. Specifically, adjustment to equalize a spacing of apertures of the slit and a spacing of interference fringes and make directions of both spacings agree with each other is done by observing the interference fringes magnified by about 1000 times with imaging lenses with an eye using a fluorescent screen and moving the position and direction of the slit manually. Since the magnification weakens the intensity of interference fringes, the adjustment requires skills and experience and accurate adjustment is difficult. Regarding a problem described in the above (2), since only one detector is used, it is essential to select and display either the amplitude image corresponding to a normal electron microscope image or the phase image, thus simultaneous display being impossible. If the observer is enabled to observe simultaneously structure information obtained from the normal electron microscope image and electromagnetic field information obtained from the phase image, it will give the observer an extra convenience. The above (3) results from a fact that, since electron beams passing through the slit are allowed to enter the detector, a part of the electron beam blocked by the slit is not used. Since the electron beams blocked the slit cannot be used effectively, there is a limit in improving detection sensitivity or detection precision. Although it is possible to capture the whole image of the interference fringes, namely, to detect all the electrons, and process them with a high-speed processor, a time to transfer data to the processor, a time required for arithmetic computing, a time to transfer the data to memory, etc. will become huge, which deprives the STEM type interference electron microscopy of its advantage that a phase image can be displayed real-time. The present invention has its object to provide a scanning interference electron microscope which is easy to set up and adjust and yet highly sensitive. The present invention solves the above-mentioned problems by detecting interference fringes of electron beam with an electron beam detector that consists of one pair of multi pixels. That is, an output of this detector is a 1-dimensional interference fringe image such that a value of each pixel is an integration value of 2-dimensional pixels along one 1-dimensional direction. Moreover, in the present invention, by mounting this detector on an externally controllable rotationable stage, a magnification of the interference fringes and a rotation direction of the detector are automatically adjusted, so that the interference fringes can be detected under conditions of highest efficiency. According to the present invention, the apparatus can detect interference fringes of the electron beams with an asymmetric 2-dimensional detector with integration capability, and adjust them at high-speed and easily, thereby being able to detect them under optimum conditions. Therefore, a scan image of a high S/N ratio can be obtained. Moreover, the use of one pair of detectors enables simultaneous display of the amplitude image and the phase image. Furthermore, unlike the conventional microscopes, this microscope uses no slit, and accordingly the whole electrons constituting the interference fringes can be used, achieving high detection efficiency. In this embodiment, an example where the present invention is applied to STEM will be described. FIG. 1 shows an example of a configuration of an STEM of this embodiment. The STEM of this embodiment comprises an electron gun 38, an illumination system 39, a specimen chamber 40, an imaging system 41, a detection system 42, a control system, etc., as a rough breakdown of the STEM. The electron gun 38 of the STEM of this embodiment consists of an electron source 1, a first anode 2, a second anode 3, an acceleration anode 4, etc. The illumination system 39 consists of an electron biprism 7, a condenser aperture 8, a first condenser lens 19, a second condenser lens 10, a scanning coil 11, an objective lens pre-field 13, an objective lens post-field 16, etc. In addition, although not illustrated specifically, the electron biprism 7 is equipped with an electron biprism fine positioning system 35, and is made movable by this. The imaging system 41 and the detection system 42 consist of a secondary electron detector 12, the stigmator 19, single or multiple imaging lenses 20, a detector 23, a rotationable stage 24, a rotation mechanism 37 of the rotationable stage, etc. In addition to the secondary electron detector 12, the STEM may be equipped with a reflected electron detector. The control system consists of a CPU 25, memory 26, a display 27, D/A converters 28, 29, a signal processor 32, etc. The D/A converter 28 is connected with constituents of the electron optical system and the imaging system through signal transmission line, and a control signal from the CPU 25 is transferred to each constituent through the signal transmission line. Moreover, although not illustrated in the figure, the display 27 is equipped with information input means, such as a keyboard and a mouse, and the system user enters desired information into the system using the information input means as an input interface. There is a case where the CPU 25, the D/A converter 28, and the signal processor 32 may be housed in a single enclosure, each as a part of a control computer. The reference numerals 30 and 31 denote a specimen stage fine positioning system and a specimen stage fine positioning sensor, both of which are attached to the rotationable stage 24. The numeral 37 denotes the rotation mechanism of the rotationable stage for moving the detector 23 and the rotationable stage 24 used in the present invention and replacing the detector 23 to another detector. The numeral 15 denotes a specimen stage, which is movable in an X-Y plane and in a Z-direction by stage drive means indicated by an arrow. First, operations of the electron optical system will be explained. The electron biprism 7 is inserted between the electron source 1 and a first condenser lens 9 and a voltage is applied to this, whereby the electron beam is splitted into two which apparently come from two virtual electron sources 6. The two splitted electron beams are focused with the first condenser lens 9, a second condenser lens 10, and further the objective lens pre-field 13, respectively, to form two micro spots 14 on a plane of a specimen 15. At this time, the two beams are so adjusted that one of the two micro spots transmits through the specimen and the other passes through a vacuum in proximity to the specimen. Note here that a distance of separation of the two micro spots becomes larger in proportion to a voltage applied to the electron biprism 7. In this embodiment, the voltage applied to the electron biprism 7 and the voltage applied to an electron beam deflection coil 11 are gang controlled in response to a magnification of an image to be observed. Although this gang control is automatically done by the CPU 25, naturally the both voltages can be set manually. This gang control is realized by setting up the voltage of the electron biprism in such a way that the distance of separation of the two micro spots on the plane of the specimen which is calculated with both a deflection angle of the electron beam by the electron biprism and the electron optical system assumes either a comparative value of the spot size of the focused electron beam in each magnification power multiplied by a predetermined multiplier or an absolute value in proportion to an inverse of each magnification value. Now, the electron beam having transmitted through the specimen and the electron beam passing though a vacuum overlap on an arbitrary plane below the specimen to generate interference fringes 17, which is magnified with the imaging lenses 20. The magnified interference fringes are recorded by the detector 23 disposed on an observation plane 22. Generally, if the electron beam is scanned with the electron beam deflection coil 11, the whole interference fringes will move. However, if the object plane of the imaging lenses 20 is adjusted to be on a pivot plane 18 of the electron beam, the interference fringes will not move even with electron beam scanning. Here, the “pivot plane” means an electron optics plane that remains immovable even when the electron beam is scanned at a fulcrum of deflection of the electron beam. Here, the imaging lenses 20 may be of one stage or a combination of multi-stage lenses according to resolution of the detector. In this embodiment, the detector 23 is placed on the externally controllable rotationable stage 24. Moreover, the electron biprism 7, the detector 23, and the rotationable stage 24 can be removed from the passage of the electron beam with the help of the fine positioning system 35 and the rotationable stage movement mechanism 37, respectively, so that these components do not hinder operations of the system as a normal scanning transmission electron microscope. Naturally, this microscope can also be used as a special purpose apparatus of the scanning interference electron microscope. Furthermore, the use of the stigmator 19 consisting of a multipole is desirable because an image of the interference fringes is compressed in a direction parallel to the interference fringes and the intensity of the electron beams is enhanced. Next, a method for observing the interference fringes by using the STEM shown in FIG. 1 will be explained using FIG. 3. First, in Step 300, the specimen is placed and held on the specimen stage 15 and carried into the vacuum chamber. Next, in Step 301, a predetermined voltage is applied to the acceleration anode 4 to accelerate the electron beam generated in the electron source 1. In Step 302, field emission current is pulled out by applying suitable voltages to the first anode 2 and the second anode 3. In Step 303, the electron biprism fine positioning system 35 is driven to move the electron biprism 7 to a predetermined position. In Step 304, the electron biprism is adjusted by using a rotational mechanism of the electron biprism and a rotational mechanism of the specimen so that the edge of the specimen becomes parallel to the direction of the electron biprism. In Step 305, the monitor 27 shows a screen used to specify observation magnification, and the apparatus user enters the observation magnification into the apparatus by input means, such as a GUI and a keyboard. The CPU 25 determines a voltage to be applied to the biprism and transfers it to the electron biprism 7 based on the entered observation magnification. The applied voltage determined by the CPU 25 is converted into an analog control signal by the D/A converter 28, and inputted into an unillustrated drive power supply for the electron biprism. Then, Step 306 is executed. Next, in Step 307, adjustment of the magnifying lens and the stigmator 19 is executed. That is, the magnification of the imaging lens 20 is suitably adjusted, the interference fringes 17 on the pivot plane 18 are converted into interference fringes 21 suitably magnified, and further these interference fringes 21 are compressed in a direction parallel to the fringes by adjusting the stigmator 19. In Step 308, formation of the interference fringes 21 is completed in this way, and in Step 309, the interference fringes 21 are grabbed by the apparatus through the detector 23. Since the present invention uses one pair of detectors, what is inputted into the apparatus is a 1-dimensional image 33 or a 1-dimensional image 34. In Step 310, the inputted interference-fringes image is 1-dimensional-Fourier transformed by the processor 32. From the results, the CPU 25 finds a current rotational speed and a peak position of the rotationable stage 24, namely, a spatial frequency giving a peak and a peak intensity, and stores them in the storage device 26. The rotation angle of the rotationable stage 24 is obtained by inputting an output of the specimen stage fine positioning sensor 31 provided in the rotationable stage 24 into the CPU 25 through the A/D converter 29. In Step 311, whether or not the current rotation angle of the rotationable stage 24 is optimum is evaluated. That is, peak intensities corresponding to the respective rotation angles which cover up to 180 degrees are called from the storage device 26, and whether or not a peak intensity corresponding to the current rotation angle is a maximum among them is evaluated. If data corresponding to rotation angles which cover up to 180 degrees is not obtained or if the current rotation angle is not optimum, the flow proceeds to Step 312, where the rotation angle is varied by a previously set angle. This operation is done by the CPU 25 by inputting a signal to the specimen stage fine positioning system 30 provided in the rotationable stage 24 through the D/A converter 28. Here, the flow returns to Step 310 again, and Step 311 and Step 312 are repeated until the rotation angle becomes optimum. If the rotation angle of the detector is determined optimum, whether or not the magnification is optimum is determined in Step 313. That is, the peak positions stored in Step 310 are scanned over a range of previously set spatial frequencies, and whether or not the peak position gives a maximum peak intensity among them is evaluated. If either of the two criteria is not satisfied, the flow goes back to Step 307, where the magnification of the magnifying lens is varied by a previously set value, and Steps 308 to 313 are repeated. If it is determined that the magnification is optimum in Step 313, the flow proceeds to Step 314, where the specimen is observed and the observation is finished in Step 315. FIG. 4 is a diagram showing the imaging system and a main part of the control system of the STEM of FIG. 1, and the operation flow shown in FIG. 3 is executed by constituents shown in FIG. 4. In FIG. 4, the drawing-out reference numeral 63 denotes interference fringes of the electron beam that passes through imaging lenses 62 and reaches a rotationable stage 66. On the rotationable stage 66, one pair of asymmetric 2-dimensional detectors with integration capability 64, 65 are placed and held. Here the “asymmetric 2-dimensional detector with integration capability” means a detector which is made up of a 2-dimensional array of multi-pixels such that a ratio of the number of pixels in one dimension and the number of pixels in the other dimension is equal to or more than two and a value obtained by integrating values of pixels along a dimension having a smaller number of pixels is outputted as a value of each pixel being arrayed along the dimension having a larger number of pixels. This function may be realized with hardware or may be realized with software. The asymmetric 2-dimensional detectors with integration capability 64, 65 are each made up of a large number of electron sensing elements, wherein signals detected by the elements are integrated in a direction along a direction of integration sequentially and is outputted finally as a 1-dimensional image. In this embodiment, the output signal from the asymmetric 2-dimensional detector with integration capability 64 and the output signal from the asymmetric 2-dimensional detector with integration capability 65 are intended to be used for phase detection and for amplitude detection, respectively, and they are designated by symbols P and A in the figure, respectively. The asymmetric 2-dimensional detectors with integration capability 64, 65 are connected with signal transmission lines 67, 68, respectively, being connected to a signal processor 69. A signal which transmits through the transmission line 67 is a signal for phase detection and a signal which transmits through the transmission line 68 is a signal for amplitude detection, and they are designated by DP and DA in FIG. 4, respectively. The signal passing through the signal processor 69 is finally inputted into a CPU 70, subjected to a predetermined operational processing, and subsequently displayed by display means 76. A D/A converter 71 is provided in order to convert the rotation angle information of the rotationable stage 66 from the CPU 70 into an analog control signal and transfer it to a fine rotation mechanism 73 for the rotationable stage 66, and also serves for a stigmator 61 and the imaging lenses 62. An A/D converter 72 is provided in order to convert a signal from a sensor 74 for detecting rotation of the rotationable stage into digital data which the CPU 70 can process. Next, a position adjustment flow of the detector will be explained in detail. Prior to observation of the specimen, it is necessary to form interference fringes first on the detector placed in the center of this scanning interference microscope, i.e., on the electron optical axis. This can be done by mechanical adjustment of the imaging lenses and adjustment of the electron beam deflection coil built in the illumination system. After this was completed, it is necessary to adjust the detector so that the interference fringes may be formed along a longitudinal direction of the detector used in the present invention. (It is necessary to adjust relatively a direction of the interference fringes and a direction of the asymmetric 2-dimensional detector with integration capability.) Here, the direction of the interference fringes of the electron beams and the direction of the detector are defined as follows. That is, the interference fringes of the electron beams are of a pattern in which an intense part and a weak part of the intensity of the electron beams are repeated in a 1-dimensionally direction in a sinusoidal manner. The 1-dimension direction in concern is defined as a direction of the interference fringes. The direction of the asymmetric 2-dimensional detector with integration capability is defined as its longitudinal direction. Adjusting the direction of the interference fringes and the direction of the detector thus defined can be achieved precisely by performing procedures as described below. These procedures will be explained using FIG. 4 similarly. First, the interference fringes 63 are made to be incident on the detector 64 and the detector 65, under appropriate conditions. The 1-dimensional image signal DA 67 which is an output of the phase detecting detector 64 or the 1-dimensional image signal DP 68 which is an output of the amplitude detecting detector 65 is subjected to 1-dimensional fast Fourier transform, namely converted to a spatial frequency spectrum, by the signal processor 69. When the spectrum is displayed on the display 76, if the direction of the interference fringes and the direction of the detector agree with each other, a clear peak 77 is observed in the spectrum. The signal processor 69 may be realized with hardware using a special board, or may be realized by executing software of Fourier transform on the CPU 70. In line with this, rotational angle detection means, such as the angle sensor 74, is provided in the rotationable stage 66, and a rotational angle of the rotationable stage 66 counting from the start of rotation is outputted as an angle signal, which is inputted into the CPU 70 though the A/D converter 72. The CPU 70 displays a phase signal inputted from the signal converter 69 on the display means 76 in synchronization with the angle signal from the A/D converter 72. Then, data showing a dependence of the peak intensity of the spectrum on the rotational angle of the rotationable stage, as shown in a graph 78, on the display means 76. Observing the height of the peak while rotating the detector, the peak intensity assumes a maximum when the angle of the rotation agrees with a best matched direction. The angles at which the peak intensity becomes maximums are determined as optimum arrangement angles of the asymmetric 2-dimensional detectors with integration capability 64, 65, respectively. Determination of the optimum arrangement angle may be selected by the apparatus user, or the apparatus may control adjustment of rotation angle so that the optimum peak is automatically selected. In the case where the apparatus user itself selects the optimum peak, the apparatus is so controlled that, when rotation of the rotationable stage is ended in a range of horizontal direction of the graph 78, the apparatus becomes a state of waiting an entry from the user. When the apparatus becomes the state of waiting an entry, the apparatus allows the apparatus user to select a peak which is considered optimum from the graph 78 displayed on the display screen 76 with input means, such as a mouse and as key board, entering information of the optimum peak into the apparatus. The information of the optimum peak entered by the apparatus user is transferred to the CPU 70, and the CPU 70 reads a rotation angle of the optimum peak from the display image (graph 78) based on the inputted information and forwards the information of the optimum angle to the D/A converter 71. The optimum angle is fed back to the fine rotation mechanism 73 installed in the rotationable stage 66. The control means of the rotationable stage rotates the rotationable stage based on the angle information fed back thereto, and optimizes the arrangement angle of the asymmetric 2-dimensional detectors with integration capability 64, 65. In the case where the apparatus optimizes the arrangement angle of the asymmetric 2-dimensional detectors with integration capability in a fully automatic manner, the CPU 70 automatically reads an optimum peak from the graph 78 and feeds it back to the fine rotation mechanism 73 for the rotationable stage through the D/A converter 71. In this case, it is not necessary to control the apparatus to be in the state of waiting for the user's entry after the end of the rotation of the rotationable stage; automatic reading of the optimum peak may be started just after the end of the rotation. It is needless to say that the apparatus needs to be adjusted in advance so that the center of the detector coincides with the center of the interference fringes before the adjustment of the optimum arrangement angle of the interference fringes described above. Note here that, since the interference fringes are integrated in a direction parallel to the interference fringes in the case of the asymmetric 2-dimensional detector with integration capability used in this embodiment, a clear peak can be obtained only when the direction of the interference fringes agrees with the direction of the detector in a highly precise manner; therefore, the direction of the two can be adjusted precisely. Moreover, integration of the interference fringes enhances the ratio excellently, and consequently the directions can be adjusted further accurately. By such procedures, it becomes possible to bring the direction of the interference fringes and the direction of the detector into agreement with each other with high precision. A next important adjustment subject is adjustment between a fringe spacing of the interference fringes, or a magnification of the interference fringes, and a pixel size of the detector. FIG. 5 shows several electron interference fringes formed under fixed conditions recorded on a high-resolution film while varying only the magnification. A film whose resolution allows interference fringes having a fringe spacing of about 3 μm at a minimum to be recorded was used. When the magnification is so reduced that the interference fringe spacing becomes 33 μm, 13 μm, 9 μm, 5.5 μm, and 3.8 μm on the film, an exposure time necessary to achieve the same optical density becomes smaller as 240 sec, 120 sec, 60 sec, 8 sec, and 4 sec, respectively. Observing a profile (FIG. 5, right row) obtained by integrating the interference fringes recorded under these conditions in a direction parallel to the fringe of the interference fringes, it is found that a highest contrast is achieved with an interference fringe spacing of 5.5 μm and an exposure time of 8 sec. As shown in this example, it is preferable to record the interference fringes with as small a magnification as possible. However, when the interference fringe spacing comes close to a resolution limit of a detector (in this case, the film), the contrast blurs because constructive interference and destructive interference cannot be recorded. The above fact teaches that in order to detect the interference fringes, it is recommended to magnify the interference fringes so as to have a best fringe spacing which complies with the resolution of the detector. So, in this embodiment, the magnifying lens 62 and the stigmator 61 are controlled by the CPU 70 and the D/A converter 71, as shown in FIG. 4, and a position and the height of the peak in the 1-dimensional Fourier conversion of the interference fringes are adjusted by the same procedures as was used in adjusting a direction of the detector. The adjustment is done by the following procedures. First, the magnifying lens 62 is so adjusted that the electron beam is incident on the center of the detector under a condition that the spacing of the interference fringes 63 is sufficiently large as compared to a size of a pixel of the detector 64 or detector 65. Then, the stigmator 61 is so adjusted that the interference fringes are compressed in a direction parallel to the interference fringes. After that, the direction of the detectors is optimized with respect to the direction of the interference fringes by the procedures described above. At this time, a peak value in a spatial frequency spectrum of the interference fringes under the optimum conditions is stored in a storage device 75. Next, the magnification of the magnifying lens 62 is made small, the same procedures are repeated, and a peak in the spectrum corresponding to a current value of the magnifying lens 62 is stored sequentially. Subsequently, the current value of the magnifying lens 62 is plotted on the horizontal axis and the peak value in the spectrum is plotted on the vertical axis. Since the peak value in the spectrum becomes a maximum at an optimum magnification of the magnifying lens 62, the magnifying lens 62 and the stigmator 61 are set to this condition and the adjustment is finished. By the above procedures, the interference fringes can be detected under the optimum conditions. Naturally, these adjustment procedures can be put in a program and be performed automatically. It goes without saying that the procedures of matching the direction of the interference fringes described above can be realized by finely tuning a mechanism for rotating the electron biprism in a plane vertical to the direction of the electron beam, except for the rotation of the detector. Now, procedures of obtaining both the amplitude image and the phase image simultaneously after setting detection of the interference fringes to be under the optimum conditions in this way will be explained using FIG. 6. Note that, in FIG. 6, a rectangular pattern that is hatched is a conceptual image of a digital output signal. First, for the amplitude image, an output signal by the asymmetric 2-dimensional detector with integration capability is obtained in the absence of specimen or under a condition that both of the two spots splitted by the electron biprism pass through a vacuum on the plane of specimen and stored in the storage means, which is designated as DA-in 80. Next, under a condition that one of the spots transmits through the specimen and the other passes through a vacuum, an output signal DA-in 79 of the asymmetric 2-dimensional detector with integration capability is obtained. The two output signals are added by an adder 81 to obtain an output signal, which is designated as DA-n+A0-in 82. Further, this is integrated for all the pixels to obtain an output signal, which is designated as IA-out 83. This output signal IA-OUT 83 is equivalent to an amplitude image of a normal electron microscope. In this embodiment, in parallel to the acquisition of the amplitude image, a phase image is obtained simultaneously using another asymmetric 2-dimensional detector with integration capability. That is, under conditions that there is no specimen or the two spots splitted by the electron biprism both pass through a vacuum on the plane of specimen as in the case of the acquisition of the amplitude image, an output signal of the asymmetric 2-dimensional detector with integration capability is obtained and recorded as DP0-in85. Next, under conditions that one of the spots transmits through the specimen and the other spot passes through a vacuum, an output signal DP-in 84 of the asymmetric 2-dimensional detector with integration capability is acquired, and the two signals are added by an adder 86 to obtain a 1-dimensional image DP-in+P0-in 87. Using a processor 89 which keeps values equal to or larger than a certain threshold of the pixels among the pixels constituting the 1-dimensional image DP-in+P0+in 87 and sets the values of other pixels to zero, a 1-dimensional image DP-OUT 90 composed of values of the pixels each having a value equal to or larger than the certain threshold is obtained. Here, for the threshold, a 1-dimensional image DP-TH 88 that is set arbitrarily by the user may be used. Alternatively, a 1-dimensional image DP-TH 88 each of whose pixels has an average value of the output signal IA-OUT 83 of the amplitude image. Each pixel value of the 1-dimensional image DP-OUT 90 thus set up is integrated over all the pixels to obtain an output signal, which is designated as IP-OUT 91. The two kinds of output signals IA-OUT 83 and IP-OUT 91 obtained in the above may be displayed, as they are, as the amplitude image and the phase image on the screen, respectively. Alternatively, as shown in the bottom of FIG. 6, another signal IP-NORMALIZED 92 may be generated from the two signals and displayed as a new phase image. Here, a computing equation used to covert the signal is given by the following expression.IP-NORMALIZED=(IP-OUT−IA-OUT)/IA-OUTThis signal IP-NORMALIZED 92 becomes an output signal of the image corresponding to the cosine of a phase. Naturally, the output signal may be further converted to obtain an output signal of an image that corresponds to a value of the phase. Incidentally, the storage device, the adder 86, and the processor 89 correspond to the storage device 26, the CPU 25, and the signal processor 32, respectively, in the configuration of the STEM shown in FIG. 1. In this way, in this embodiment, the amplitude image and the phase image can be acquired simultaneously. In order to display the two images simultaneously, the two may be displayed independently on the screen of the display. Alternatively, a signal IA-OUT 94 of the amplitude is brought into correspondence with a Lightness value, as shown in FIG. 7, a signal IP-OUT 95 corresponding to the cosine of a phase or a signal obtained by further converting it into a phase value is brought into correspondence with a Hue value of the HLS color model, and this is converted to the RGB model with a converter 96 to be displayed in a display 97. Thus, the phase image and amplitude image are simultaneously displayed, overlaying one image on the other in the same display. The simultaneous display of the two images makes possible for the user to observe both a structure which is recognizable from the amplitude image and a potential distribution or a magnetic field distribution of the sample which is recognizable from the phase image, thereby making it easy to observe the both images being correlated with each other. FIGS. 8A to 8D show another example of this embodiment. In this example, this embodiment is applied to dopant profile evaluation of a semi-conductor transistor. First, a voltage is applied to the electron biprism in the absence of specimen, the interference fringes magnified with an imaging lens is made to be incident on the asymmetric 2-dimensional detector with integration capability. After that, the electron beams are deflected with a deflection coil, and thereby the electron beams take an arrangement as shown in FIG. 8A. Here, one of two splitted electron beam spots 98, 99 is adjusted to pass through a vacuum and the other is adjusted to transmit through the specimen. Next, the two splitted electron beam spots are scanned in a direction of scanning 100, and at each predetermined scanning distance, the interference fringes are acquired. FIG. 8B shows a comparative relation between the specimen and the electron beam spot at the time when the electron beam is scanned as far as the central portion of a semiconductor thin film specimen. Further the scanning is continued so as to complete the scanning of the electron beam as far as a desired range (FIG. 8C), and subsequently an image corresponding to a dopant profile in an area 107 can be obtained from an image which corresponds to sequentially acquired cosine values of phases of the electron beam or values obtained by converting them into phases. FIGS. 9A to 9D show further another example of the embodiment. In this, the STEM is applied to magnetic domain structure evaluation of a magnetic thin film. First, in the absence of specimen, a voltage is applied to the electron biprism and interference fringes magnified with an imaging lens are made to be incident on the asymmetric 2-dimensional detector with integration capability. After that, by deflecting the electron beam with a deflection coil, the electron beams take an arrangement as shown in FIG. 9A. Here, two splitted electron beam spots 108, 109 are so adjusted that one of them passes through a vacuum in proximity to the specimen and the other transmits through the specimen. Next, while the two splitted electron beam spots are being scanned in a direction of scanning 110, the interference fringes are acquired at each predetermined scanning distance. FIG. 9B shows a comparative relation between the specimen and an electron beam spot at a time when the electron beam is scanned as far as the central part of a semiconductor thin film specimen. Further the scanning is continued so as to complete the scanning of the electron beam as far as a desired range (FIG. 9C), and subsequently contour line displays 119, 120 which correspond to a magnetic domain structure 113 in a magnetic thin film 112 and a stray magnetic field in a vacuum 111 in proximity to the specimen can be obtained. Note that, in this embodiment, the apparatus outputs cosine values of phases rather than values obtained by converting the cosine values of phases into phases, whereby a display corresponding to magnetic lines of force can be obtained directly. The present invention relates to a scanning interference electron microscope used for evaluation of electric and magnetic characteristics of a micro domain.
051669625	description	DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS FIG. 1 is a sectional view showing the schematic arrangement of an X-ray mask according to the first embodiment of the present invention. Referring to FIG. 1, reference numeral 11 denotes a partially etched support frame of Si or the like; 12, an X-ray transmitting thin film serving as a mask substrate consisting of SiC; and 13, an X-ray absorber pattern consisting of tungsten or the like. The arrangement of this embodiment is basically identical to that of a conventional example shown in FIG. 10 except that the X-ray transmitting thin film 12 has a three-layered structure of 12a-12b-12a. FIG. 2 shows the section of the X-ray transmitting thin film 12 in an enlarged scale. Referring to FIG. 2, a 1-.mu.m thick SiC film is formed on an Si substrate through a low pressure CVD method, and the ratio of C to Si in two surface portions 12a of the film is larger than that in a central portion 12b. More specifically, as shown in FIG. 3, while source gases were introduced into a reactor 22 containing an Si substrate 21 (support frame 11), the 1 .mu.m thick SiC film 12 was formed on the Si (100) substrate 21, which was heated xto 1,050.degree. C. beforehand. The internal pressure of the reactor 22 was set at 1 kPa. Note that the CVD apparatus is not limited to that of FIG. 3 but may be an apparatus using a reaction tube. The gases used were silane (concentration 3%, hydrogen diluted), acetylene (concentration 10%, hydrogen diluted), and hydrogen. The flow rate of silane was fixed at 67 sccm, and that of hydrogen was fixed at 500 sccm. In the thin film 12 shown in FIG. 2, each portion 12a was formed at an acetylene flow rate of 10 sccm for three minutes, and the portion 12b was formed at an acetylene flow rate of 9 sccm for 39 minutes. Note that a 0.7-.mu.m thick SiC film 14 was also formed on the lower surface of the Si substrate 21 through a low pressure CVD method at a pressure of 1 kPa, a substrate temperature of 1,050.degree. C., a silane flow rate of 100 sccm, an acetylene flow rate of 30 sccm, and a hydrogen flow rate of 450 sccm. Portions except for a 25-mm square region in the central portion of the SiC film on its lower surface were covered with an aluminum etching mask, and the SiC film in the remaining portion was removed by reactive ion etching at an RF applied power of 300 W by using a gas mixture of 25 sccm of carbon tetrafluoride and 20 sccm of oxygen. Subsequently, a portion 16 of the remaining SiC film was used as an etching mask to etch away the Si substrate 21 exposed on the lower surface by using an aqueous 30% solution of potassium hydroxide at about 90.degree. C., thereby forming the support frame 11 described above. The visible light transmittance of the 25-mm square SiC film 12 of the completed mask, which remained in the form of a thin film, was measured by using a spectrophotometer. With the above method, a film having a visible light transmittance of 42% at a wavelength of 633 nm could be obtained. By contrast, the visible light transmittance of a 1-.mu.m thick SiC film formed with the flow rates of silane, acetylene, and hydrogen fixed as in conventional methods is at most 33% for an acetylene flow rate of 10 sccm and 38% for that of 9 sccm. That is, according to this embodiment, it is found that the visible light transmittance can be improved by 10% or more simply by changing the gas flow rate (acetylene) by about 10% during the film formation. In addition, a satisfactory effect could be obtained from the thin film 12 shown in FIG. 2 even when only one portion 12a was formed. As described above, according to this embodiment, the visible light transmittance of the X-ray transmitting thin film 12 serving as a mask substrate can be increased by forming the X-ray transmitting thin film 12 using a stacked film consisting of layers having slightly different compositions of SiC. In this case, since the X-ray transmitting thin film 12 consists of the same material system, the film can be easily formed in a single reactor only by changing the flow rates of source gases in a CVD method. In addition, when an X-ray mask is formed by using this X-ray transmitting thin film 12, an alignment between the mask and a wafer can be performed with a high precision. This makes it possible to realize an X-ray mask which is essential in putting micropatterning of VLSI elements of the next generation using X-ray exposure transfer techniques into practical use. FIG. 4 is a sectional view showing the section of an X-ray transmitting thin film according to the second embodiment of the present invention in an enlarged scale. In this embodiment, a film 12a was also formed in a middle portion of the film 12b previously shown in FIG. 3 at an acetylene flow rate of 10 sccm for three minutes. The other conditions including the entire film thickness of this second embodiment were the same as those of the above first embodiment. Also in this case, as in the first embodiment, an SiC film was also formed on the lower surface and selectively etched, and the remaining SiC film was used as a mask to etch away the Si substrate exposed on the lower surface The visible light transmittance of a 25-mm square SiC film remaining in the form of a thin film was measured by using a spectrophotometer. With the above method, a film having a visible light transmittance of 48% at a wavelength of 633 nm could be obtained. This embodiment is different from the above first embodiment only in that the acetylene flow rate was varied from 9 sccm to 10 sccm in the middle portion. However, the visible light transmittance could be improved by 6% compared with that of the first embodiment. FIG. 5 is a sectional view showing the section of an X-ray transmitting thin film according to the third embodiment of the present invention in an enlarged scale. Also in this embodiment, a low pressure CVD method was used to form a 1-.mu.m thick SiC film on an Si substrate with a plane index of (100), which was heated to 1,050.degree. C. at a pressure of 1 kPa, and a plurality of layers having different compositions were stacked in this film. As in the first embodiment, the gases used were silane (concentration 3%, hydrogen diluted), acetylene (concentration 10%, hydrogen diluted), and hydrogen. While the flow rates of silane and hydrogen were fixed at 67 sccm and 500 sccm, respectively, the film formation was alternately performed at acetylene flow rates of 9.5 sccm and 8 sccm for three minutes each and for a total of 45 minutes such that the uppermost layer and the lowest layer were formed at a flow rate of 9.5 sccm. In the SiC film 12, films 12c were formed at an acetylene flow rate of 9.5 sccm, and films 12d were formed at that of 8 sccm. Also in this case, as in the first embodiment described above, an SiC film was also formed on the lower surface and selectively etched, and the remaining SiC film was used as a mask to etch away the Si substrate exposed on the lower surface. Thereafter, the visible light transmittance of a 25-mm square SiC film remaining in the form of a thin film was measured by using a spectrophotometer. With the above method, a film having a visible light transmittance of 68% at a wavelength of 633 nm could be obtained. This value of visible light transmittance belongs to the highest class among those of usual SiC masks obtained before coated with an antireflection film. That is, according to this embodiment, it is found that the visible light transmittance can be largely improved by 70% or more simply by changing the gas flow rate by about 20% during the film formation. In addition, a higher visible light transmittance can be naturally obtained when an antireflection film consisting of, e.g., silicon oxide, is formed on this film. FIGS. 11A to 11D are views showing a method of forming the X-ray absorber pattern 13 and the support frame 11 common to the above first to third embodiments in the order of steps. First, in accordance with the manufacturing steps described above, an X-ray transmitting thin film 12 having a layer structure according to any of the first to third embodiments is formed on an Si substrate 21. Subsequently, the SiC film 14 is formed on the lower surface of the substrate 21, and the central portion of the SiC film 14 is removed by reactive ion etching. A tungsten (W) film 15 serving as an X-ray absorber is formed on the surface of the SiC film 12 by a sputtering method. A resist is coated on the W film 15, and a predetermined pattern is formed on the resist by using an electron beam writing apparatus. The resultant resist is used as an etching mask to transfer the pattern of the resist onto the W film 15 by reactive ion etching, and the remaining resist is removed to form an X-ray absorber pattern 13. Lastly, a residual portion 16 of the SiC film 14 on the lower surface side of the substrate 21 is used as a mask to etch away the exposed portion of the substrate 21, thereby forming the support frame 11. Detailed conditions for forming the support frame 11 are as described above. FIG. 12 is a view for explaining an exposure method using the X-ray mask according to the present invention common to the first to third embodiments. The manufacture of a semiconductor device requires transfer of several different patterns. Therefore, in order to perform relative positioning between the respective patterns, it is necessary to align a mask 32 and a wafer 34 with a high precision. In the present invention, a diffraction grating-like alignment mark 36 formed on the peripheral portion of the circuit pattern on the mask 32 by using the same material W as the X-ray absorber and a diffraction grating-like alignment mark 38 (essentially the same pattern as the mark 36) formed beforehand on the wafer are used to perform alignment for the gap and the horizontal relative position between the mask 32 and the wafer 34 through a heterodyne method using an He-Ne laser beam (wavelength 633 nm). As shown in FIG. 12, a laser beam 44 including information concerning the alignment between the mask 32 and the wafer 34 reaches a detector 42 after passing through the mask substrate twice. Therefore, the S/N ratio of a signal can be largely improved by using the mask substrate having a high visible light transmittance, and this enables a highly precise alignment. An X-ray resist is spin-coated on the wafer 34 beforehand, and X rays 46 having a wavelength of about 1 nm are radiated from the upper portion of the drawing surface while a feedback operation for the alignment is performed, thereby performing transfer. After the transfer, the resist is developed with an proper developer to form a resist pattern. This resist pattern is used to process an underlying substrate, thus making it possible to manufacture a semiconductor device with a fine pattern. Note that the present invention is not limited to the above embodiments. In each of the above embodiments, the film formation of SiC is performed using silane, acetylene, and hydrogen. However, the same effect can be obtained by using other gases, for example, gases containing Si atom, such as dichlorosilane, trichlorosilane, tetrachlorosilane, silicon tetrafluoride, and disilane, and gas mixtures of these gases, in place of silane; and gases containing C atom, such as methane, ethane, ethylene, and propane, and gas mixtures of these gases, in place of acetylene. In addition, the same effect can be obtained when not only the flow rates of acetylene and hydrogen used to dilute the acetylene but also the flow rate of another gas, such as silane, is simultaneously changed. Conversely, the same effect can be obtained when the flow rate of only one type of gas is varied by using a nondiluted gas. Furthermore, when gases containing halogen groups, such as fluorine, chlorine, and hydrogen chloride, are introduced into a reactor during the film formation, the same effect as when the ratio of C atoms to Si atoms in the source gases is changed can be obtained. This is so because the reactivity and the rate of reaction of the above gases with respect to Si atoms being deposited or the source gas of Si atoms are different from those with respect to C atoms being deposited or the source gas of C atoms. For this reason, by changing the flow rates of the gases containing halogen groups, such as fluorine, chlorine, and hydrogen chloride, during the film formation, the visible light transmittance can be improved as in the above embodiments. This effect of introducing gases containing halogen groups will be described below. A low pressure CVD apparatus was used in the film formation of SiC, and an Si(100) wafer of 3 inches in diameter was used as a substrate. This wafer was not subjected to any special pretreatment after it was purchased but directly placed on an SiC-coated graphite susceptor in the apparatus. The setting of substrate temperature was performed by induction-heating the substrate by using a radio frequency from outside a bell jar. The substrate temperature was measured by using a pyrometer from outside the bell jar, and the film formation was performed at 1,050.degree. C. 1% Hydrogen-diluted silane (SiH.sub.4) and 1% hydrogen-diluted acetylene (C.sub.2 H.sub.2) were used as source gases, hydrogen (H.sub.2) was used as a carrier gas, and hydrogen chloride gas (HCl) was added as described above. The flow rate of each gas was controlled by a mass-flow controller. These gases were exhausted by a dry pump with an exhaust power of 7,000 l/min via an automatic pressure regulating valve interlocked with a capacitance absolute manometer. In order to change the source gas composition ratio, the flow rate of the acetylene gas was changed while the flow rate of the diluted silane gas was fixed at 300 sccm. At this time, the flow rate of the hydrogen gas was simultaneously changed so that the total gas flow rate was kept constant at 580 sccm. The pressure was set at 1 kPa upon film formation. The film thickness was measured by observing the section by using an SEM after the film formation. FIG. 6 is a graph showing the change in film stress as a function of the source gas composition ratio (C/Si). The stress measurement was performed by preparing a substrate whose initial deflection or warp was measured in advance by using a flatness tester before the film formation, and calculating from the difference between the initial warp and the warp of the substrate after the film formation in accordance with the balancing conditions of an elastic material. FIG. 6 reveals that a tensile stress of about 1.times.10.sup.9 dyn/cm.sup.2, which is a desirable value as the stress of an SiC film, can be obtained by addition of HCl; that is, C/Si=1.1 when HCl=3 sccm, and C/Si=0.8 when HCl=7 sccm. When the addition amount of HCl is increased, the peak value of the curve is shifted in the direction along which the C/Si value decreases. The reason for this is that because HCl reacts with Si more strongly than with C, the effective source gas composition ratio on the surface becomes equivalent to that when the C/Si ratio is large. FIG. 7 is a graph showing the change in visible light transmittance as the function of the source gas composition ratio (C/Si). The measurement of the visible light transmittance was performed by using a sample with a film thickness of 1.0 .mu.m. Referring to FIG. 7, each bar of data indicates a transmittance which can be obtained by changes in film thickness. The measurement of the visible light transmittance was performed as follows. First, an SiC film was also formed on the lower surface of a substrate, and only a central 25-mm square region of the film on the lower surface was removed by reactive ion etching using a gas mixture of carbon tetrafluoride (CF.sub.4) and oxygen (O.sub.2). Subsequently, a portion of the Si substrate exposed on the lower surface was removed by etching with an aqueous potassium hydroxide (KOH) solution at about 90.degree. C. Thereafter, the visible light transmittance of the central 25-mm square region of the SiC film remaining in the form of a membrane was measured by a spectrophotometer. As can be seen from FIG. 7, a visible light transmittance of 50% or more, which is desirable as a mask substrate, can be easily obtained by addition of HCl (C/Si=0.7 to 1.0). Especially when the HCl flow rate was 7 sccm and the source gas composition ratio was about 0.8, it is possible to simultaneously achieve the optimum stress described above and the desirable visible light transmittance of 50% or more. When the surface roughness with respect to the source gas composition ratio (C/Si) was measured, the results shown in FIG. 8 were obtained. FIG. 8 reveals that an SiC film with a small surface roughness can be formed by addition of HCl (C/Si=0.7 to 0.8). In particular, a film having a surface roughness of about 15 nm can be obtained by adding 3 sccm of HCl and performing film formation at a composition ratio of C/Si=0.8. A film having a surface roughness of about 20 nm could be obtained even under the film formation conditions (HCl=7 sccm, C/Si=0.8) which satisfy the standard levels of both the stress and the visible light transmittance described above. FIG. 9 is a graph showing the crystal orientation parameter as a function of the source gas composition ratio (C/Si). The HCl flow rate was set at 3 sccm, and the crystal orientation parameter was defined as EQU {I.sub.(111) -I.sub.(200) }/{I.sub.(111) +I.sub.(200) } in accordance with an X-ray diffraction peak strength. As is apparent from FIGS. 8 and 9, this parameter has a correlation with the surface roughness. From this fact, it is understood that in order to obtain a film with a small surface roughness, a film having an intense (200) reflection, i.e., a strong <100> orientation property need only be formed. The effect of improving the visible light transmittance obtained by introducing the halogen group can be obtained by using an additional impurity source gas such as arsine, phosphine, or diborane as well. That is, each of these impurities As, P, and B, for example, has different substitution reaction properties and different rates of reaction with respect to an Si atom and a C atom. Therefore, it can be expected that an effect equivalent to that obtained by changing the ratio of C atoms to Si atoms in source gases can be obtained by changing the flow rate of a source gas of the above impurity as in a case wherein the gas containing the halogen group is used. In the above embodiments, SiC is used as the material of an X-ray transmitting thin film formed as a mask substrate. However, the present invention is also effective when another material, e.g., BN or SiN is used as the film. In addition, the film formation method is not limited to a low pressure CVD method but may be an atmospheric pressure CVD method, a plasma CVD method, an ECR-CVD method, a photo excited CVD method, and the like. Furthermore, the present invention can be variously modified without departing from the spirit and scope of the present invention. Additional advantages and modifications will readily occur to those skilled in the art. Therefore, the invention in its broader aspects is not limited to the specific details, and illustrated examples shown and described herein. Accordingly, various modifications may be made without departing from the spirit or scope of the general inventive concept as defined by the appended claims and their equivalents.
	description	The invention relates generally to non-invasive imaging such as single photon emission computed tomography (SPECT) imaging. More particularly, the invention relates to swappable collimators for use in non-invasive imaging. SPECT is used for a wide variety of imaging applications, such as medical imaging. In general, SPECT systems are imaging systems that are configured to generate an image based upon the impact of photons (generated by a nuclear decay event) against a gamma-ray detector. In medical and research contexts, these detected photons may be processed to formulate an image of organs or tissues beneath the skin. To produce an image, one or more detector assemblies may be rotated around a subject. Detector assemblies are typically comprised of various structures working together to receive and process the incoming photons. For instance, the detector assembly may utilize a scintillator assembly (e.g., large sodium iodide scintillator plates) to convert the photons into light for detection by an optical sensor. This scintillator assembly may be coupled by a light guide to multiple photomultiplier tubes (PMTs) or other light sensors that convert the light from the scintillator assembly into an electric signal. In addition to the scintillator assembly-PMT combination, pixilated solid-state direct conversion detectors (e.g., CZT) may also be used to generate electric signals from the impact of the photons. This electric signal can be easily transferred, converted and processed by electronic modules in a data acquisition module to facilitate viewing and manipulation by clinicians. Typically, SPECT systems further include a collimator assembly that may be attached to the front of the gamma-ray detector. In general, the collimator assembly is designed to absorb photons such that only photons traveling in certain directions impact the detector assembly. For example, multi-hole collimators comprised of multiple, small-diameter channels separated by lead septa have been used. With these multi-hole collimators, photons that are not traveling through the channels in a direction generally parallel to the lead septa are absorbed. In addition, while parallel-hole collimators are typically used, collimators also may have converging holes for image magnification or diverging holes for minifying the image. For improved resolution, a pinhole aperture collimator may be used. Pinhole aperture collimators are generally collimators with one or more small pinhole apertures therein. By way of example, an improved image resolution may be obtained with a pinhole aperture collimator, e.g., if the subject is closer to the pinhole than the pinhole is to the gamma-ray detector. SPECT systems may be used for a variety of different applications each of which may require different resolutions and sensitivities. By way of example, small organ imaging may require higher resolution and lower sensitivity, whereas imaging a large volume (such as for possible lesions) typically may require higher sensitivity with lower resolution. Accordingly, it would be desirable to provide an imaging system with adjustable performance based, for example, on the particular application. In accordance with one embodiment, the present technique provides a method of adjusting performance of an imaging system. The method includes removing a slit aperture collimator from the imaging system. The imaging system includes the slit aperture collimator, at least one of a crossed-slit collimator on a side of the slit aperture collimator or a septa assembly having one or more septa spaced on a side of the slit aperture collimator, and a detector assembly. The detector assembly is configured to detect collimated gamma rays emanating from a subject in a field of view of the imaging system. The method further includes removing either the crossed-slit collimator or the septa assembly from the imaging system. The method further includes inserting a pinhole aperture collimator into the imaging system. In accordance with another embodiment, the present technique provides another method of adjusting performance of an imaging system. The method includes removing a pinhole aperture collimator from the imaging system. The imaging system includes the pinhole aperture collimator and a detector assembly. The detector assembly is configured to detect collimated gamma rays emanating from a subject in a field of view of the imaging system. The method further includes inserting a slit aperture collimator into the imaging system. The method further includes inserting a septa assembly or a crossed-slit collimator into the imaging system. In accordance with another embodiment, the present technique provides an imaging system including a collimator support base and a detector assembly. The collimator support base is configured to interchangeably accept a slit aperture collimator and a pinhole aperture collimator. The slit aperture collimator has either a corresponding septa assembly or a corresponding crossed-slit collimator. The detector assembly is configured to detect collimated gamma rays emanating from a subject in a field of view of the imaging system and generate one or more signals in response to the detected gamma rays. FIG. 1 illustrates an exemplary SPECT system 10 for acquiring and processing image data in accordance with exemplary embodiments of the present technique. In the illustrated embodiment, SPECT system 10 includes a collimator assembly 12 and a detector assembly 14. The SPECT system 10 also includes a control module 16, an image reconstruction and processing module 18, an operator workstation 20, and an image display workstation 22. Each of the aforementioned components will be discussed in greater detail in the sections that follow. As illustrated, a subject support 24 (e.g. a table) may be moved into position in a field of view 26 of the SPECT system 10. In the illustrated embodiment, the subject support 24 is configured to support a subject 28 (e.g., a human patient, a small animal, a plant, a porous object, etc.) in position for scanning. Alternatively, the subject support 24 may be stationary, while the SPECT system 10 may be moved into position around the subject 28 for scanning. Those of ordinary skill in the art will appreciate that the subject 28 may be supported in any suitable position for scanning. By way of example, the subject 28 may be supported in the field of view 26 in a generally vertical position, a generally horizontal position, or any other suitable position (e.g., inclined) for the desired scan. In SPECT imaging, the subject 28 is typically injected with a solution that contains a radioactive tracer. The solution is distributed and absorbed throughout the subject 28 in different degrees, depending on the tracer employed and, in the case of living subjects, the functioning of the organs and tissues. The radioactive tracer emits electromagnetic rays 30 (e.g., photons or gamma quanta) known as “gamma rays” during a nuclear decay event. As previously mentioned, the SPECT system 10 includes the collimator assembly 12 that receives the gamma rays 30 emanating from the field of view 26. The collimator assembly 12 is generally configured to limit and define the direction and angular divergence of the gamma rays 30. In general, the collimator assembly 12 is disposed between the detector assembly 14 and the field of view 26. As will be discussed in more detail below, the collimator assembly 12 may be configured to interchangeably accept a slit aperture collimator and a pinhole aperture collimator so that the performance of the SPECT system 10 may be modified. As will be appreciated by those of ordinary skill in the art, the slit aperture collimator may have a corresponding septa assembly or a corresponding crossed-slit collimator while the pinhole aperture collimator generally may not have a corresponding septa assembly. Accordingly, the collimator assembly 12 generally may contain slit apertures or pinholes apertures therethrough such that gamma rays 30 aligned with either the slit or pinhole apertures pass through the collimator assembly 12. Moreover, the collimator assembly 12 may contain a radiation absorbent material, such as lead or tungsten, for example, so that gamma rays 30 that are not aligned with the slit or pinhole apertures should be at least substantially, if not completely, absorbed by the collimator assembly 12. Referring again to FIG. 1, the collimator assembly 12 extends at least partially around the field of view 26. In exemplary embodiments, the collimator assembly 12 may extend up to about 360° around the field of view 26. By way of example, the collimator assembly 12 may extend from about 180° to about 360° around the field of view 26. The gamma rays 30 that pass through the collimator assembly 12 impact the detector assembly 14. Due to the collimation of the gamma rays 30 by the collimator assembly 12, the detection of the gamma rays 30 may be used to determine the line of response along which each of the gamma rays 30 traveled before impacting the detector assembly 14, allowing localization of each gamma ray's origin to that line. In general, the detector assembly 14 may includes a plurality of detector elements configured to detect the gamma rays 30 emanating from the subject 28 in the field of view 26 and passing through one or more apertures defined by the collimator assembly 12. In exemplary embodiments, each of the plurality of detector elements in the detector assembly 14 produces an electrical signal in response to the impact of the gamma rays 30. As will be appreciated by those of ordinary skill in the art, the detector elements of the detector assembly 14 may include any of a variety of suitable materials and/or circuits for detecting the impact of the gamma rays 30. By way of example, the detector elements may include a plurality of solid-state detector elements, which may be provided as one-dimensional or two-dimensional arrays. In another embodiment, the detector elements of the detector assembly 14 may include a scintillation assembly and PMTs or other light sensors. Moreover, the detector elements may be arranged in the detector assembly 14 in any suitable manner. By way of example, the detector assembly 14 may extend at least partially around the field of view 26. In certain embodiments, the detector assembly 14 may include modular detector elements arranged around the field of view 26. Alternatively, the detector assembly 14 may be arranged in a ring that may extend up to about 360° around the field of view 26. In certain exemplary embodiments, the detector assembly 14 may extend from about 180° to about 360° around the field of view 26. The ring of detector elements may include flat panels or curved detector surfaces (e.g., a NaI annulus). In one exemplary embodiment, the ring may comprise in the range from 9-10 solid-state detector panels with each detector panel comprising four detector modules. Those of ordinary skill in the art will appreciate that the ring need not be circular, for example, the detector elements may be arranged in an elliptical ring or be contoured to the body profile of the subject 28. In addition, in certain exemplary embodiments, the detector assembly 14 may be gimbaled on its support base, e.g., so that arbitrary slice angles may be acquired. To acquire multiple lines of response emanating from the subject 28 in the field of view 26 during a scan, the collimator assembly 12 may be configured to rotate about the subject 28 positioned within the field of view 26. In accordance with exemplary embodiments, the collimator assembly 12 may be configured to rotate with respect to the detector assembly 14. By way of example, the detector assembly 14 may be stationary while the collimator assembly 12 may be configured to rotate about the field of view 26. Alternatively, the detector assembly 14 may rotate while the collimator assembly 12 is stationary. In certain exemplary embodiments, the collimator assembly 12 and the detector assembly 14 may both be configured to rotate, either together or independently of one another. Alternatively, if sufficient pinhole apertures and/or slit apertures are provided through the collimator assembly 12, then no rotation may be required. Also, if the slit apertures are orthogonal to the longitudinal axis of the collimator assembly 12 (as illustrated below with respect to FIG. 4), then no rotation may be required. SPECT system 10 further includes a control module 16. In the illustrated embodiment, the control module 16 includes a motor controller 32 and a data acquisition module 34. In general, the motor controller 32 may control the rotational speed and position of the collimator assembly 12, the detector assembly 14, and/or the position of the subject support 26. The data acquisition module 34 may be configured to obtain the signals generated in response to the impact of the gamma rays 30 with the detector assembly 14. For example, the data acquisition module 34 may receive sampled electrical signals from the detector assembly 14 and convert the data to digital signals for subsequent processing by the image reconstruction and processing module 18. Those of ordinary skill in the art will appreciate that any suitable technique for data acquisition may be used with the SPECT system 10. By way of example, the data needed for image reconstruction may be acquired in a list or a frame mode. In one exemplary embodiment of the present technique, gamma ray events (e.g., the impact of gamma rays 30 on the detector assembly 14), gantry motion (e.g., collimator assembly 12 motion and subject support 24 position), and physiological signals (e.g., heart beat and respiration) may be acquired in a list mode. For example, a time-stamp may be associated with each gamma ray event (e.g., energy and position) or by interspersing regular time stamps (e.g., every 1 ms) into the list of gamma ray events. The physiological signals may be included in the list, for example, when they change by a defined amount or with every regular time stamp. In addition, gantry motion may also be included in the event lists, for example, when it changes by a defined amount or with every regular time stamp. The list mode data may be binned by time, gantry motion or physiological gates before reconstruction. List mode may be suitable in exemplary embodiments where the count rate is relatively low and many pixels record no counts at each gantry position or physiological gate. Alternatively, frames and physiological gates may be acquired by moving the gantry in a step-and-shoot manner and storing the number of events in each pixel during each frame time and heart or respiration cycle phase. Frame mode may be suitable, for example, where the count rate is relatively high and most pixels are recording counts at each gantry position or physiological gate. In the illustrated embodiment, the image reconstruction and processing module 18 is coupled to the data acquisition module 34. The signals acquired by the data acquisition module 34 are provided to the image reconstruction and processing module 18 for image reconstruction. The image reconstruction and processing module 34 may include electronic circuitry to provide the drive signals, electronic circuitry to receive acquired signals, and electronic circuitry to condition the acquired signals. Further, the image reconstruction and processing module 34 may include processing to coordinate functions of the SPECT system 10, to implement reconstruction algorithms suitable for reconstruction of the acquired signals. The image reconstruction and processing module 34 may include a digital signal process, memory, a central processing unit (CPU) or the like, for processing the acquired signals. As will be appreciated, the processing may include the use of one or more computers within the image reconstruction and processing module 34. The addition of a separate CPU may provide additional functions for image reconstruction, including, but not limited to, signal processing of data received, and transmission of data to the operator workstation 20 and image display workstation 22. In one embodiment, the CPU may be confined within the image reconstruction and processing module 34, while in another embodiment a CPU may include a stand-alone device that is separate from the image reconstruction and processing module 34. The reconstructed image may be provided to the operator workstation 20. The operator workstation 20 may be utilized by a system operator to provide control instructions to some or all of the described components and for configuring the various operating parameters that aid in data acquisition and image generation. An image display workstation 22 coupled to the operator workstation 20 may be utilized to observe the reconstructed image. It should be further noted that the operator workstation 20 and the image display workstation 22 may be coupled to other output devices, which may include printers and standard or special purpose computer monitors. In general, displays, printers, workstations, and similar devices supplied with the SPECT system 10 may be local to the data acquisition components, or may be remote from these components, such as elsewhere within the institution or hospital, or in an entirely different location, linked to the image acquisition system via one or more configurable networks, such as the Internet, virtual private networks, and so forth. By way of example, the operator workstation 20 and/or the image reconstruction and processing module 18 may be coupled to a remote image display workstation 36 via a network (represented on FIG. 1 as Internet 38). Furthermore, those of ordinary skill in the art will appreciate that any suitable technique for image reconstruction may be used with the SPECT system 10. In one exemplary embodiment, iterative reconstruction (e.g., ordered subsets expectation maximization, OSEM) may be used. Iterative reconstruction may be suitable for certain implementations of the SPECT system 10 due, for example, to its speed and the ability to tradeoff reconstruction resolution and noise by varying the convergence and number of iterations. While in the illustrated embodiment, the control module 16 (including the data acquisition module 34 and the motor controller 32) and the image reconstruction and processing module 18 are shown as being outside the detector assembly 14 and the operator workstation 20. In certain other implementations, some or all of these components may be provided as part of the detector assembly 14, the operator workstation 20 and/or other components of the SPECT system 10. Those of ordinary skill in the art will appreciate that the performance of the SPECT system 10 is at least partially based on the collimator assembly 12 selected for use therewith. For example, pinhole aperture collimators may be used, in certain embodiments, for small field of view imaging. In certain embodiments, when using a pinhole aperture collimator multiple images may be formed with the subject at different positions within the field of view to form a composite whole-body image. However, this technique generally requires more time to acquire than a whole-body image obtained with a slit aperture collimator. Furthermore, the slit and pinhole apertures collimators typically have different spatial resolutions and sensitivities. Different applications, however, may benefit from operating with different resolutions and sensitivities. By way of example, small organ imaging may require higher resolution and lower sensitivity, whereas imaging a large volume (such as for possible lesions) typically may require higher sensitivity with lower resolution. To provide different resolutions and sensitivities, multiple collimator assemblies may be provided for each SPECT system with each collimator assembly having a different performance point. An embodiment of the present technique provides for the exchange of collimator assemblies in the SPECT system 10. More particularly, an embodiment of the present technique provides for a SPECT system 10 configured to interchangeably accept a slit aperture collimator with a corresponding septa assembly or a corresponding crossed-slit aperture collimator and a pinhole aperture collimator. In general, the pinhole aperture collimator is not used with a corresponding septa assembly or crossed-slit aperture collimator, although there may be special circumstances in which one wishes to impose a two-dimensional slice restriction on the three-dimensional character of the pinhole aperture collimator. FIG. 2 illustrates an exemplary SPECT system 10 configured to interchangeably accept a slit aperture collimator 40 with a corresponding septa assembly 42 and a pinhole aperture collimator 44. In the illustrated embodiment, the SPECT system 10 includes a slit aperture collimator 40, a septa assembly 42 and a detector assembly 14. As illustrated, a portion of the detector assembly 14 is removed to illustrate the components of the SPECT system 10, particularly the slit aperture collimator 40 and the septa assembly 42. In one embodiment, the slit aperture collimator 40 may be removeably coupled to a collimator support base 46. The collimator support base 46 may be coupled to a motor (not depicted) to enable rotation of the slit aperture collimator 40. Moreover, the collimator support base 46 may be configured to interchangeably accept the slit aperture collimator 40 and the pinhole aperture collimator 44. Any of a variety of techniques may be used to couple the slit and pinhole aperture collimators 40 and 44 to the collimator support base 46. Further, a septa support assembly (e.g., septa support assembly 81 on FIG. 10) may independently support the septa assembly 42. The septa support assembly may be capable of removing the septa assembly 42 axially from the region of the detector assembly 14 to enable the exchange of slit aperture collimator 40 and the pinhole aperture collimator 44. Alternatively, the septa assembly 42 may be coupled to the slit collimator 40, and thus be capable of co-rotating with it and being removed or inserted with it. To change performance of the SPECT system 10, it may be desired to exchange the slit aperture collimator 40 for the pinhole aperture collimator 44. For example, the pinhole aperture collimator 44 may be selected for use in the SPECT system 10 for small field of view imaging. In general, exchange of the slit aperture collimator 40 for the pinhole aperture collimator 44 may be based on a number of factors, included the particular imaging application. Accordingly, the slit aperture collimator 40 and septa assembly 42 may be removed from the SPECT system 10. In one embodiment, the slit aperture collimator 40 may be de-coupled from the collimator support base 46 and removed from the SPECT system 10. After removal of the slit aperture collimator 40, the pinhole aperture collimator 44 may be inserted into the SPECT system 10. For example, the pinhole aperture collimator 44 may be coupled to the collimator support base 46. As previously mentioned, the slit aperture collimator 40 and the pinhole aperture collimator 44 may be exchanged to change the performance of the SPECT system 10. Accordingly, FIGS. 3-6 describe exemplary slit aperture collimators, having corresponding septa assemblies or crossed-slit collimators, and pinhole aperture collimators that may be exchanged in accordance with embodiments of the present technique. Referring now to FIG. 3, a portion of the SPECT system 10 is illustrated, having a slit aperture collimator 40. In the illustrated embodiments, the SPECT system 10 includes a slit aperture collimator 40 having one or more slit apertures 48 (e.g., slit apertures 48a-48h) therein, a septa assembly 42 having one or more septa 50 spaced on a side of the slit aperture collimator 40 and a detector assembly 14. As illustrated, a portion of the detector assembly 14 is removed to illustrate the components of the SPECT system 10. In exemplary embodiments, the slit aperture collimator 40 may be removeably coupled to the collimator support base 46 to allow exchange of the slit aperture collimator 40 with a pinhole aperture collimator 44. As previously mentioned, it may be desired to exchange the slit aperture collimator 40 and corresponding septa assembly 42 for the pinhole aperture collimator 44 to change the performance of the SPECT system 10. By way of example, the pinhole aperture collimator 44 may be configured to provide a different resolution and/or sensitivity than the slit aperture collimator 40. In general, the slit aperture collimator 40 and the septa assembly 42 may be arranged such that the one or more slit apertures 48 and the one or more septa 50 define one or more pathways for gamma rays emanating from a subject 28 placed in the field of view 26. Gamma rays aligned with one of the slit/septa pathways should pass through the slit aperture collimator 40 and the septa assembly 42 and impact the detector assembly 14, while gamma rays that are not aligned with one of the slit/septa pathways should not pass therethrough. Those of ordinary skill in the art will appreciate that the slit apertures 48 and the septa 50 generally may define a two-dimensional fan-beam imaging geometry. In the illustrated embodiment, the slit aperture collimator 40 has one or more slit apertures 48 therein. As illustrated, the slit apertures 48 may extend in a direction generally parallel to the longitudinal axis 52 of the slit aperture collimator 40. In addition, the slit aperture collimator 40 may include one or more sections spaced around the longitudinal axis 52 thereof such that spaces between the sections define the slit apertures 48. By way of example, the spaced sections may be or include one or more panels 54 (e.g., panels 54a-54h) spaced around the longitudinal axis 52 of the slit aperture collimator 40 to define the slit apertures 48. As illustrated, eight panels 54 are spaced around the longitudinal axis 52 to define eight slit apertures 48. The slit apertures 48 may be referred to as generally one dimensional because the length of the slit apertures 48 is typically long in comparison to their width. For example, the length of the slit apertures 48 may be four, five, ten, or more times greater than their respective width. For support, the panels 54 may be coupled by a mechanical coupling mechanism, such as bands (rings) 56 illustrated in FIG. 3. By way of example, each of the bands 56 may be coupled to each of the panels 54 at the respective ends of the slit aperture collimator 40. As illustrated, the bands 56 may be configured to hold the panels 54 in a generally cylindrical arrangement. Further, while the panels 54 are illustrated in FIG. 3 as curved sections, the present technique encompasses the use of sections that are not curved. In addition, while the panels 54 of the slit aperture collimator 40 are illustrated as separate sections, the present technique encompasses the use of a slit aperture collimator 40 that is unitary. That is, the slit aperture collimator 40 may be fabricated as a solid piece having one or more slit apertures 48 therein. Furthermore, in certain exemplary embodiments, the slit aperture collimator 40 may be constructed as a unitary piece in which the slit apertures 48 are filled by a material that provides mechanical support but that also allows most gamma rays to pass through the slit apertures 48 without interaction. As previously mentioned, one or more septa 50 may be spaced on a side of the slit aperture collimator 40 opposite from the field of view 26. In the illustrated embodiment, each of the septa 50 is generally annular-shaped and spaced along the longitudinal axis 52 of the slit aperture collimator 40. The septa 50 may be arranged, for example, to provide the desired slice information for the SPECT system 10. As illustrated, the septa 50 are generally parallel to each other and generally perpendicular to the longitudinal axis 52 of the slit aperture collimator 40. In this embodiment, the septa 50 may define the transaxial slice information for the SPECT system 10 while the slit apertures 48 provide the longitudinal information. Those of ordinary skill in the art will appreciate that the septa may also be arranged in a generally diverging or converging configuration to alter the slice definition by either magnifying or minifying the axial field of view. In addition, the slit aperture collimator 40 and the septa 50 may each have a thickness sufficient to absorb any gamma rays that do not pass through the slit/septa pathways. By way of example, the slit aperture collimator 40 may have a thickness in the range of from about 10 mm to about 30 mm and the septa 50 may each have a thickness in the range of from about 0.1 mm to about 2 mm. Those of ordinary skill will appreciate that the required thickness to absorb gamma rays depends upon the energy of the gamma rays and the material properties of the slit aperture collimator 40 and the septa 50. Further, the thickness of the slit aperture collimator 40 should provide adequate mechanical strength to support the weight of the collimator and to allow rotation without unpredictable shape distortion. Those of ordinary skill in the art will appreciate that the resolution and sensitivity of the SPECT system 10 may be based in part on the width of the slit apertures 48 and the spacing of the septa 50. In general, the slit apertures 48 and septa 50 may have the same or different widths, with different widths providing different resolving power. By way of example, the slit apertures 48 and/or the spacing between each of the septa 50 may have two or more different widths. In exemplary embodiments, each of the slit apertures 48 and spacing between septa 50 may have a width in the range of from about 0.1 mm to about 10 mm, typically in the range of from about 1 mm to about 5 mm. The various slit apertures 48 and septa spacing may have a distribution of various sizes, and thus differing spatial resolutions and sensitivities. The image reconstruction algorithm should appropriately model the system response of the various apertures. While the preceding discussion of FIG. 3 has described the slit apertures 48 in the slit collimator 40 as extending in a direction generally parallel to the longitudinal axis 52 of the slit collimator 40, and the orthogonal septa 50 spaced along the longitudinal axis 52 of the slit aperture collimator 40, one of ordinary skill in the art will recognize that the present technique may be implemented with collimator assemblies having alternative slit configurations. By way of example, as illustrated by FIG. 4, the slit apertures 48 (e.g., slit apertures 48a-48c) may extend in a direction generally perpendicular to the longitudinal axis 52 of the slit aperture collimator 40 while the septa 50 may extend longitudinally and radially from the slit aperture collimator 40. Alternatively, the slit apertures 48 may extend in a direction generally oblique to the longitudinal axis 52 of the slit aperture collimator 40 and thus describe spirals. Those of ordinary skill in the art will appreciate that the septa assembly 42 may be replaced by a crossed-slit collimator as illustrated in FIG. 5. By way of example, an inner slit aperture collimator 60 is shown with inner slit apertures 64 generally parallel to the longitudinal axis 52. An outer slit collimator (or crossed-slit collimator) 62 is shown with outer slit apertures 66 generally perpendicular to the longitudinal axis 52. As will be discussed in more detail below, the inner and outer slit aperture collimators 60 and 62 should be configured such that the inner slit apertures 64 and the outer slit apertures 66 define one or more apertures therethrough. In exemplary embodiments, the slit direction in the inner slit collimator 60 may be chosen to be perpendicular to the longitudinal axis 52 and the slit direction in the outer slit collimator 62 may be chosen to be generally parallel to the longitudinal axis 52. Further, in exemplary embodiments, the slit directions may be chosen to be oblique to the longitudinal axis 52 and thus describe spirals with the inner and outer slit apertures 64 and 66 in the inner and outer slit aperture collimators 60 and 62 generally orthogonal to each other. Moreover, in exemplary embodiments, the inner and outer slit apertures 64 and 66 may be oblique to each other. Referring now to FIG. 6, a portion of the detector assembly 14, the inner slit aperture collimator 60 and the outer slit aperture collimator 62 are shown to illustrate the apertures defined by the alignment of the inner and outer slit apertures 64 and 66, in accordance with an embodiment of the present technique. As previously mentioned, the inner and outer slit aperture collimators 60 and 62 should be configured such that the inner slit apertures 64 and the outer slit apertures 66 define one or more apertures 68 therethrough. Gamma rays 30 that do not pass through the one or more apertures 68 should be absorbed by the inner and outer slit aperture collimators 60 and 62. In the illustrated embodiment, the apertures 68 are defined by the intersection of the inner slit apertures 64 and the outer slit apertures 66. The apertures 68 allow gamma rays 30 emanating from the field of view 26 to pass through the inner and outer slit aperture collimators 60 and 62 and impact the detector assembly 14. Referring now to FIG. 7, a pinhole aperture collimator 44 having one or more pinhole apertures 70 is illustrated, in accordance with embodiments of the present technique. In the illustrated embodiment, a detector assembly 14 encircles the pinhole aperture collimator 44. As illustrated, a portion of the detector assembly 14 is removed to illustrate the pinhole aperture collimator 44. In exemplary embodiments, the pinhole aperture collimator 44 may be removeably coupled to the collimator support base 46 to allow exchange of the pinhole aperture collimator 44 with a slit aperture collimator 40. As previously mentioned, it may be desired to exchange the pinhole aperture collimator 44 for the slit aperture collimator 40 with corresponding septa assembly 42 or corresponding crossed-slit collimator (such as outer slit aperture collimator 62 on FIG. 5) to change the performance of the SPECT system 10. By way of example, the slit aperture collimator 40 may be configured to provide a different resolution and/or sensitivity than the pinhole aperture collimator 44. In general, gamma rays aligned with one of the pinhole apertures 70 should pass through the pinhole aperture collimator 44, while gamma rays that are not aligned with one of the pinhole apertures 70 should be absorbed by the pinhole aperture collimator 44. Accordingly, the pinhole aperture collimator 44 should have a thickness sufficient to absorb any gamma rays that do not pass through the pinhole apertures 70. By way of example, the pinhole aperture collimator 44 may have a thickness in the range of from about 10 mm to about 30 mm. Those of ordinary skill will appreciate that the required thickness to absorb gamma rays depends upon the energy of the gamma rays and the material properties of the pinhole aperture collimator 44. Further, the thickness of the pinhole aperture collimator 44 should provide adequate mechanical strength to support the weight of the pinhole aperture collimator 44 and to allow rotation without unpredictable shape distortion. In certain exemplary embodiments, the pinhole apertures 70 may be filled with a material that allows most gamma rays to pass through the pinhole apertures 70 without interaction. In the illustrated embodiment, the pinhole apertures 70 in the pinhole aperture collimator 44 are arranged in two rows. The pinhole apertures 70, however, may be arranged in the pinhole aperture collimator 44 in a variety of different configurations. In exemplary embodiments, the pinhole apertures 70 may be arranged in the pinhole aperture collimator 44 in one, two, three, or more rows or in other ordered or pseudo-random patterns. Those of ordinary skill in the art will appreciate that the pinhole apertures 70 generally define a three-dimensional cone-beam imaging geometry. While the pinhole apertures 70 are illustrated as having a generally circular configuration, those of ordinary skill in the art will appreciate that the pinhole apertures 70 may have any suitable geometry. By way of example, the pinhole apertures 70 may be configured as having aperture configurations that are substantially polygonal (e.g., three-sided, four-sided, five-sided, six-sided, and so forth), or substantially curved (e.g., elliptical, circular, and so forth). Those of ordinary skill in the art will appreciate that the resolution and sensitivity of the SPECT system 10 is based in part on the cross-sectional area of the pinhole apertures 70. In general, the pinhole apertures 70 may have the same or different cross-sectional areas. By way of example, the pinhole apertures 70 may have two or more different cross-sectional areas. In exemplary embodiments, each of the pinhole apertures 70 may have a width in the range of from about 0.1 mm to about 10 mm, typically in the range of from about 1 mm to about 5 mm. The various pinhole apertures 70 may have a distribution of various sizes, and thus differing spatial resolutions and sensitivities. The image reconstruction algorithm should appropriately model the system response of the various apertures. While the slit aperture collimator 40 and pinhole aperture collimator 44 are illustrated herein as being generally cylindrically shaped, the present technique encompasses the employment of collimator assemblies that are not generally cylindrically shaped. By way of example, the slit aperture collimator 40 (or pinhole aperture collimator 44) may be or include a flat panel having one or more slit apertures 48 (or pinhole apertures 70) therein. Furthermore, one of ordinary skill in the art will recognize that the collimators and detectors may be combined in modules and positioned to view portions of the field of view. If only a few collimator/detector modules are deployed, then they may be moved to a plurality of positions during image acquisition in order to acquire sufficient data for tomographic image reconstruction. Alternatively, if sufficient collimator/detector modules are deployed, then they may remain stationary during image acquisition and yet acquire sufficient data for tomographic image reconstruction. Furthermore, those of ordinary skill in the art will appreciate that the efficiency of gamma ray detection is based on the number of apertures, such as slit apertures 48 in FIGS. 3, 4, and 5 and pinhole apertures 70 in FIG. 7. By way of example, a collimator assembly configured to have a large number of slit or pinhole apertures 48 and 58 would typically require less or no rotation to obtain a sufficient number of angular projections for image reconstruction. Accordingly, the number of the slit or pinhole apertures 48 and 58 may be adjusted to provide the desired imaging sensitivity for a desired imaging time. Those of ordinary skill in the art will appreciate that the number and spacing of the slit and pinhole apertures 48 and 58 should be chosen with consideration of the efficient utilization of the detector assembly 14 and the performance of the image reconstruction and processing module 18. For example, limited overlap of gamma ray lines of response impacting on the detector assembly 14 may be acceptable. While specific reference in the present discussion is made to a SPECT system, it should be appreciated that the present technique is not intended to be limited to this or any other specific type of imaging system or modality. Rather, exemplary embodiments of the present technique may be used in conjunction with other imaging modalities, e.g., coded-aperture astronomy. In addition, SPECT system 10 may be combined with a second imaging system, such as a CT system or a magnetic resonance imaging (MRI) system. By way of example, the SPECT system 10 may be combined in the same gantry with a CT system. As illustrated in FIG. 8, a SPECT/CT imaging system includes SPECT system 10 and CT system 72. By way of example, the SPECT system 10 and the CT system 72 are shown as separate modules, aligned along a common longitudinal axis, and sharing a single subject support 24. As illustrated by FIG. 9, CT system 72 includes a source 74 of X-ray radiation configured to emit a stream of radiation 76 in the direction of the field of view 26 and an X-ray detector assembly 78 configured to generate one or more signals in response to the stream of radiation. Those of ordinary skill in the art will appreciate that in the third-generation CT configuration illustrated in FIG. 9, the source 74 and the X-ray detector assembly 78 generally rotate in synchrony around the field of view 26 while acquiring a plurality of lines of response passing through the subject 28, so that an X-ray tomographic attenuation image may be reconstructed. Other CT configurations may be employed, including the shared use of at least a portion of the SPECT detector assembly 14 as the X-ray detector assembly 78. Further, the SPECT and CT images may be acquired sequentially, in any order, by repositioning the subject, or concurrently by sharing the detector array. The images generated with the CT system 72 may then be used to generate gamma ray attenuation maps, for example, to calculate attenuation and/or scatter correction during the SPECT image reconstruction. In addition, the CT anatomical images may be combined with the SPECT functional images. FIGS. 10-14 illustrate an exemplary method for removing a septa assembly 42 and a slit aperture collimator 40 from a SPECT system 10 in accordance with one embodiment of the present technique. Referring now to FIG. 10, a SPECT system 10 is illustrated having a detector assembly 14, a slit aperture collimator 40 and a septa assembly 42. As illustrated, the detector assembly 14 may include detector modules 80. The septa assembly 42 may be coupled to a septa support assembly 81. In exemplary embodiments, the septa support assembly 81 may be configured for removal of the septa assembly 42 from the SPECT system 10. In the illustrated embodiment, the septa support assembly 81 includes a septa support arm 82, a septa support table 84 and a rail 86. As illustrated, one end of the septa assembly 42 may be coupled to the septa support arm 82. The bottom of the septa support arm 82 may be coupled to the septa support table 84. The septa support table 84 may be coupled to the rail 86. In exemplary embodiments, the septa support table 84 may be slidably coupled to the rail 86. Referring now to FIG. 1, the septa assembly 42 may be removed from the SPECT system 10 in accordance with one embodiment of the present technique. In the illustrated embodiment, the septa assembly 42 may be removed axially from the region of the SPECT system 10 surrounded by the detector assembly 14. As illustrated, the septa support table 84 may be configured to slide on the rail 86 in the axial direction to enable removal of the septa assembly 42. Removal of the septa assembly 42 may enable the subsequent removal of the slit aperture collimator 40 from the SPECT system 10. Referring now to FIG. 12, the collimator support base 46 may be de-coupled from the frame 87 of the SPECT system 10 in accordance with one embodiment of the present technique. As illustrated, the collimator support base 46 may be coupled to the frame 87 of the SPECT system 10. Any of a variety of suitable mechanisms may be used to couple the collimator support base 46 to the frame 87. In the illustrated embodiment, the collimator support base 46 may include latch 88 into which the latch pin 90 may be inserted. To unlatch the collimator support base 46, the latch pin 90 may be removed from the latch 88 of the collimator support base 46, for example. Once the collimator support base 46 has been unlatched, the collimator support base 46 may be lowered, as illustrated by FIG. 13, to facilitate removal of the slit aperture collimator 40, in accordance with one embodiment of the present technique. As illustrated, lowering the collimator support base 46 may involve rotating the collimator support base 46 about an axis. In one exemplary embodiment, the collimator support base 46 may be coupled to a pair of shock-absorbing arms 92 to, for example, control the lowering of the collimator support base 46. Alternatively, in one exemplary embodiment, the collimator support base 46 may be configured to be moved axially from the region of the SPECT system 10 surrounded by the detector assembly 14. As illustrated by FIG. 14, the slit aperture collimator may be decoupled from the collimator support base 46 and removed from the SPECT system 10. In accordance with embodiments of the present technique, a pinhole aperture collimator (such as pinhole aperture collimator 44 on FIG. 7) may then be inserted into the SPECT system 10. Those of ordinary skill in the art will appreciate that FIGS. 10-14 and the accompanying description describe one suitable method for the removal of a slit aperture collimator 40 from the SPECT system 10. Any of a variety of other suitable methods for the removal of collimators from the SPECT system 10 is encompassed by the present technique. While only certain features of the invention have been illustrated and described herein, many modifications and changes will occur to those skilled in the art. It is, therefore, to be understood that the appended claims are intended to cover all such modifications and changes as fall within the true spirit of the invention.
062467417	claims	1. A fuel assembly for a pressurized water reactor, comprising a lower nozzle to be disposed on a lower core plate of said pressurized water reactor; an upper nozzle having hold-down spring means for holding down said lower nozzle against said lower core plate; a plurality of control rod guide thimbles for guiding control rods extending through said upper nozzle toward said lower core plate; a plurality of supporting grids mounted on said control rod guide thimbles; a number of fuel rods held by said supporting grids in parallel with said control rod guide thimbles; each of said control rod guide thimbles shaped to form a thin tubular dashpot at a lower portion of said control rod for reducing falling speed of said control rods; wherein said dashpot includes a large diameter section having a diameter substantially equal to that of said control rod guide thimble formed at a lower portion of said dashpot, and a small diameter section having a diameter smaller than that of said large diameter section formed at an upper portion of said dashpot, and wherein said small diameter section has an effective length dimensioned so as to fall within a range of 0.03 L to 0.1 L, where L represents an entire length of said control rod guide thimble. 2. A fuel assembly according to claim 1, wherein the effective length of said small diameter section ranges from 0.04 L to 0.06 L. 3. A fuel assembly according to claim 1, wherein the length of said dashpot is dimensioned so as to fall within a range of 0.16 L to 0.18 L.
	summary
	description	As shown in FIG. 1, an object 1 to be X-rayed is located in an X-ray tunnel 2 of an X-ray testing machine 3. Disposed inside the X-ray tunnel 2 is an adjustable diffraction apparatus 10. The diffraction apparatus 10 comprises a collimator/detector arrangement 11 and an X-ray source 12. The collimator/detector arrangement 11 is aimed at an X-ray beam FXxe2x80x2, which is preferably a primary beam emitted as a xe2x80x98pencil beamxe2x80x99 from the X-ray source 12, which is preferably disposed beneath a transport device 4 for the object to be tested in the X-ray tunnel 2. The collimator/detector arrangement (11) is mounted to be adjustable both height-wise and laterally (in the Z and Y directions, respectively, as shown in the figure by arrows) by means of adjustment elements 5, not shown in detail here, connected thereto. The X-ray source 12 is mounted on adjustment elements 6, and can also be adjusted laterally in the Y direction parallel to lateral adjustments of the collimator/detector arrangement (11). The collimation/detector arrangement 11 and the X-ray source 12 are guided synchronously, for which purpose the elements 5 and 6 (which can be, for example, linear guidance with a spindle drive) are actuated at the same time. This movement can be coordinated by a computer 30, not shown in detail. The object 1 to be X-rayed is located, with its items 7, 8, on the transport device 4. If the primary beam FXxe2x80x2 of an X-ray source hits a material, this primary beam FXxe2x80x2 is known to be partially deflected at the crystal-lattice structure of the material as scatter radiation FXxe2x80x3 (as known from Bragg""s Law). Accordingly, the energy spectrum obtained with the energy-sensitive detector yields the crystal structure, and thus the identity of the material. In particular, explosives can be identified and distinguished in this manner. FIG. 2 shows, in detail, the diffraction apparatus 10 according to one embodiment of the present invention for making such X-ray diffraction measurements. The collimator 13 comprises a round slot 15 which defines a predetermined angle "THgr"M in the form of a truncated cone such that, of the scatter radiation emanating from the tested point GM7 of the item 7 in the object, only the components that fall within a specific angle "THgr"M are allowed through to the detector 14. The energy-sensitive detector 14 located behind the collimator 13 detects the scatter radiation FXxe2x80x3 passing through round slot 15 at the scatter angle "THgr"M. To attain a primary beam FXxe2x80x2 from the X-ray source 12, a collimator arrangement 16, for example an apertured-diaphragm arrangement, is mounted in front of the X-ray source 12. The diffraction apparatus should be aligned to the location of the material to be determined in order to make a X-ray diffraction measurement. If the position information in two spatial coordinates (e.g., transport-device position X and lateral position Y) for the items 7 and 8, for example, is known from a lower or prior test stage, the respective missing coordinate e.g. height must be continuously scanned in a measuring sweep. For this purpose, the transportation device 4 and the collimation/detector arrangement 11 travel to an initial position specified for the respective item 7. From there, the measuring sweep is initiated such that the arrangement 11 travels, as necessary, in its height direction and laterally, synchronously with the X-ray source 12, in the direction of the missing coordinate. The signals recorded by the detector during a measuring sweep are stored in one or more energy spectra and compared in a known manner to known energy spectra in the computer 30. This comparison thus yields the material type, particularly for explosive material. If the predetermined points GM7 and GM8 are known in three spatial coordinates, the collimator/detection arrangement 11 and the X-ray source 12 of the diffraction apparatus 10 are displaced and aligned to points GM7 and GM8 one after the other. The scatter radiation FXxe2x80x3 of the X-ray source 12, which is deflected at the crystal lattice of the items 7 or 8, is captured through the round slot 15 of the collimator 13. No further adjustment of the collimator/detection arrangement 11 is necessary during the respective measurement. It is also possible to combine the coordinate information from the lower test stage and the additional, spatial information from the higher stage, possibly supplemented by numerous measurements along numerous measuring paths, and thus determine the volume and the precise spatial position of, for example, the item 8 in the object 1. FIG. 3 illustrates an advantageous embodiment of the annular-slot collimator 13. A central, blind-bore-like opening 17 is preferably integrated into the collimator 13. The opening 17 is closed in the direction toward the detector 14 disposed behind it. A first detection device 21 and, disposed behind it at a defined distance, a second detection device 22, are located in the opening 17. The first detection device 21 is embodied as a detector for relatively lower X-ray energies, and the second detection device 22 is embodied as a detector for higher X-ray energies. This collimator 13 can be used, for example, to additionally perform a conventional material detection through the determination of the average atomic number of the material of the item 7 or 8. The combination of this atomic number and the determined energy spectrum can attain an improved identification of the material of the item 7 or 8. This is of particular significance if the item 7 or 8 contains a highly-absorbent material. Often, lower energies of the central beam FXxe2x80x2 are absorbed in the material, so the corresponding lines of diffraction are missing in the measured energy spectrum. This absence can be reported to the computer with the additional determination of material, and considered in the comparison for the evaluation. In addition, the detection devices 21, 22, which can also comprise, for example, quadrant detectors, can perform a precise spatial orientation (alignment) of the collimation/detection arrangement 11 relative to the X-ray source 12. The alignment itself is effected without an object 1 being located between the collimator/detector arrangement 11 and the X-ray source 12. To this end, the collimator 13 described in conjunction with FIG. 2 has the additional opening 17 with the detection devices, which was not shown in detail in FIG. 2 in order to provide a clear overview. Of course, modifications are possible within the scope of the concept of the invention. For example, other diffraction apparatuses 10 can be used, as are described in the state of the technology, in which case the diffraction apparatus 10, as disclosed in the description, for example, is to be adjustable. The invention now being fully described, it will be apparent to one of ordinary skill in the art that many changes and modifications can be made thereto without departing from the spirit or scope of the invention as set forth herein.
043604960	summary	The invention relates to a cooling system for auxiliary systems of a nuclear installation for the removal of heat from heat exchangers which, firstly are connected on the primary side to lines which may contain radioactive liquids or gases; secondly are disposed in a secured area of the nuclear installation; and thirdly have connections on the secondary side thereof for cooling-liquid lines. For many of these systems it must be ensured that a given temperature (for instance 30.degree. C.) of the cooling liquid provided for the removal of heat from these auxiliary systems is maintained in operation. In addition, the temperature of the cooling liquid must not drop too far so that material stresses in the heat exchangers for the removal of heat from the auxiliary systems are kept low. The cooling system must furthermore be constructed so that even in the event of a failure of the power supply from the normal network, heat removal from the auxiliary systems is still assured. These three requirements have not generally been met in such cooling systems in the past. It is accordingly an object of the invention to provide a cooling system for auxiliary systems of a nuclear installation which overcomes the hereinafore-mentioned shortcomings of the heretofore-known devices of this general type, and which meets all three of these requirements without the need of having to provide a supplemental water line or supply basin. With the foregoing and other objects in view there is provided, in accordance with the invention, a cooling system for auxiliary systems of a nuclear installation for heat removal from heat exchangers, the heat exchangers being connected on the primary side thereof to lines which may contain radioactive liquids or gases, being disposed in a secured area of the nuclear installation, and having connections on the secondary side thereof for cooling liquid lines, comprising an outgoing line for the cooling liquid connected to the connection on the secondary side of the heat exchangers, a dry cooling tower having cooling elements connected to the outgoing line, a return line for the cooling liquid connected to the cooling elements, a refrigeration loop (compressor, condenser and throttle point) having a supplemental heat exchanger with the primary side thereof connected in the return line, a bypass line connected from the outgoing to the return line parallel to the cooling elements and supplemental heat exchanger, and a control valve connected in the bypass line. In accordance with another feature of the invention, the secured area has a boundary, and there is provided a fast-acting shut off valve connected in each of the outgoing and return lines at the boundary. In accordance with a further feature of the invention, there is provided a rising line connected to the return line in the secured area, and an expansion tank being connected to the rising line and disposed in the secured area. In accordance with an added feature of the invention, the cooling liquid is an antifreeze medium which ensures unrestrained operation of the dry cooling tower for safe heat removal from the nuclear installation. In accordance with a concomitant feature of the invention, a plurality of redundant cooling systems are provided for one nuclear installation, the dry cooling towers of the individual cooling systems being disposed at different locations. Other features which are considered as characteristic for the invention are set forth in the appended claims. Although the invention is illustrated and described herein as embodied in a cooling system for auxiliary systems of a nuclear installation, 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.
	summary
054694809	claims	1. A mid-loop operating method carried out during nuclear refueling, characterized in that a round-about pipe conduit is additionally installed between a suction pipe conduit and a discharge pipe conduit of a residual heat removing pump; and a flow rate adjusting valve is installed on said round-about pipe conduit, so that the flow passing through said pump can be maintained at a predetermined normal operation flow level during a mid-loop operation, and round-about flow rate is gradually increased in accordance with the decrease of a residual heat of an atomic reactor, whereby a suction flow rate is maintained at a predetermined level, and introduction of air into the residual heat removing pump is prevented. the facility comprising: a round-about pipe conduit 4 installed between a suction pipe conduit 2 and a discharge pipe conduit 3 of a residual heat removing pump 1; and a flow rate adjusting valve 5 installed on said round-about pipe conduit 4. 2. A mid-loop operating facility for nuclear power plants carried out during nuclear refueling,
049869604	description	DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS Referring to the drawings, it is seen in FIG. 2 that the invention is generally referred to by the numeral 10. As seen in FIG. 1 nuclear reactor vessel 12 is generally comprised of core support structure 14, fuel assemblies 16, shroud 18, inlet nozzle 20, and outlet nozzle 22. The flow of coolant through reactor vessel 12, as symbolized by the arrows, is down annulus 24 via inlet 20, up through and around fuel assemblies 16, and out to steam generators not shown through outlet 22. End fitting 10 is positioned between fuel assemblies 16 and upper grid plate 26. End fitting 10 is generally comprised of main body portion 28, lower plate 30, and spring means 32. Main body portion 28, as seen in side view in FIG. 2 and partial cutaway top view in FIG. 3, is substantially box shaped and has two open opposed sides. An outwardly extending shoulder 34 is provided adjacent the upper edge and substantially at the midpoint of each side of main body portion 28. Shoulders 34 contact upper grid plate 26 in the core of reactor vessel 12 to retain end fitting 10 in the installed position. Upward pressure is provided from fuel assemblies below end fitting 10 and are typically in contact with end fitting 10 by means of guide tubes not shown. Main body portion 28 is provided with slots 38 adjacent the upper portion of each corner for receiving one end of spring means 32. An aperture 39 or handling window is provided on at least one side of main body portion 28 for ease of handling end fitting 10 during initial assembly or reconstitution of fuel assemblies 16. A bore 40 is provided at each corner and along the longitudinal axis of main body portion 28 for slidable mounting relative to lower plate 30. Lower plate 30 is square or rectangular in shape and provided with a guide pin 42 at each corner. In the preferred embodiment, guide pins 42 are threadably engaged in a bore in lower plate 30 and then tack welded in place. A slot 44 is provided in lower plate 30 on each side and offset from the end for receiving one end of spring 32. Lower plate 30 is adapted for attachment to fuel assemblies 16 by guide tubes or other suitable means known in the industry. Bores 48 are provided in lower plate 30 for passage of guide tubes therethrough. Openings 50 are provided in lower plate 30 for flow of coolant therethrough. The pattern shown is for illustrative purposes only and may be adapted to suit the particular application. In the preferred embodiment, spring means 32 is in the shape of a sideways V or hairpin spring. Spring 32 has a first angled end designed to fit into slot 44 in lower plate 30. The second straight end is received in slot 38 in main body portion 28 and welded in place. The first end of spring 32 is offset from the end of lower plate 30. This causes end fitting 10 to have preloaded tension against guide tubes used in fuel assemblies 16 and it is less prone to vibration induced spring failure. In the assembled state, main body portion 28 is resiliently biased away from lower plate 30. In this manner, pressure against lower plate 30 from fuel assemblies 16 is accommodated by compression of spring 32 and movement of lower plate 30 toward main body portion 28. Main body portion 28 is prevented for sliding off guide pins 42 by the use of stop means 46 attached to the end of each guide pin 42 after main body portion 28 has been mounted thereon. It should be noted that any suitable means such as a stop washer or a nut engaged on the end of each guide pin 42 may be used. In operation, end fitting 10 is assembled by welding the second end of spring 32 in slot 38 of main body portion 28. The first end of spring 32 is inserted into slot 44 in lower plate 30 as main body portion 28 is slidably mounted on pins 42. Stop means 46 is then attached to the end of pins 42 to prevent main body portion 28 from sliding off pins 42. During initial assembly of a reactor fuel assembly or during retrofit operations while reconstituting a fuel assembly, lower plate 30 is then attached to the top of a fuel assembly 16 by any suitable means known in the industry. Because many varying and differing embodiments may be made within the scope of the inventive concept herein taught and because many modifications may be made in the embodiment herein detailed in accordance with the descriptive requirement of the law, it is to be understood that the details herein are to be interpreted as illustrative and not in a limiting sense.
	summary
048896799	description	DESCRIPTION OF THE PREFERRED EMBODIMENT In a nuclear reactor it is desirable that steam generator tubes remain leak-tight so that radioactive primary fluid remains everywhere separated from non-radioactive secondary fluid to avoid mingling the radioactive primary fluid with the secondary fluid. Precautionary measures may be necessary to ensure that through-wall cracking and corrosion do not occur in a tube so that the tube remains leak-tight. As described in more detail hereinafter, such precautionary measures may be to expand the tube into engagement with the tube sheet, the support plate or like structure, or to sleeve the tube. Referring to FIG. 1, a steam generator is referred to generally as 20 and comprises a generally cylindrical outer shell 30 having a cylindrical upper portion 40 and a cylindrical lower portion 50. Disposed in upper portion 40 is moisture separating means 54 for separating a steam-water mixture so that entrained water is removed from the steam-water mixture. Disposed in lower portion 50 is an inner shell 55 which is closed at its top end except for a plurality of openings disposed in its top end for allowing passage of the steam-water mixture from inner shell 55 to moisture separating means 54. Inner shell 55 is open at its bottom end, which inner shell 55 defines an annulus 56 between inner shell 55 and outer shell 30. Disposed in inner shell 55 is a vertical steam generator tube bundle 60 having a plurality of vertical, U-shaped steam generator tubes 70 therein, which may be mill annealed, thermally treated Inconel 600. Disposed at various locations along the length of bundle 60 are a plurality of horizontal, circular tube support plates 80, which may be Type 405 stainless steel, having holes 82 therein for receiving each tube 70, for laterally supporting tubes 70 and for reducing flow induced vibration in tubes 70. Additional support for tubes 70 is provided in the U-bend region of bundle 60 by a plurality of anti-vibration bars 85 which may be chrome-plated Inconel. Referring again to FIG. 1, disposed in lower portion 50 and below a bottom-most support plate 86 is a horizontal, circular tube sheet 90 having a plurality of vertical apertures 100 therethrough for receiving the ends of tubes 70 and for supporting the ends of tubes 70, which ends of tubes 70 extend a predetermined distance through apertures 100. Tube sheet 90, which may be a nickel-molybdenum-chromium-vanadium alloy clad in Inconel, is sealingly attached, which may be by welding, around its circumferential edge to a hemispherical channel head 110. Disposed in channel head 110 is a vertical, semi-circular divider plate 120 sealingly attached, which may be by welding, to channel head 110 along the circumferential edge of divider plate 120 and sealingly attached, which may be by welding, to tube sheet 90 along the flat edge of divider plate 120. Divider plate 120 divides channel head 110 into an inlet plenum chamber 130 and an outlet plenum chamber 140. Still referring to FIG. 1, disposed on outer shell 30 below tube sheet 90 are a first inlet nozzle 150 and a first outlet nozzle 160 in fluid communication with inlet plenum chamber 130 and with outlet plenum chamber 140, respectively. A plurality of manway holes 170 are disposed on outer shell 30 below tube sheet 90 for providing access to inlet plenum chamber 130 and outlet plenum chamber 140. Disposed on outer shell 30 above tube bundle 60 is a second inlet nozzle 180, which is connected to a perforated, horizontal and generally toroidal feedring 182 disposed in upper portion 40 for allowing entry of non-radioactive secondary fluid into upper portion 40 through inlet nozzle 180 and through the perforations (not shown) of feedring 182. A second outlet nozzle 190 is disposed on the top of upper portion 40 for exit of steam from steam generator 20. During operation of steam generator 20, radioactive primary fluid, which may be water obtaining a temperature of approximately 620 degrees Fahrenheit, enters inlet plenum chamber 130 through first inlet nozzle 150 and flows through tubes 70 to outlet plenum chamber 140 where the primary fluid exits steam generator 20 through first outlet nozzle 160. Non-radioactive secondary fluid, which may be water, enters feedring 182 through second inlet nozzle 180, which is in fluid communication with feedring 182, and flows downwardly from the perforations of feedring 182 through annulus 56 until the secondary fluid is in fluid communication with tube sheet 90. The secondary fluid then leaves annulus 56 flowing upwardly by natural convection through bundle 60 where the secondary fluid boils and vaporizes into a steam-water mixture due to conductive heat transfer from the primary fluid to the secondary fluid through the walls of tubes 70 which comprise bundle 60 and which function as heat conductors. The steam-water mixture flows upwardly from bundle 60 and is separated by moisture separating means 54 into saturated water and dry saturated steam which may obtain a minimum quality of approximately 99.75 percent. The saturated water flows downwardly from moisture separating means 54 and mixes with the secondary fluid flowing downwardly from feedring 182. Thus, as the secondary fluid enters feedring 182 through second inlet nozzle 180, dry saturated steam exits steam generator 20 through second outlet nozzle 190. In a manner well known in the art of nuclear power production, the dry saturated steam is ultimately transported to a heat sink (not shown) after the dry saturated steam exits steam generator 20 through second outlet nozzle 190. Moreover, in a nuclear reactor the primary fluid is radioactive; therefore, steam generator 20 is designed such that the radioactive primary fluid is nowhere in direct fluid communication with the non-radioactive secondary fluid in order that the secondary fluid is not radioactively contaminated by mingling with the primary fluid. FIRST EMBODIMENT OF THE INVENTION Turning now to FIG. 2, there is illustrated the first embodiment of the present invention, which is an eddy current probe apparatus generally referred to as 200 operatively disposed in a steam generator tube 210 having an interior wall 212, which tube 210 is to be hydraulically diametrically expanded into engagement with tube sheet 90 for closing a gap 220 existing between tube 210 and tube sheet 90. Closing gap 220 mitigates vibration of tube 210 against tube sheet 90 and deposition of contaminates within gap 220, which vibration and deposition of contaminates might otherwise cause surface and volume flaws to develop in tube 210. It will be understood that tube 210 is any one of tubes 70, the plurality of which tubes 70 comprise tube bundle 60. Tube 210, which is received through aperture 100, is attached to tube sheet 90 by a weldment 230 which secures tube 210 within aperture 100. As shown in FIG. 2, probe apparatus 200 may be in fluid communication with a fluid reservoir 240 having fluid, such as water, or gas therein and having a fluid feed conduit 250 connected thereto, which fluid feed conduit 250 extends from fluid reservoir 240 to probe apparatus 200 for supplying fluid to probe apparatus 200 so that probe apparatus 200 is capable of hydraulically diametrically expanding tube 210 into engagement with tube sheet 90. Referring again to FIG. 2, each tube 70, including tube 210, also extends through an associated hole 82 which is formed through each support plate 86. It will be appreciated that probe apparatus 200 may be suitably translated in tube 210 to a location adjacent support plate 86 for hydraulically diametrically expanding tube 210 into engagement with support plate 86 so that a crevice 260 existing between tube 210 and support plate 86 is closed in a manner similar to the manner of closing gap 220. Probe apparatus 200 may be suitably translated in tube 210 by a suitable probe drive apparatus such as that described in U.S. Pat. No. 4,087,748 entitled "Pneumatic Drive Device for a Probe, Particularly an Eddy Current Measuring Probe" and issued in the name of Michel Pigeon et al., the disclosure of which is hereby incorporated by reference. Also illustrated in FIG. 2 is a steam generator tube 270 which has been expanded by the present invention into engagement with tube sheet 90 and support plate 86 for closing gap 220 and crevice 260, respectively. Referring to FIG. 3, probe apparatus 200 is shown disposed in tube 210 for hydraulically diametrically expanding tube 210 into engagement with tube sheet 90 for closing gap 220 which exits between tube sheet 90 and tube 210. Probe apparatus 200 includes a generally cylindrical support body, generally referred to as 280. Support body 280 comprises an elongated generally cylindrical stud 290, which may have a bulged portion 300 of larger diameter at or near the middle portion thereof. Coaxially integrally formed at one end of stud 290 is a generally cylindrical externally threaded first stud end 310, which may be of smaller diameter. Coaxially integrally formed at the other end of stud 290 is a generally cylindrical externally threaded second stud end 320 which may be of the same diameter as first stud end 310. As shown in FIG. 3, stud 290 may generally inwardly taper from bulged portion 300 to first stud end 310 and second stud end 320. Surrounding stud 290 is a generally cylindrical expansible sleeve 322, which may be Pellethane CPR-2103-55D available from The Upjohn Company, CPR Division, located in Torrance, Calif., for diametrically expanding tube 210 into engagement with tube sheet 90. Pellethane is a urethane elastoplastic polymer having hydrolytic stability and high performance at high and low temperatures. This material possesses chemical and solvent resistance and can be used for dynamic and load-bearing applications. It will be appreciated that sleeve 322 may be used also for diametrically expanding tube 210 into engagement with support plate 86 (see FIG. 1). When used for closing gap 220 or 260, sleeve 322 may be approximately 1.3 inches long. As described presently, sleeve 322 has a first sleeve end 324 and a second sleeve end 326 capable of being sealingly compressed against stud 290 for sealingly connecting first sleeve end 324 and second sleeve end 326 to stud 290 so that hydraulic fluid does not escape sleeve 322 when sleeve 322 is inflatably expanded into engagement with tube 210. Referring again to FIG. 3, coaxially threadably attached to first stud end 310 is a generally cylindrical first end cap 330 having a first step bore 340 therein for threadably engaging first stud end 310 and for compressing first sleeve end 324 sealingly against stud 290. First step bore 340 has an internally threaded first smaller portion 350 for threadably engaging the external threads of first stud end 310. First step bore 340 also includes an unthreaded first larger portion 360 of larger diameter for compressing first sleeve end 324 sealingly against stud 290. Similarly, coaxially threadably attached to second stud end 320 is a generally cylindrical second end cap 370 having a second step bore 380 therein. Second step bore 380 includes an internally threaded second smaller portion 390 for threadably engaging the external threads of second stud end 320. Second step bore 380 also includes an unthreaded second larger portion 400 of larger diameter for compressing second sleeve end 326 sealingly against stud 290. First larger portion 360 and second larger portion 400 may outwardly taper from first smaller portion 350 and second smaller portion 390 of first step bore 340 and second step bore 380, respectively, for matingly compressing first sleeve end 324 and second sleeve end 326 against the associated inwardly tapering portions of stud 290. Thus, first sleeve end 324 is sealingly interposed between first larger portion 360 of first step bore 340 and stud 290 for sealingly connecting first sleeve end 324 to stud 290. Similarly, second sleeve end 326 is thus sealingly interposed between second larger portion 400 of second step bore 380 and stud 290 for sealingly connecting second sleeve end 326 to stud 290. In this manner, first sleeve end 324 and second sleeve end 326 are sealingly compressed against stud 290 so that hydraulic fluid does not escape sleeve 322 when sleeve 322 is inflatably expanded into engagement with tube 210. Still referring to FIG. 3, second end cap 370 may include a generally frusto-conical cavity 410 therein in fluid communication with fluid feed conduit 250 at one end thereof and with second stud end 320 at the other end thereof for delivering fluid from fluid feed conduit 250 to second stud end 320. Also formed through stud 290 is a channel 420 which may extend from second stud end 320 to an external surface 430 of stud 290 on the in-board side of sleeve 322 for channeling fluid to and withdrawing fluid from sleeve 322. The fluid referred to immediately above flows from fluid feed conduit 250, through cavity 410, through channel 420, and to sleeve 322, so that sleeve 322 is hydraulically pressurized thereby. As shown in FIG. 3, channel 420 may have an inverted generally L-shaped configuration as it extends from second stud end 320 to external surface 430. Referring to FIG. 4, at least one groove 440, which may have a substantially square-shaped cross-sectional configuration, is formed in stud 290 around the circumference of bulged portion 300 for matingly receiving an associated integral eddy current coil 450 therein, wherein each eddy current coil 450 may have a generally square cross-sectional configuration for matingly lodging in groove 440. Eddy current coil 450 is thereby connected to stud 290 when it is lodged in groove 440. Eddy current coil 450 includes a plurality of electricity conducting wires 460 therein, which may be copper and approximately 0.005 inch in diameter in cross-section, that surround a portion of bulged portion 300 on the in-board side of sleeve 322 for detecting the variations in the electromagnetic characteristics of tubesheet 90 or support plate 86. When eddy current coil 450 is moved in tube 210 to near tube sheet 90, eddy current coil 450 is capable of electromagnetically detecting the presence of a tube sheet inner edge 462. Similarly, when eddy current coil 450 is moved in tube 210 to near support plate 86, eddy current coil 450 is capable of electromagnetically detecting the presence of a support plate inner edge 464 (see FIG. 2). Eddy current coil 450 is thus capable of electromagnetically detecting the location of tube sheet 90 or support plate 86 relative to eddy current coil 450 and sleeve 322. Moreover, eddy current coil 450 is capable of continuously electromagnetically detecting the extent of diametrical expansion of tube 210 by inducing eddy currents in tube 210 and transmitting the variations in the eddy currents to an eddy current measuring unit (not shown). Again referring to FIG. 4, each of the plurality of wires 460 obtains a round cross-sectional configuration and is relatively thin for resisting the pressurization of sleeve 322 when sleeve 322 is hydraulically expanded. The pressurization may be between approximately 12,000 and 20,000 pounds per square inch gauge. However, it will be understood that a plurality of voids 466 are defined between wires 460 when wires 460 are wound about bulged portion 300 in groove 440 and stacked adjacent one to the other. It is undesirable to obtain voids 466 in eddy current coil 450 because the fluid used to pressurize and expand sleeve 322 also may seep into voids 466 between wires 460 thereby pressurizing and expanding eddy current coil 450. Pressurization of eddy current coil 450 in this manner may cause eddy current coil 450 to catastrophically rupture or operate in a manner not intended. Thus, in the preferred embodiment (see FIG. 4) eddy current coil 450 is constructed of relatively thin wire, which may be approximately 0.005 inch in diameter, for reducing the total volume of voids 466 in eddy current coil 450. Wires 460 are arranged in an annular configuration around bulged portion 300 for continuously instantaneously electromagnetically detecting and measuring the extent of diametrical expansion of tube 210. Alternatively, the extent of diametrical expansion of tube 210 may be detected and measured by suitably detecting and measuring the change in pressure in fluid feed conduit 420. It will be appreciated that the thinner each wire 460, the smaller is the total volume of voids defined between adjacent wires 460. Alternatively, the plurality of wires 460 may be selected such that a square cross-sectional configuration (see FIG. 5) is obtained for each wire 460 for resisting the pressurization of eddy current coil 450 when sleeve 322 is hydraulically pressurized and expanded. It will be understood that the use of wires 460 each having a square cross-sectional configuration substantially eliminates voids 466. As shown in FIG. 6, sleeve 322 is capable of diametrically expanding when fluid, which should be transparent to eddy current waves, flows through channel 420 to sleeve 322 for pressurizing sleeve 322. When sufficiently pressurized, sleeve 322 diametrically expands into engagement with tube 210 for diametrically expanding tube 210 into engagement with tube sheet 90 or support plate 86 for closing gap 220 or gap 260, respectively. It will be appreciated that as tube 210 is expanded a predetermined radial distance beyond its elastic limit, tube 210 becomes permanently deformed. The extent of diametrical expansion of sleeve 322 and the extent of diametrical expansion of tube 210 are controlled such that tube 210 expands only to a predetermined radial distance. As described hereinabove, eddy current coil 450 is capable of detecting, by its annular configuration, the extent of diametrical expansion of tube 210 by detecting the variations in the electromagnetic field induced in tube 210 as tube 210 diametrically expands. Of course, eddy current coil 450 is electronically connected to a suitable display device or eddy current measuring unit (not shown) chosen from those commonly available in the art for detecting, monitoring, measuring and displaying the electrical signals from eddy current coil 450. Eddy current coil 450 is also electronically connected to an electrical supply (not shown) for supplying electricity to eddy current coil 450 so that eddy current coil 450 is capable of inducing eddy currents in an adjacent structure. If tube 210 has surface and volume flaws, tube 210 may not remain leak-tight because tube 210 may leak and rupture due to the flaws thereby causing the radioactive primary fluid to mingle with the non-radioactive secondary fluid surrounding tube 210. A typical practice in the art, commonly referred to as sleeving, is to dispose another tube within tube 210 so that radioactive primary fluid can not mingle with the non-radioactive secondary fluid. Therefore, referring to FIG. 7, coaxially disposed in tube 210 is a tubular sealing member 470 for sealing tube 210 by sleeving tube 210 so that sealing member 470 presents a barrier within tube 210 between the primary fluid flowing in tube 210 and the secondary fluid flowing externally around tube 210 and so that sealing member 470 structurally seals and reinforces tube 210 in the region of the flaws. In this manner, primary fluid can not mingle with the secondary fluid when tube 210 has surface and volume flaws because sealing member 470 defines a barrier within tube 210 between the primary fluid and the secondary fluid for preventing the mingling of primary and secondary fluids. However, in order to suitably sleeve tube 210, sealing member 470 should be diametrically expanded into engagement with tube 210. As described presently, probe apparatus 200 is capable of diametrically expanding sealing member 470 into sealing engagement with tube 210 such that tube 210 is suitably sleeved. Referring to FIG. 8, expansible sleeve 322 is shown diametrically expansibly engaging sealing member 470 such that sealing member 470 diametrically expansibly engages tube 210 for sleeving tube 210. When used for sleeving, expansible sleeve 322 may be approximately 4 inches long. When sealing member 470 diametrically expands, tube 210 diametrically expands to like extent due to the diametric force of expansion radially exerted by sealing member 470 against tube 210 as sealing member 470 diametrically expands into engagement with tube 210. It will be appreciated that sealing member 470 and tube 210 are expanded a predetermined radial distance beyond their respective elastic limits so that sealing member 470 and tube 210 are permanently deformed after the expansion process is completed. As described hereinabove, tube 210 may have surface and volume flaws; hence, diametrical expansion of tube 210 excessively beyond its elastic limit may cause tube 210 to catastrophically rupture. Therefore, the extent of diametrical expansion of sleeve 322 and hence the extent of diametrical expansion of sealing member 470 are controlled such that sealing member 470 and tube 210 diametrically expand and deform only to a predetermined radial distance sufficient to avoid catastrophic rupture of tube 210. The extent of diametrical expansion of tube 210 and sealing member 470 are detected by instantaneously and continuously monitoring and measuring the variations in the electromagnetic field induced in sealing member 470 and tube 210 by eddy current coil 450. Alternatively, the extent of diametrical expansion of sealing member 470 and tube 210 may be monitored and measured by suitably monitoring and measuring the change in fluid pressure in fluid feed conduit 420. When sealing member 470 and tube 210 expand a predetermined radial distance, the hydraulic fluid flowing through channel 420 to sleeve 322 is withdrawn from sleeve 322, thereby depressurizing sleeve 322 and deflating sleeve 322, such that sleeve 322 no longer expansibly diametrically engages sealing member 470. The change in fluid pressure as a function of time in fluid feed conduit 420 may be suitably monitored and measured with the aid of a hydraulic expansion unit (not shown) such as HYDROSWAGE brand hydraulic expander manufactured by Haskel, Inc. of Burbank, Calif. This expansion unit includes a low pressure supply system and pressure intensifier, a control box for controlling the operation of the pressure intensifier, and a solenoid valve capable of controlling the flow of hydraulic fluid from the pressure intensifier to sleeve 322 via feed conduit 420. The pressure intensifier may be controlled, via the control box, by a tube expansion control circuit (not shown) which operates in conjunction with a pressure transducer (not shown). The pressure transducer is in turn fluidly connected to the outlet of the pressure intensifier and electrically connected to the expansion control circuit. The control circuit is capable of raising the hydraulic pressure acting against tube 210 or sealing member 470 by a predetermined percentage above the contact pressure after full contact has been made with tube sheet 90 or tube 210, respectively. Thus, the hydraulic expansion unit acting in cooperation with the tube expansion control circuit is capable of monitoring and measuring the hydraulic pressure as a function of time during the expansion process such that leaks or over-pressure conditions which could over-expand tube 210 and/or sealing member 470 are detectable. SECOND EMBODIMENT OF THE INVENTION FIGS. 9 and 10 illustrate the second embodiment of the present invention. As shown in FIGS. 9 and 10, the second embodiment of the invention differs from the first embodiment at least with respect to the number of eddy current coils 450a and 450b and associated grooves 440a and 440b. As shown in FIGS. 9 and 10, two eddy current coils 450a and 450b are each wound in an associated groove 440a and 440b, respectively. Eddy current coils 450a and 450b are separated by a predetermined longitudinal distance- X and, as similar to the first embodiment, are interposed between sleeve 322 and bulged portion 300. As will be described in more detail hereinafter, increasing the number of eddy current coils 450 to two increases the accuracy of determining the location of support plate 86. In the second embodiment, the outside edges of eddy current coils 450a and 450b along the longitudinal axis of stud 90 are spaced at approximately the same distance as the width of support plate 86, which may be approximately 0.75 inch wide, so that the eddy current coils 450a and 450b ; may detect with precision the location of support plate 86 with respect to the longitudinal axis of tube 210. When eddy current coils 450a and 450b are connected to conventional eddy current probe circuitry, such coil spacing yields a lissajous curve with a point intersection whenever the longitudinal edges of eddy current coils 450a and 450b are flush with the top and bottom edges of support plate 86. It will be understood that the spacing X of eddy current coils 450a and 450b does not interfere with the use of eddy current coils 450a and 450b for detecting defects or deposits along the walls of tube 210. In addition, when tube 210 is sleeved, an interference joint is formed at the interface of sealing member 470 and tube 210 when sealing member 470 expansibly engages tube 210. Therefore, it will be understood that spacing of eddy current coils 450a and 450b does not interfere with the use of eddy current coils 450a and 450b for mapping the profile of the interference joint which is formed when sealing member 470 expansibly engages tube 210. When probe apparatus 200 is used to close gap 220 existing between tube sheet 90 and tube 210 or to close gap 260 existing between support plate 86 and tube 210, probe apparatus 200 is translated in tube 210 by operating a suitable probe drive apparatus (not shown) which is connected to probe apparatus 200 and which may be selected from those well known in the art of non-destructive examination. Probe apparatus 200 is translated in tube 210 to a predetermined location at or near tube sheet 90 or support plate 86 for closing gap 220 or gap 260, respectively. The location of tube sheet 90 or support plate 86 is detected by eddy current coil 450, as described hereinabove, when eddy current coil 450 is at or near tube sheet 90 or support plate 86 and when eddy current coil 450 induces an electromagnetic field in tube sheet 90 or support plate 86. When probe apparatus 200 reaches its predetermined location at or near tube sheet 90 or support plate 86, hydraulic fluid is caused to flow from fluid reservoir 240, through fluid feed conduit 250, through cavity 410, to second stud end 320, through channel 420 and to sleeve 322 for pressurizing sleeve 322 and thusly diametrically expanding sleeve 322. As sleeve 322 diametrically expands, sleeve 322 diametrically expansibly engages tube 210, thereby exerting a diametrical force of expansion against tube 210 until tube 210 diametrically outwardly expands beyond its elastic limit into engagement with tube sheet 90 or support plate 86 for permanently closing gap 220 or gap 260, respectively. The diametrical expansion of tube 210 is suitably continuously instantaneously monitored by eddy current coil 450, as specified hereinabove, so that tube 210 is diametrically expanded only to a predetermined radial extent. When tube 210 is expanded to the predetermined radial distance, the hydraulic fluid is suitably withdrawn from sleeve 322, through channel 420, to second stud end 320, through cavity 410, through fluid feed conduit 250, and into fluid reservoir 240 such that sleeve 322 no longer hydraulically expansibly diametrically engages tube 210. When the expansion process for closing gap 220 or 260 described immediately above is completed, probe apparatus 200 is suitably translated out of tube 210 by suitably operating the probe drive apparatus connected to probe apparatus 200. On the other hand, when probe apparatus 200 is used to sleeve tube 210, probe apparatus 200 is translated to a predetermined location in sealing member 470, which is disposed in tube 210 opposite that portion of tube 210 having flaws, by suitably operating the probe drive apparatus which is connected to probe apparatus 200. Probe apparatus 200 is thus translated in sealing member 470 to the predetermined location which may be at that portion of tube 210 having flaws. When probe apparatus 200 reaches its predetermined location, hydraulic fluid is caused to flow from fluid reservoir 240, through fluid feed conduit 250, through cavity 410, to second stud end 320, through channel 520 and to sleeve 322 for pressurizing sleeve 322 and thusly diametrically expanding sleeve 322. As sleeve 322 diametrically expands, sleeve 322 expansibly diametrically engages sealing member 470 which is disposed in tube 210, thereby exerting a diametrical force of expansion against sealing member 470 sufficient to cause sealing member 470 to diametrically expand into sealing engagement with tube 210 for sleeving tube 210. The diametrical expansion of tube 210 is monitored by eddy current coil 450 in the manner described above so that tube 210 diametrically expands only by a predetermined radial extent. The annular configuration of wires 460 comprising eddy current coil 450 allows eddy current coil 450 to instantaneously continuously monitor the diametrical expansion of sealing member 470 and tube 210. After sealing member 470 and tube 210 are expanded the predetermined radial distance, the hydraulic fluid is suitably withdrawn from sleeve 322, through channel 420, to second stud end 320, through cavity 410, through fluid feed conduit 250, and into fluid reservoir 240 such that sleeve 322 no longer hydraulically expansibly diametrically engages sealing member 470. When the sleeving process described immediately above is completed, probe apparatus 200 is suitably translated out of sealing member 470 by suitably operating the probe drive apparatus connected to probe apparatus 200. Of course, it will be understood that modifications and variations may be effected without departing from the spirit and scope of the novel concepts of the present invention. Therefore, this invention provides an eddy current probe apparatus having an expansible sleeve thereon capable of hydraulically diametrically expanding a tube into engagement with a tube sheet or tube support plate and capable of hydraulically diametrically expanding a tubular sealing member into engagement with the interior wall of the tube, which probe is also capable of detecting the location of the tube sheet or support plate and the extent of diametrical expansion of the sealing member and tube.
056573600	claims	1. A reactor container which includes a dry well and a wet well and which is provided with a dry well cooling system, said dry well cooling system comprising: an in-dry-well heat exchanger disposed in the dry well of the reactor container; an in-dry-well blower connected to a primary side of the in-dry-well heat exchanger; first and second circulation pipes connected to the in-dry-well heat exchanger; a normal cooling system connected to the secondary side of the in-dry-well heat exchanger through the circulation pipes and including an equipment cooling pump, an equipment cooling heat exchanger and a seawater pump which are operatively connected to each other; and a standby cooling system connected to said first and second circulation pipes at a position located in between said normal cooling system and said in-dry-well heat exchanger. 2. A reactor container according to claim 1, wherein said standby cooling system comprises a standby cooling pump, a standby cooling heat exchanger connected to said standby cooling pump and a standby seawater pump connected to the standby cooling heat exchanger. 3. A reactor container according to claim 2, wherein a normal power supply and an emergency power supply are connected to said in-dry-well blower, said seawater pump, said standby cooling pump and said standby seawater pump. 4. A reactor container according to claim 1, wherein said standby cooling system comprises a seawater circulation line for directly circulating a seawater therethrough and a seawater pump for pumping up the seawater. 5. A reactor container according to claim 1, wherein said standby cooling system comprises a standby cooling pump and an air cooler connected to each other.
053234294	description	DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS Referring to FIG. 1, a tubular conduit 32 traverses a wall 40 of the reactor vessel of a nuclear reactor, and is subject to stress-corrosion damage due to chemical reaction with the coolant and due to mechanical stresses caused by variations in thermal expansion and pressure. The penetration 32 shown can be one of a plurality of penetrations that traverse the reactor head package, and define pressure fittings 42 whereby mechanical means such as control rod guides and electrical means such as signal lines pass through the pressure barrier between the reactor vessel and the containment building (not shown). The penetrations 32 to be monitored are the conventional pressure sealing penetration tubes otherwise used for the control rod guides, signal lines and the like. At least one of the penetration tubes 32, and preferably a characteristic sample or subset of the penetration tubes is instrumented as a means to assess corrosion of the sample tubes, and also to estimate the corrosion of comparable tubes which are not similarly instrumented. The reactor vessel, which is not shown in detail, is arranged to enclose a quantity of nuclear fuel and a coolant to pass over the fuel for carrying away heat. The coolant normally is water, forming an electrolyte 50 with various ions in solution. The penetrations 32 define conduits subject to stress-corrosion damage due to operation of the nuclear reactor. Although the penetrations 32 are partly exterior to the reactor vessel, they are subjected to chemical action from the coolant or electrolyte 50 as well as stress due to temperature and pressure conditions. Like most metals subjected to such conditions, the walls of the penetrations corrode and can become cracked over time. According to the invention, the ongoing extent of such corrosion is assessed. Information gathered in this manner can be used to track the accumulated corrosion of the penetration tube walls for planning maintenance, and also is useful for detecting conditions of increased stress and corrosion, especially at distinct areas which are vulnerable due to the techniques by which the penetration tubes are attached to the reactor vessel head. As appropriate, the results of the monitoring may involve taking actions to decrease the rate at which the corrosion occurs, e.g., adjusting the chemical makeup of the coolant, or actions to replace or repair penetration tubes which are failing or subject to impending failure. For assessing corrosion, the chemical reactions affecting the penetration wall are detected electrically. At least one electrochemical sensor arrangement or cell 60, and preferably an array 64 of sensors, is mounted in at least one of the penetrations 32 traversing the reactor wall 40. Each sensor cell 60 has a working electrode 72 and a reference electrode 74, insulated from one another and from the wall 82 of the respective penetration tube 32. The working electrode 72 and the reference electrode 74 are placed immediately adjacent an area of the penetration wall 82 to be monitored, the respective cell responding substantially to corrosion in that area. The penetration wall 82 and the electrodes 72, 74 are exposed commonly to the electrolyte 50 during operation of the nuclear reactor. In the preferred embodiment, separately wired electrodes are positioned at different places on the surface of the sensor probe, for monitoring corrosion at a plurality of distinct areas. Axially spaced groups 86 of circumferentially spaced electrode arrays 88 allow separate monitoring of areas around the sensor probe. For example, for monitoring corrosion in the area of passage through the vessel head, which is about 13 cm or 5 inches thick, three to five axially spaced sets of eight circumferentially spaced sensing electrodes can be provided. This arrangement provides for 24 to 40 separately distinguishable corrosion monitoring areas on the internal wall of the penetration tube. It would also be possible to mount an array of sensing electrodes so as to encompass the outer surface of a penetration tube that protrudes into the reactor vessel. However, the exemplary embodiments shown are arranged to monitor corrosion from the inside of the penetration tube, specifically in the area of its attachment to the vessel head, where stresses on the penetration tube make corrosion a problem. The respective electrodes are coupled via signal lines 92 to instrumentation as shown in FIG. 3 for capturing potential and current information and analyzing the data to determine the extent of potential and current noise. The signal lines 92 are routed out of the penetration via a standard pressure fitting 42 operable to maintain the pressure boundary. The signal conductors 92 couple to a detector circuit 94 the voltage and current signals developed at the electrodes 72, 74. The penetration wall 82 is also coupled to the detector circuit for coupling the wall as a working electrode to the detector. The voltage and current signals vary as a function of electrochemical activity leading to stress-corrosion damage of the penetration wall. The extent of corrosion is a function of exchange currents that pass between the electrodes and the electrolyte and produce current and voltage signals between the respective sections of the penetration wall and the electrodes associated therewith. The detector circuit 94 is coupled to the signal conductors 92 and is operable to encode data representing at least one of electrochemical potential, electrochemical impedance, and current passing through the electrolyte between the electrodes and the wall 82. Preferably, the current and voltage noise levels in these signals are assessed. The data thereby developed is read out for assessing deterioration of the penetration wall as a function of the electrochemical activity. There are a number of specific components of electrical activity that can be used to reflect the extent of corrosion of a metal in an electrolyte, and reference can be made to the disclosures mentioned in the foregoing prior art section of the Specification for specific examples, which are hereby incorporated in their entireties. Briefly, the signal conductors 92 are preferably coupled to measurement circuits operable to amplify and encode the electrical potential levels and current dissipation through the wall, the electrode and the electrolyte, which signals are representative of the level of corrosion and the extent of stress-corrosion damage which is accumulating. Some parameters which can be monitored include electrochemical impedance as a function of frequency, galvanic current between the electrode and the wall, electrochemical potential noise and electrochemical current noise. Electrochemical impedance is measured by analyzing the response of the corrosion interface to an applied sinusoidal potential waveform over a range of frequencies, e.g. 0.1 Hz to 10 KHz. This gives information on the resistance/capacitance characteristics of the corroding surface. At the higher frequencies, the impedance can be related to the solution resistance of the electrolyte in the circuit including the penetration wall and the electrode, and can also be related to the extent of accumulated scale and/or similar deposits that are present. The response at lower frequencies can be related to the polarization resistance (or DC impedance value) of the sensor circuit. By subtracting the solution resistance, an accurate representation of the resistance to charge transfer at the corrosion interface can be determined. A lower charge transfer resistance indicates a higher rate of corrosion, and vice versa. Zero resistance ammetry can be used to determine the galvanic current between two electrodes, in this case between the penetration wall and the electrode therein. Normally, the penetration wall and the electrode are of dissimilar metals, which inherently produce a galvanic current when coupled as a cell. This technique can also determine the galvanic current between nominally identical electrodes, which typically are at least different enough to take up slightly different potentials. When the penetration wall and the electrode are coupled via a zero-resistance ammeter, a measurable current flows. The DC value of the coupling current during active corrosion is proportional to the level of corrosion activity then in progress on the electrodes. Electrochemical potential noise is a low level random fluctuation of the electrochemical corrosion potential. The fluctuation is typically of a low amplitude (e.g., less than a millivolt), and a low frequency (e.g., 1 Hz and lower). By measuring the low frequency variation in the electrochemical potential, a time varying signal can be developed that can be correlated against the mode of corrosion attack. For example, pitting corrosion and crevice attack produce clearly distinct signatures in measured electrochemical potential noise. Electrochemical current noise can also be measured. The current noise is similar to the potential noise, except that fluctuations in the coupling current between similar electrodes are recorded and analyzed. An estimate of the overall rate of corrosion can be made from the electrochemical current noise output signal after calibrating the sensor cell empirically, using controlled weight loss exposure measurements. The penetration tube 32 comprises a tubular conduit traversing a wall defined by the head structure 40 of the reactor pressure vessel. The conduit is circular in cross section and is fitted closely into a bore 98 in the vessel head 40. The vessel head is dome shaped. However, the penetration tubes 32 are parallel to one another. As a result, the longitudinal axes of the penetration tubes are disposed at an angle relative to the plane of the wall of the head structure 40 at the penetration 32. As shown in FIG. 1, a result is that the welds 102 which attach the penetration tube 32 to the vessel head 40 are of different sizes and are disposed at different axial positions along the penetration tube 32. The penetration tube 32 can protrude inwardly of the vessel wall, as also shown in FIG. 1. As a result of this mounting arrangement, the penetration tube 32 is especially subject to stress-corrosion cracking adjacent the welds 102, namely in the area where the penetration tube 32 traverses the vessel head 40. The electrodes 72, 74 of the sensing cells 60 are preferably located in this area, where the penetration tube is vulnerable. The penetration 32 used for corrosion monitoring is similar to the other penetrations in the head structure 40 of the reactor pressure vessel. As a result, corrosion of the penetrations generally can be assessed by measuring the corrosion occurring at the instrumented penetration. At least one penetration is instrumented; however it is also possible to instrument a plurality of penetrations as shown in FIG. 4, for separately assessing corrosion at different positions of the vessel head. For example, the instrumented penetrations can be disposed diametrically opposite one another or otherwise spaced around the circumference of the head package, and/or placed at different radial distances from the centerline of the reactor vessel. The instrumented penetrations are provided with pressure fittings 42 in the same manner as the penetrations used for control rod guides or signal lines for other sensors, such as temperature, pressure, nuclear flux and the like. Referring to FIGS. 1 and 2, a plurality of paired electrodes 72, 74 are arranged around the circumference of the probe at axially spaced levels 86. The two electrodes 72, 74 of each pair 110 interact through the electrolyte 50 primarily with the nearest portion of the wall of the penetration tube 32, allowing the arrayed electrodes to develop signals specific to distinct areas. Electrical insulators 112 are interspersed between the electrodes in the probe, the electrodes and insulators being mounted on a supporting post 120 or spring mounting, for example as in the probe of international patent application PCT/GB87/00500. Accordingly, the penetration tube 32 is used as one of the working electrodes for each measurement. The two counter electrodes 72, 74 are positioned along the tube in the vicinity of the potential cracking site, preferably where stresses are highest, e.g., due to the mounting of the penetration tube. As shown schematically in FIG. 3, the current signal is sensed between one electrode 124 and the penetration wall to obtain the coupling current and current noise signals. The voltage signal is sensed between the other electrode 126 and the penetration wall as the other electrode to develop the potential reference signal and potential noise signal. These signals are coupled to digitizing means 130 operable to sample the data, and to a processor 132 that analyzes the sample data numerically. The electrodes used for voltage and current measurements can be fabricated from the same material as the tube. Multiple electrodes can be installed within the penetration tube for any or all of the measurements, e.g., with electrode sets 110 disposed 180.degree. apart around the circumference or spaced by 45.degree. as shown, to detect variations in the corrosion conditions around the inside circumference of the tube, and/or at different axial positions. The specific arrangement of the electrode pairs can be varied with the type and dimensions of the penetration, and with the extent of local area monitoring desired. Similarly, one or a plurality of penetration tubes 32 traversing the vessel head can be instrumented in this manner as shown in FIG. 4, to obtain complete data respecting corrosion of the penetration tubes. Using electrochemical noise measurement to assess stress-corrosion cracking and similar corrosion of the reactor vessel penetrations has several advantages. The probe design can be extremely simple and rugged. Therefore, the probe readily qualifies as safe in the severe environment of the reactor vessel, and reliable long-term measurements can be expected. The electrochemical noise measurement technique is also quite accurate, being capable of detecting crack initiation before visually detectable damage can occur. The technique also detects crack propagation, enabling estimates of crack depth. Since analyzing the noise signal effectively measures the free corrosion potential of the penetration wall and the electrode(s), no polarization of the specimens is required, which could potentially accelerate the corrosion process by providing energy for ion exchange. Analysis of the noise signals not only allows the level of corrosion to be assessed, but also helps to identify the fundamental electrochemical and corrosion processes at work. This information is critical for root cause analyses and failure studies. Finally the technique is excellent for monitoring complex localized corrosion events such as the typical stress-corrosion cracking experienced in reactor vessel penetrations 32 in the area of the reactor vessel wall 40. The circuitry needed to capture and analyze the signals developed from the sensors can include high input impedance amplifiers coupled to data processing means operable to analyze the data for frequency specific data. Preferably, the outputs of the device are coupled to suitable display and/or readout devices for graphic, tabular, summary reporting, and potentially for the triggering of maintenance alarms. The data is also stored for reference, and can be communicated remotely via modem or other communication means. An integrated package of circuitry specifically for electrochemical noise analysis that can be applied to the measurements taken according to the invention is available from CAPCIS MARCH, Limited (CML), Manchester, UK, under the product name DENIS (an acronym for Digital Electrochemical Noise Integrated System). The probe is coupled to the data acquisition and analysis circuitry in the same manner as other process monitoring variables generally. Various intermediate and ultimate elements such as sampling analog to digital converters, multiplexers, cables and other signal lines, data acquisition equipment, and electrochemical noise analyzers can be provided. Preferably, electrochemical noise analysis is employed via software running on a processor coupled to process the data, and the software can include maintenance predictive functions for estimating the remaining useful life of the penetration tubes. The probe is preferably dimensioned and arranged to be compatible with an existing type of reactor head adaptor tube, and preferably is a unitary structure that can simply be inserted in the penetration 32, sealed by the pressure fitting 42 and wired to the detector circuits 94 for operation. The electrode array placed within the tube can be located in the tube internal diameter near the location where cracking has been observed in penetration tubes of this type, namely in the area adjacent the junction with the reactor walls. The probe pressure boundary qualification is also an important concern. Preferably, an end cap design or similar configuration similar to the pressure closures used with existing sensing cable arrangements is employed. The probe is fabricated in accordance with applicable regulatory Code requirements and using Code materials as applicable. The probe is a durable device, comprised substantially of solid metal materials for the support and the electrodes, coupled via electrical insulation 112 so as to maintain electrical isolation of the tube 32 and the electrodes except via ionic current flow through the electrolyte 50. As installed in the head adaptor tube and utilizing suitably durable materials for the probe, its insulating materials and pressure fittings, the probe does not compromise plant safety margins with respect to loads such as pressure, temperature, seismic shock, flow vibration, etc. Thus the probe can be arranged to survive the effects of radiation exposure over a design life that at least exceeds the specifications applicable to the respective penetration tube 32. Preferably, the monitoring system is operated on a continuous basis during normal operation of the plant. Data analysis software provides a continuous on-line indication of corrosion activity for operation monitoring and for maintenance planning purposes. The invention having been disclosed, variations will now be apparent to persons skilled in the art. Whereas the invention is intended to encompass not only the foregoing specific embodiments but a range of equivalent variations as well, reference should be made to the appended claims rather than the foregoing examples, in order to assess the scope of the invention in which exclusive rights are claimed.
	abstract	A motor controller determines a drive speed of a motor for driving a filter such that an area shifting cycle of the filter synchronizes with a cardiac cycle. The filter starts to rotate before radiation emission, and continues to rotate at a constant speed during two successive radiation emissions. An emission controller controls timing of two emissions in accordance with a phase of the filter. The area shifting cycle of the filter and the cardiac cycle are previously synchronized, and thus a high-performance filter switching device for switching or shifting the filter in synchronization with the emission timing becomes unnecessary. Adjusting the area shifting cycle of the filter in accordance with the cardiac cycle realizes two radiation emissions in two successive cardiac cycles. As a result, an exposure time is shortened.
053012127	abstract	The irradiated component (1) comprises at least one tubular wall with its axis vertical and fastened inside a well (2) in a concrete structure (3). The connecting elements between the concrete structure (3) and the component (1) are destroyed, the component (1) is displaced in successive steps in the vertical direction along its axis on the inside of the well (2), and the wall of the component is cut, so as to obtain blocks (26) of irradiated material of the wall, at the upper level of the well (2) of the concrete structure (3) after each displacement of the component (1). The cut blocks (26) are disposed of for the purpose of elimination or storage. The displacement of the component can be obtained by a pull or push on its lower bottom.
	abstract	A system for radioisotope production uses fast-neutron-caused fission of depleted or naturally occurring uranium targets in an irradiation chamber. Fast fission can be enhanced by having neutrons encountering the target undergo scattering or reflection to increase each neutron's probability of causing fission (n, f) reactions in U-238. The U-238 can be deployed as one or more layers sandwiched between layers of neutron-reflecting material, or as rods surrounded by neutron-reflecting material. The gaseous fission products can be withdrawn from the irradiation chamber on a continuous basis, and the radioactive iodine isotopes (including I-131) extracted.
	summary
	description	This application is a Continuation of U.S. application Ser. No. 12/764,163 filed Apr. 21, 2010, which is a Division of U.S. application Ser. No. 12/128,524, filed May 28, 2008, which is a Continuation of and claims the benefit and priority under 35 USC §120 from U.S. Ser. No. 11/348,333, filed Feb. 7, 2006, the entire contents of each of which are incorporated herein by reference. U.S. Ser. No. 11/348,333 is a Division of U.S. Pat. No. 7,139,352, U.S. Ser. No. 09/749,547, issued Nov. 21, 2006, and claims the benefit of priority under 35 U.S.C. §119 from Japanese Patent Application No. 2000-049031, filed Feb. 25, 2000, and Japanese Patent Application No. 11-375240, filed Dec. 28, 1999. 1. Field of the Invention The present invention relates to a reactivity control rod for a core, a core of a nuclear, a nuclear reactor and a nuclear power plant. More particularly, the present invention relates to a reactivity control rod for a core, which can elongate the lifetime of the core, a core of a nuclear reactor composed of the reactivity control rod, which can have a long lifetime, a nuclear reactor which is cooled by a liquid metal and is able to reduce scattering of the liquid metal so as to be made into a small size thereof and a nuclear power plant which comprises the nuclear reactor. 2. Description of the Related Art A conventional liquid metal cooled nuclear reactor with a small size, that is, a fast reactor is disclosed in U.S. Pat. No. 5,420,897. Moreover, a conventional fast reactor has a structure for moving a neutron reflector in a vertical direction so as to adjust (control) a leakage of neutron from the core thereof, thus to compensate a change of reactivity of the core due to a burn-up (combustion) thereof. In the aforesaid conventional liquid metal cooled nuclear reactors, an intermediate heat exchanger is arranged in a reactor vessel. A primary coolant performs the heat exchanging operation with a secondary coolant in the intermediate heat exchanger, and the exchanged secondary coolant is circulated to a steam generator arranged outside the reactor vessel so as to generate a steam. Namely, the conventional liquid metal cooled nuclear reactor has a structure of requiring a steam generator for generating a steam, an electromagnetic pump for circulating a secondary coolant between the reactor vessel and the steam generator, and piping equipments connecting them. An activated liquid metal such as sodium is used as each of the coolants. For this reason, the reactor vessel and a facility using the liquid metal arranged around the reactor vessel have complicated structures, so that there is the possibility that an auxiliary facility is required in preparation for a leakage of the activated liquid metal, fire caused thereby or the like. Moreover, in the conventional liquid metal cooled nuclear reactor, the liquid metal which is easily activated, such as sodium is used as the coolant. That is, in the steam generator, the liquid metal which is easily activated reacts to water to generate a steam. For this reason, in cases where a water leakage occurs in a heating tube of the steam generator, it is difficult to avoid an occurrence of an accident caused by the reaction between the sodium and the leaked water. The reaction between the sodium and the leaked water causes a reaction product, so that, in order to prevent the reaction product from directly being radiated, a secondary cooling system facility must be required. In addition, a facility for housing the reaction product must be required so that there is the possibility that the reactor system, as a whole, is made into a large size thereof, and that the cost of manufacturing the reactor system is made to be increased. Furthermore, the electromagnetic pump is arranged in a liquid metal; however, it is coaxially arranged in series on a downstream side (lower side) of the intermediate heat exchanger in view of a heat resistant characteristic of a large-sized conductive coil of the electromagnetic pump or the like. On the other hand, each of tube plates arranged above and below the intermediate heat exchanger has a structure which is easy to receive a thermal stress, and an enlargement of its diameter causes an increase of the thermal stress so that it is taken into consideration to prevent each of the tube plates from being made into a large size thereof. As described above, in the conventional liquid metal cooled reactor, the intermediate heat exchanger and the electromagnetic pump are vertically arranged in series; for this reason, the reactor is made into a large size thereof in its height direction (in its axial direction). The reactor with a large size in its axial direction has a structure which is easily oscillated, thereby making it unstable. On the other hand, in a conventional neutron reflector migration type of fast reactor, when elongating the lifetime of the core thereof, it must be necessary to make long the length of fuel assembly in the core. That is, according to the progress of combustion of the fuel assembly, a reactivity of the fuel assembly becomes negative. Therefore, in order to offset the negative reactivity, a neutron reflector is left up from a lower portion of the core to cover the height thereof so as to improve the ability of reflecting neutron, thereby increasing a positive reactivity of the neutron reflector, so that a reactivity of the whole core of the reactor needs to be set to 0; that is, it is necessary to make the reactor operate so as to keep a combustion in a critical state. Thus, in order to elongate an operating period of the reactor, a fuel length of the fuel assembly must be made long. Furthermore, in cases where the fuel length of the fuel assembly is made long, the reactor vessel of the reactor becomes long as a whole; as a result, there is the possibility of deteriorating the economics of the reactor. Furthermore, there are problems of causing a change of reactivity by deformation of the core in the lifetime thereof the core, an increase of pullout force of the fuel assembly. The present invention is made in view of the aforesaid problems in the related art. Accordingly, it is an object of the present invention to provide a nuclear reactor, which is capable of limiting a space for housing a liquid metal used as a coolant into an inside of a reactor vessel thereof so as to prevent scattering of the coolant to the outside thereof, whereby it is possible to make simple the whole structure of the nuclear reactor with a cooling facility, and to make compact the whole structure thereof, and to provide a nuclear power plant comprising the nuclear reactor. In order to achieve such object, according to one aspect of the present invention, there is provided a nuclear reactor in which a primary coolant is contained, including: a core composed of nuclear fuel, the coolant moving upwardly from the core by an operation thereof; an annular steam generator arranged in an upper side of the core into which the upwardly moving coolant flows and adapted to transfer heat in the coolant into water therein to generate a steam; a passage structure that defines a coolant passage for the coolant to an outside of the core, the heat-transferred coolant in the annular steam generator flowing downwardly in the coolant passage so as to flow into the core, thereby moving upwardly; and a reactor vessel arranged to surround the coolant passage so as to contain the core, the annular steam generator and the coolant passage therein. In order to achieve such object, according to another aspect of the present invention, there is provided a nuclear power plant comprising a nuclear reactor in which a coolant is contained, the nuclear reactor including: a core composed of nuclear fuel, the coolant moving upwardly from the core by an operation thereof; an annular steam generator having a plurality of heat transfer tubes and arranged in an upper side of the core into which the upwardly moving coolant flows, the annular steam generator transferring heat in the coolant with water in the heat transfer tubes to generate a steam; a passage structure that defines a coolant passage for the coolant to an outside of the core, the heat-transferred coolant in the annular steam generator flowing downwardly in the coolant passage so as to flow into the core, thereby moving upwardly; and a reactor vessel arranged to surround the coolant passage so as to contain the core, the annular steam generator and the passage means therein; a feed water branch pipe connecting to corresponding to heat transfer tubes; a steam branch pipe connecting to corresponding to heat transfer tubes, the feed water branch pipe and the steam branch pipe independently penetrating through a reactor container facility; a first feed water pipe; a steam pipe, the feed water branch pipe and the steam branch pipe being connected to the first feed water pipe and the steam pipe outside the reactor container facility, respectively; a steam bypass pipe branching from the steam branch pipe and provided with a steam separator having a bottom portion; an air conditioner provided for the steam separator via a steam facility pipe thereof; and a second feed water pipe with a feed-water pump, the bottom portion of the steam separator being connected through the second feed water pipe to the feed water branch pipe. In order to achieve such object, according to further aspect of the present invention, there is provided a reactivity control rod for use in a reactor core and for controlling a reactivity therein, comprising: a tube portion; and a mixture filled in the tube portion, the mixture being made by mixing a neutron absorber that absorbs a neutron and a neutron moderator that moderates a neutron. In order to achieve such object, according to still further aspect of the present invention, there is provided a reactor core in a core barrel of a nuclear reactor, comprising: a plurality of fuel assemblies contained in the core barrel; and a mixture contained in the core barrel, the mixture being made of a neutron absorber that absorbs a neutron in the core and a neutron moderator that moderates a neutron therein so that a reactivity of the core is controlled. According to the present invention, it is possible to reduce a heat value dispersed to the outside, thereby improving a heat efficiency thereof, and to make the reactor vessel compact into a small size as a whole, thereby securely preventing a leakage of the liquid metal. Furthermore, according to the present invention, because the whole of the reactor vessel is kept at a suitable temperature, and is protected from a rapid heat transit, it is possible to secure a structural safety of the reactor, and to make an operation of the reactor for a long period. In addition, after a shutdown of the reactor, because a natural circulating force generated by heating of the core and radiation from the reactor vessel is effectively used, it is possible to stably carry out a decay heat removal operation of the reactor. Still furthermore, in particular, the shape of the reactor is miniaturized in its longitudinal direction, and therefore, it is possible to prevent a contact of the liquid metal with the water so as to make an operation of the reactor for a long period. The preferred embodiments of the present invention will be described below with reference to the accompanying drawings. In these embodiments, as one example of a nuclear reactor according to the present invention, a liquid metal cooled reactor is described Next, with reference to FIG. 23, a fast reactor 1G according to a sixth embodiment of the present invention will be described below. FIG. 23 shows principal parts of the fast reactor 1G in this sixth embodiment, and corresponds to FIG. 19A. In FIG. 23, for simplification of explanation, like reference numerals are used to designate the same parts as FIG. 19A. The fast reactor 1G of the sixth embodiment is different from the fast reactor 1F of the above fifth embodiment in that a neutron absorber 124 with a neutron moderator is provided above the neutron reflector 4. The neutron absorber 124 with the neutron moderator includes a material produced by mixing a neutron moderator and a neutron absorber. Conventionally, the upper portion of the neutron reflector 4 is formed into a cavity in order to improve its value. In this sixth embodiment, the neutron absorber 124 with the neutron moderator is mounted into the cavity. According to the structure, in addition to the effect of the fifth embodiment, because the neutrons irradiated from the core 2A is moderated to be absorbed in the neutron absorber 124, it is possible to give a neutron shielding function to the reactor 1G, and to simplify the upper structure of the reactor 1G. FIGS. 15A, 15B and FIG. 16 show a third embodiment of the present invention. These FIGS. 15A, 15B and FIG. 16 are correspondent to FIG. 3 and FIG. 4 as described before, respectively, and show a liquid surface state and a flow of primary coolant in an operation of reactor. In FIG. 15A and FIG. 16, arrows “a” show flowing directions of the primary coolant. A liquid metal cooled nuclear reactor 1B of this third embodiment basically has the same structure as that of the above first embodiment, and therefore, overlapping explanation is omitted with reference to FIG. 1 and FIG. 2. The liquid metal cooled nuclear reactor 1B of this third embodiment differs from the above first embodiment in that the steam generator 14 is provided with an opening portion 44 of the inner shell 23 of the steam generator 14, which communicates with a cover gas space 45 of the reactor vessel 9, and is located at the upper portion from the liquid surface of the reactor vessel 9. Moreover, in this third embodiment, each of the heating tubes 16 of the steam generator 14 has a double pipe structure provided with an inner tube 16S and an outer tube 16T surrounding an outer periphery of the inner tube 16S, as shown in FIG. 15B. In addition, the reactor 1B comprises a continuous leakage monitoring unit 46 that detects a leakage in both outer and inner tubes 16T and 16S. If a large-scale water leakage occurs in a liquid metal by simultaneous breakdown of the double tubes, a water vapor or bubble of the reaction product caused by contacting the liquid metal with the water is transferred to the surroundings from the leakage portion. In this case, in the heat exchange portion, a gas transferred upwardly from the leakage portion flows to a cover gas space of the steam generator 14. On the other hand, a gas transferred downwardly from there flows through each liquid surface of the space between the intermediate shell 25 and the outer shell 24 and the space between the outer shell 24 and the reactor vessel 9 to the cover gas space of the steam generator 14. At that time, the opening portion 44 of the inner shell 23 operates so that the cover gas space 45 of the reactor vessel 9 communicates with the cover gas space of the steam generator 14. Therefore, the water vapor or bubble of the reaction product by the large-scale water leakage generated in the liquid metal is all guided to the cover gas space 45 of the reactor vessel 9 through the opening portion 44. In this third embodiment, even if a large-scale water leakage occurs in the heating tube 16 of the steam generator 14, it is possible to maintain a safety of the reactor 1B without mixing the bubble into the core 2. Incidentally, in the third embodiment, partial modification may be made. For example, as shown in FIG. 16, the lower end portion 23b of the inner shell 23 of the steam generator 14 in the reactor 1C may be arranged at a position lower than the lower end portion 24a of the outer shell 24 thereof and the lower end portion 25a of the intermediate shell 25 in the primary coolant outlet portion of the steam generator 14. According to the above construction of the modification, if a large-scale water leakage occurs, because the lower end portion 23b of the steam generator inner shell 23 in the primary coolant outlet portion of the steam generator 14 is arranged at the position lower than the lower end portion 25a of the intermediate shell 25 and the lower end portion 24a of the outer shell 24, a gas transferred downwardly of water vapor or reaction product generated by the leakage selectively flows to the upper cover gas space of the steam generator 14 via each liquid surface of the space between the intermediate shell 25 and the outer shell 24 and the space between the outer shell 24 and the reactor vessel 9. Moreover, the opening portion 44 of the steam generator 23 operates so that the cover has space 45 of the reactor vessel 9 communicates with the cover gas space of the steam generator 14, whereby the water vapor or bubble of the reaction product by the large-scale water leakage generated in the liquid metal is all guided to the cover gas space 45 of the reactor vessel 9. In this modification of the third embodiment, even if a large-scale water leakage occurs in the heating tube 16 of the steam generator 14, it is possible to maintain a safety of the reactor 1C without mixing bubble into the core 2. Furthermore, in this third embodiment, another modification with a construction may be made. More specifically, as shown in FIG. 17, the reactor 1D comprises a detecting unit 47 that detects a peculiar change in flow rate generated due to a pressure rise of the shell side of the steam generator 14 by using a change in a current of the electromagnetic pump 13. In addition, the reactor comprises an operation control unit 48 that performs a control for stopping the operation of the electromagnetic pump 13 by a detected signal outputted from the detecting unit 47. In addition, the lower end portion 23b of the steam generator inner shell 23 is arranged at a position lower than the lower end portion 24a of the outer shell 24 and the lower end portion 25a of the intermediate shell 25. According to the above construction of the reactor 1D in the another modification, the following operation is carried out. That is, if a water vapor or reaction product gas is generated in the steam generator 14 by a large-scale water leakage, the pressure rise brings about a change in a flow rate of the primary coolant in the steam generator 14. The change in the flow rate of the primary coolant in the electromagnetic pump 13 is detected by the detecting unit 47 via the outlet portion of the steam generator 14 and the coolant passage 5, and then, the electromagnetic pump 13 stopped by the control of the operation unit 47 and, after that, the electromagnetic pump 13 is again operated. In this case, a gas transferred downwardly in the steam generator 14 is transferred selectively to the upward cover gas space thereof via each liquid surface of the space between the intermediate shell 25 and the outer shell 24 and the space between the outer shell 24 and the reactor vessel 9. Because the lower end portion 23b of the steam generator inner shell 23 is arranged at the position lower than the lower end portion 25a of the intermediate shell 25 and the lower end portion 24a of the outer shell 24. Moreover, the opening portion 44 of the steam generator 23 operates so that the reactor vessel 9 communicates with the cover gas space 45 of the steam generator 14. Therefore, the water vapor or bubble of the reaction product by the large-scale water leakage generated in the liquid metal is all guided to the cover gas space 45 of the reactor vessel 9 so that, even if a large-scale water leakage occurs in the heating tube 16 of the steam generator 14, it is possible to maintain a safety of the reactor 1D without mixing a bubble into the core 2. Moreover, another construction of a further modification according to the third embodiment may be made according to the present invention. For example, the outer tube 16T is arranged at a gap to the outer periphery of the inner tube 16S so that an inert gas such as helium or the like is sealed in the gap. Furthermore, in order to detect a leakage in both inner and outer tubes 16S and 16T, a continuous leakage monitoring unit such as a helium pressure gage, a moisture content concentration monitor or the like, is provided for the reactor according to the modification. According to the above construction of the reactor in the further modification, the heating tube 16 has a double tube structure, and the continuous leakage monitoring unit detects a leakage in both inner and outer tubes 16S and 16T by the inert gas such as helium or the like sealed in the gap between the inner and outer tubes 16S and 16T so that it is possible to securely prevent a contact of the water in the tubes 16S and 16T with the liquid metal of the shell side of the steam generator 14. Accordingly, with the above construction, because of preventing the water from contacting the liquid metal, it is possible to make a stable operation of the reactor for a long period. Next, a fast reactor according to a seventh embodiment of the present invention will be described below. According to this seventh embodiment, the reactivity control assembly has the structure in that the distribution of the neutron moderator in the diametrical direction of the cladding tube 121 is gradually dense toward an inside of the cladding tube 121. The fast reactor of the seventh embodiment has almost the same effects as the fifth embodiment. Besides, according to the fast reactor of this seventh embodiment, it is possible to prevent a reduction of the initial neutron absorption effect, and to provide a linear reduction of the reactivity. Therefore, according to this seventh embodiment, the reactivity is linear, and the excess reactivity change by the burn-up is linear in appearance. Therefore, it is possible to linearly carry out the burn-up control by the operation of the neutron reflector 4, and thus, to carry out the operation of the neutron reflector 4 at an approximately constant speed, thereby readily performing the burn-up control. Next, a fast reactor according to an eighth embodiment of the present invention will be described below. According to this eighth embodiment, the mixture 122 in the cladding tube 121 of the reactivity control assembly 119 is formed so that the neutron moderator and the neutron absorber are mixed to be filed in the cladding tube 121, and, in this embodiment, as the neutron moderator, graphite is used. The eighth embodiment has almost the same effects as the fifth embodiment. Besides, because of using the graphite as the neutron moderator, it is possible to improve the safety of the fast reactor under the condition of high temperature, to increase the flexibility of designing the fast reactor and to correspond to the fast reactor wherein a coolant outlet temperature thereof is made high. Next, a fast reactor according to a ninth embodiment of the present invention will be described below. In this ninth embodiment, as shown in FIG. 20 in the fifth embodiment, the neutron absorber rod 123 is produced by mounting, as the mixture 122, the neutron moderator and the neutron absorber into the cladding tube 121 by a vibration compaction process. More specifically, in the case of mixing zirconium hydride and gadolinium as the mixture 122 of the neutron moderator and the neutron absorber, both zirconium hydride and gadolinium are weighted by a predetermined amount, and thereafter, are molded like granules. These granules are gradually put from a top opening portion of the cladding tube 121 whose bottom end is sealed, to be filled therein, while vibration is applied to the cladding tube 121 by a vibrator. After vibration filling, an upper plug is attached onto the top opening portion of the cladding tube 121 to be sealed thereto, and thus, the neutron absorber rod 123 is completed. In this case, the cladding tube 121 is attached on a vibration base of the vibrator, and then, a predetermined vibration is applied the cladding tube 121 thereby. According to this eighth embodiment, it is possible to simplify a process for forming the neutron absorber rod 123 containing the neutron moderator, and to carry out a remote control in forming of the neutron absorber rod 123. Furthermore, even in the case where the neutron moderator or the neutron absorber is a dangerous material such as a radioactive material, the neutron absorber rod 123 can be readily formed. Next, a fast reactor according to a tenth embodiment of the present invention will be described below. In this tenth embodiment, the cladding tube 121 or the wrapper tube 120 shown in FIG. 20 in the fifth embodiment is provided at its inner surface with an inside coat for preventing hydrogen from being transmitted, for example, a chromium coating layer. The chromium coating layer contacts with the mixture 122 of the neutron moderator and the neutron absorber, for example, the mixture of zirconium hydride and gadolinium. According to this tenth embodiment, the reactivity control assembly 119 is provided at its inner surface with the inside coat for preventing hydrogen from being transmitted, and then, the reactivity control assembly 119 is mounted into the center portion of the core 2A as shown in FIG. 19A and FIG. 19B. According to the structure, it is possible to prevent hydrogen generated by the burn-up in the core 2A from leaking outside the reactivity control assembly 119. Other effects are the same as the above fifth embodiment. Next, a fast reactor according to an eleventh embodiment of the present invention will be described below. In this eleventh embodiment, in order to improve a neutron absorptive power of the reactivity control assembly 119, the neutron absorber rod 123 is formed with the mixture 122 made by mixing a fission product (FP) as a neutron absorber and a zirconium hydride as a neutron moderator, and the neutron absorber rod 123 is mounted in the core 2A. According to this eleventh embodiment, the fission product (FP) is used as the neutron absorber, and thereby, it is possible to effectively use a radioactive material generated by another reactor, and thus, to contribute for a reduction of fission products. Other effects are the same as the fifth embodiment. Next, a fast reactor according to a twelfth embodiment of the present invention will be described below. In this twelfth embodiment, a mixture 122 of the neutron moderator and a thermal neutron absorber, for example, zirconium hydride and gadolinium in the fifth embodiment, is filled in the fuel assembly 116 at the vicinity of the central portion of the core, and thereby improving a neutron absorptive power. According to this twelfth embodiment, the fuel assembly 116 is provided with the mixture of the neutron moderator and a thermal neutron absorber, and thereby, there is no need of mounting the reactivity control assembly 119 in the central portion of the core. Further, this serves to readily make a design of the neutron absorber rod mounted in the center of the core or a neutron absorptive channel. Incidentally, in this embodiment, the mixture 122 is filled in the fuel assembly 116 in the vicinity of the central portion of the core. However, the present invention is not limited to the structure. That is, the neutron absorber may be filled in one of the fuel assemblies 116 in the vicinity of the central portion of the core, and the neutron moderator may be filled in another one of the fuel assembles 116 which is also in the vicinity of the central portion thereof. Next, a fast reactor according to a thirteenth embodiment of the present invention will be described below. In this thirteenth embodiment, in each of the aforesaid fast reactors, a mixture of a neutron moderator and a neutron absorber, for example, zirconium hydride and gadolinium, is provided in a burnable poison assembly at the central portion of the core, and thereby, a void reactivity of the final burn-up is transferred to a positive side. The reflector control type of fast reactor of this embodiment has the same function as the fifth embodiment. In general, in the fast reactor, with the burn-up of the core, a void reactivity rises to a positive side. This means that in the final burn-up, the positive reactivity is increased by spectral hardening in the case where void is generated. However, as this embodiment, in the case of the fast reactor, which is provided with the neutron absorber rod with the neutron moderator, in the final burn-up, an absorptive effect is reduced in a small neutron energy range. For this reason, in the final burn-up, the burn-up to fission is great in a low neutron energy range as compared with a general fast reactor. As a result, in the final burn-up, no transfer to a positive reactivity is made with respect to spectral hardening by coolant void generation. Therefore, in the final burn-up, the void reactivity is hard to be transferred to the positive side, and therefore, it is possible to improve safety of the fast reactor. Next, a fast reactor according to a fourteenth embodiment of the present invention will be described below. In this fourteenth embodiment, lead or lead-bismuth alloy is used in place of sodium used as the liquid metal coolant in the fifth embodiment. Other construction is the same as the fifth embodiment. According to this fourteenth embodiment, a fast neutron is moderated so as to be absorbed in the neutron absorber, and thereby, it is possible to improve a neutron absorptive power, and to provide a fast reactor which has a high neutron breeding ratio, thereby elongating a lifetime of the core. In this embodiment, the volume percent ratios of the neutron moderator and the neutron absorber in the neutron absorber rod 123 in the reactivity control assembly 119 mounted in the core 2A are not uniformed but different according to different positions in the axial direction of the core 2A. That is, the volume percent ratio of a predetermined portion of the mixture 122 in the neutron absorber rod 123 of the reactivity control assembly 119, which has a height in the axial direction thereof corresponding to the height H1 of the core 2A, is X1 to Y1, wherein the X1 percent is bigger than the Y1 percent, and the volume percent ratio of another predetermined portion of the mixture 122 in the neutron absorber rod 123 of the reactivity control assembly 119, which has a height in the axial direction thereof corresponding to the height H2 of the plenum is X2 to Y2, wherein the X2 percent is bigger than the Y2 percent, and the X1 percent and the Y1 percent are bigger than the X2 percent and the Y2 percent, respectively. Incidentally, in the above embodiments, the primary coolant, such as the liquid metal is circulated by means of the electromagnetic pump, but the present invention is not limited to the structure. That is, the electromagnetic pump is omitted in each reactor in each embodiment of the present invention, and the primary coolant is circulated by a natural circulating force generated by, for example, the heating of the core, the radiation from the reactor vessel and the like. In this modification, it is further possible to reduce the cost of manufacturing the reactor, and because of no use of the electromagnetic pump, it is possible to improve the safety of each reactor in the present invention. Furthermore, in the fifth embodiment to the fifteenth embodiment of the present invention, as a nuclear reactor, the liquid metal cooled type of fast reactor is applied, but the present invention is not limited to the structure. That is, in the fifth embodiment to the fifteenth embodiment, as a nuclear reactor, a light water reactor is able to be applied to the present invention, which has the described system for cooling the core, and furthermore, other nuclear reactors can be applied to the present invention. While there has been described what is at present considered to be the preferred embodiments and modifications of the present invention, it will be understood that various modifications which are not described yet may be made therein, and it is intended to cover in the appended claims all such modifications as fall within the true spirit and scope of the invention.
	summary
	summary
046631178	claims	1. A nuclear reactor fuel assembly of the kind comprising multi-pins located in a wrapper of hexagonal cross-section and stabilised by a series of spaced grids of cellular structure, characterised in that a plurality of elements extend lengthwise of the wrapper at locations between the vertices of the wrapper and are secured to the grids to form with the grids a cage structure within the wrapper, and each element includes inwardly directed projections at each grid location which extend into overlapping relation with the cellular structure of each grid without encroaching on the open cross-secion of each cell. 2. An assembly as claimed in claim 1 in which each grid includes a pin accommodating cell at each vertex of the wrapper. 3. An assembly as claimed in claim 1 in which each element is in the form of an elongate plate and said inwardly directed projections are constituted by pressed out portions of the plate. 4. An assembly as claimed in claim 1 in which the edge cells of each grid are fomed with axially projecting tabs which are secured in face-to-face relation with said inwardly directed projections of said elements. 5. An assembly as claimed in claim 3 in which the edge cells of each grid are of unitary construction. 6. An assembly as claimed in claim 1 in which at least some of said elements are provided with elastically yieldable formations for bearing against the internal faces of the wrapper and provide lateral support for the cage within the wrapper. 7. An assembly as claimed in claim 1 in which said elements are anchored to the wrapper at only one axial location, the cage being free of connection to the wrapper over the remainder of the length of said elements. 8. An assembly as claimed in claim 1 in which there are two groups of said elements each associated with a respective set of grids so as to form two cage structures at successive axial locations within the grid. 9. An assembly as claimed in claim 8 in which said cage structures are separated by a gap. 10. An assembly as claimed in claim 8 in which each cage structure is anchored to the wrapper at the end remote from the other cage structure and is free of connection to the wrapper over the intervening space between the anchored ends of the cagte structures. 11. A nuclear reactor fuel assembly of the kind comprising multi-pins located in a wrapper of hexagonal cross-section and stabilised by a series of spaced grids of cellular structure, characterised in that a plurality of elements extend lengthwise of the wrapper at locations between the vertices of the wrapper and are secured to the grids to form with the grids a cage structure within the wrapper, and the cells at the periphery of each grid are formed with axially projecting tabs and the tabs of at least some adjacent cells are of differing length and are secured together to leave the longer length partially exposed, the exposed portion being in face-to-face relation with, and secured to, an inwardly directed projection of a respective element. 12. An assembly as claimed in claim 11 in which each grid includes a pin accommodating cell at each vertex of the wrapper. 13. An assembly as claimed in claim 11 in which the edge cells of each grid are of unitary construction. 14. An assembly as claimed in claim 11 in which at least some of said elements are provided with elastically yieldable formations for bearing against the internal faces of the wrapper and provide lateral support for the cage within the wrapper. 15. An assembly as claimed in claim 11 in which said elements are anchored to the wrapper at only one axial location, the cage being free of connection to the wrapper over the remainder of the length of said elements. 16. An assembly as claimed in claim 15 in which said cage structures are separated by a gap. 17. An assembly as claimed in claim 15 in which each cage structure is anchored to the wrapper at the end remote from the other cage structure and is free of connection to the wrapper over the intervening space between the anchored ends of the cage structures.
	summary
	description	This application is a continuation-in-part of U.S. patent application Ser. No. 15/467,840 filed Mar. 23, 2017, which is a continuation-in-part of U.S. patent application Ser. No. 15/402,739 filed Jan. 10, 2017, which is a continuation-in-part of U.S. patent application Ser. No. 15/348,625 filed Nov. 10, 2016, which is a continuation-in-part of U.S. patent application Ser. No. 15/167,617 filed May 27, 2016, which is a continuation-in-part of U.S. patent application Ser. No. 15/152,479 filed May 11, 2016, which is a continuation-in-part of U.S. patent application Ser. No. 14/216,788 filed Mar. 17, 2014, which is a continuation-in-part of U.S. patent application Ser. No. 13/087,096 filed Apr. 14, 2011, which claims benefit of U.S. provisional patent application No. 61/324,776 filed Apr. 16, 2010, all of which are incorporated herein in their entirety by this reference thereto. Field of the Invention The invention relates generally to imaging and treating a tumor. Discussion of the Prior Art Cancer Treatment Proton therapy works by aiming energetic ionizing particles, such as protons accelerated with a particle accelerator, onto a target tumor. These particles damage the DNA of cells, ultimately causing their death. Cancerous cells, because of their high rate of division and their reduced ability to repair damaged DNA, are particularly vulnerable to attack on their DNA. Patents related to the current invention are summarized here. Proton Beam Therapy System F. Cole, et.al. of Loma Linda University Medical Center “Multi-Station Proton Beam Therapy System”, U.S. Pat. No. 4,870,287 (Sep. 26, 1989) describe a proton beam therapy system for selectively generating and transporting proton beams from a single proton source and accelerator to a selected treatment room of a plurality of patient treatment rooms. Imaging Lomax, A., “Method for Evaluating Radiation Model Data in Particle Beam Radiation Applications”, U.S. Pat. No. 8,461,559 B2 (Jun. 11, 2013) describes comparing a radiation target to a volume with a single pencil beam shot to the targeted volume. P. Adamee, et. al. “Charged Particle Beam Apparatus and Method for Operating the Same”, U.S. Pat. No. 7,274,018 (Sep. 25, 2007) and P. Adamee, et. al. “Charged Particle Beam Apparatus and Method for Operating the Same”, U.S. Pat. No. 7,045,781 (May 16, 2006) describe a charged particle beam apparatus configured for serial and/or parallel imaging of an object. K. Hiramoto, et.al. “Ion Beam Therapy System and its Couch Positioning System”, U.S. Pat. No. 7,193,227 (Mar. 20, 2007) describe an ion beam therapy system having an X-ray imaging system moving in conjunction with a rotating gantry. C. Maurer, et.al. “Apparatus and Method for Registration of Images to Physical Space Using a Weighted Combination of Points and Surfaces”, U.S. Pat. No. 6,560,354 (May 6, 2003) described a process of X-ray computed tomography registered to physical measurements taken on the patient's body, where different body parts are given different weights. Weights are used in an iterative registration process to determine a rigid body transformation process, where the transformation function is used to assist surgical or stereotactic procedures. M. Blair, et.al. “Proton Beam Digital Imaging System”, U.S. Pat. No. 5,825,845 (Oct. 20, 1998) describe a proton beam digital imaging system having an X-ray source that is movable into a treatment beam line that can produce an X-ray beam through a region of the body. By comparison of the relative positions of the center of the beam in the patient orientation image and the isocentre in the master prescription image with respect to selected monuments, the amount and direction of movement of the patient to make the best beam center correspond to the target isocentre is determined. S. Nishihara, et.al. “Therapeutic Apparatus”, U.S. Pat. No. 5,039,867 (Aug. 13, 1991) describe a method and apparatus for positioning a therapeutic beam in which a first distance is determined on the basis of a first image, a second distance is determined on the basis of a second image, and the patient is moved to a therapy beam irradiation position on the basis of the first and second distances. Problem There exists in the art of charged particle cancer therapy a need for accurate, precise, and rapid imaging of a patient and/or treatment of a tumor using charged particles in a complex room setting. The invention comprises an automated updated and optionally auto-implemented cancer treatment plan apparatus and method of use thereof. Elements and steps in the figures are illustrated for simplicity and clarity and have not necessarily been rendered according to any particular sequence. For example, steps that are performed concurrently or in different order are illustrated in the figures to help improve understanding of embodiments of the present invention. The invention comprises a method and apparatus for treating a tumor of a patient with positively charged particles in a treatment room, comprising the steps of: (1) providing an initial radiation treatment plan; (2) a main controller implementing the initial radiation treatment plan, as a current radiation treatment plan, using the positively charged particles delivered from a synchrotron, along a beam transport line, through a nozzle system proximate the treatment room, and into the tumor; (3) concurrent with the step of implementing, imaging the tumor, such as with protons, to generate a current image; (4) upon detection of movement of the tumor relative to surrounding constituents of the patient using the current image, the main controller, using computer implemented code, automatically generating an updated treatment plan, the updated treatment plan becoming the current radiation treatment plan; and (5) repeating the steps of implementing, imaging, and generating an updated treatment plan at least n times, where n is a positive integer of at least one. Optionally, the step of automatically generating an updated treatment plan further comprising the step of an unsupervised computer implemented algorithm using a set of computer coded inputs to automatically generate the updated treatment plan. In combination, the above described embodiment is used with an X-ray imaging and charged particle beam treatment or imaging system comprising the steps of: rotating an X-ray imaging system, configured to deliver the X-rays, around both a first rotation axis and the patient; imaging the patient using X-rays from the X-ray imaging system; and passing the positively charged particles through an exit port of a nozzle system, the nozzle system connected to a synchrotron via a first beam transport line, the positively charged particles passing into the patient from the exit port along a z-axis and at least one of: (1) treating the tumor with the positively charged particles and (2) imaging the patient with residual charged particles comprising the positively charged particles after transmitting through the patient. In one case, a first cone beam X-ray source and a second cone beam X-ray source are positioned on a first side of the patient and at least one two-dimensional X-ray detector is positioned on an opposite side of the patient from the first cone beam X-ray source. In combination, the above described embodiment is used with a multiplexed proton tomography imaging apparatus and method of use thereof. For example, a method for imaging a tumor of a patient comprises the steps of: (1) simultaneously detecting spatially resolved positively charged particle positions passing through each of a set of cross-section planes, where the cross-section planes are both prior to and posterior to the patient along a path of the positively charged particles; (2) determining a prior vector for each of the individual positively charged particles entering a patient using the detected positions; (3) determining a posterior vector for each of the individual positively charged particles exiting the patient using the detected positions; (4) generating a path, a best path, and/or a probable path of each positively charged particle through the patient; and (5) generating an image of the patient using the n probable proton paths. In one case, an imaging system: (1) delivers a set of n protons from a synchrotron: through a beam transport system exit nozzle, through a proton radial cross-section beam expander, through a first prior imaging sheet, through a second prior imaging sheet, through a patient position, through at least one posterior imaging sheet, and into a scintillation material of a beam energy scintillation detector system, where the first prior imaging sheet is positioned between the proton radial cross-section beam expander and the patient position, where the second prior imaging sheet is positioned between the proton radial cross-section beam expander and the patient position; (2) simultaneously detects spatially resolved both prior and posterior position photon emissions, resultant from passage of multiple protons; (4) determines both a prior vector and a posterior vector for each proton; and (5) determines a path for each proton through the patient and uses the determined paths, optionally and preferably with residual energy determinations, to generate an image of the patient. In combination, a method of double exposure imaging of a tumor of a patient is performed using hardware, using a detector responsive to both X-rays and positively charged particles, simultaneously, and/or in either order. The preferably near-simultaneous double exposure yields enhanced resolution due to the imaging rate versus patient movement, no requirement of a software overlay step, and associated errors, of the X-ray based image and the positively charged particle based image, and enhancement of an X-ray image, the enhancement resultant from a differing physical interaction of the positively charged particles with the patient compared to interactions of X-rays and the patient. Further, resolution enhancements utilize individual particle tracking, as measured using detection screens, to determine a probable intra-patient path. Optionally, residual energy positively charged particles, having passed through a primarily X-ray detector, are used to generate a second/dual image at a secondary detector, such as a detector based on scintillation resultant from proton absorbance. In combination, a method for imaging a tumor of a patient using X-rays and positively charged particles comprises the steps of: (1) generating an X-ray image using the X-rays directed from an X-ray source, through the patient, and to an X-ray detector, (2) generating a positively charged particle image: (a) using the positively charged particles directed from an exit nozzle, through the patient, through the X-ray detector, and to a scintillator, the scintillator emitting photons when struck by the positively charged particles and (b) generating the positively charged particle image of the tumor using a photon detector configured to detect the emitted photons, where the X-ray detector maintains a static position between said the nozzle and the scintillator during the step of generating a positively charged particle image. Individual images are optionally and preferably collected as a function of relative rotation of the patient and the imaging elements to form a three-dimensional image, such as via tomography. In combination, a method and apparatus is described for determining a position of a tumor in a patient for treatment of the tumor using positively charged particles in a treatment room. More particularly, the method and apparatus use a set of fiducial markers and fiducial detectors to mark/determine relative position of static and/or moveable objects in a treatment room using photons passing from the markers to the detectors. Further, position and orientation of at least one of the objects is calibrated to a reference line, such as a zero-offset beam treatment line passing through an exit nozzle, which yields a relative position of each fiducially marked object in the treatment room. Treatment calculations are subsequently determined using the reference line and/or points thereon. The inventor notes that the treatment calculations are optionally and preferably performed without use of an isocenter point, such as a central point about which a treatment room gantry rotates, which eliminates mechanical errors associated with the isocenter point being an isocenter volume in practice. For example, a set of fiducial marker detectors detect photons emitted from and/or reflected off of a set of fiducial markers positioned on one or more objects in a treatment room and resultant determined distances and/or calculated angles are used to determine relative positions of multiple objects or elements in the treatment room. Generally, in an iterative process, at a first time objects, such as a treatment beamline output nozzle, a specific position of a patient relative to a tumor, a scintillation detection material, an X-ray system element, and/or a detection element, are mapped and relative positions and/or angles therebetween are determined. At a second time, the position of the mapped objects is used in: (1) imaging, such as X-ray, positron emission tomography, and/or proton beam imaging and/or (2) beam targeting and treatment, such as positively charged particle based cancer treatment. As relative positions of objects in the treatment room are dynamically determined using the fiducial marking system, engineering and/or mathematical constraints of a treatment beamline isocenter is removed. In combination, a method and apparatus for imaging a tumor of a patient using positively charged particles, comprising the steps of: (1) sequentially delivering from an output nozzle, connected to a first beam transport line, to the patient: a first set of the positively charged particles comprising a first mean energy and a second set of the positively charged particles comprising a second mean energy, the second mean energy at least two mega electron Volts different from the first mean energy; (2) after transmission through the patient, sequentially detecting: a first residual energy of the first set of the positively charged particles and a second residual energy of the second set of the positively charged particles; and (3) determining a water equivalent thickness of a probed path of the patient using the first residual energy and the second residual energy. The detection step optionally uses a scintillation material and/or an X-ray detector material to detect the residual energy positively charged particles. Use of a half-maximum of a Gaussian fit to output of the detection material as a function of energy, preferably using three of more detected residual energies, yields a water equivalent thickness of the sampled beam path. In combination, an apparatus and method of use thereof are used for directing positively charged particle beams into a patient from several directions. In one example, a charged particle delivery system, comprising: a controller, an accelerator, a beam path switching magnet, a primary beam line from the accelerator to the path switching magnet, and a plurality of physically separated beam transport lines from the beam path switching magnet to a single patient treatment position is used, where the controller and beam switching magnet are used to direct sets of the positively charged particles through alternatingly selected beam transport lines to the patient, tumor, and/or an imaging detector. Optionally, during a single session and at separate times, a single repositionable treatment nozzle is repositioned to interface with each beam transport line, such as to a terminus of each beam transport line, which allows the charged particle delivery system to use one and/or fewer beam output nozzles that are moved with nozzle gantries. A single nozzle with first and second axis scanning capability along with beam transport lines leading to various sides of a patient allow the charged particle delivery system to operate without movement and/or rotation of a beam transport gantry and an associated beam transport gantry. Beam transport line gantries are optional as one or more of the beam transport lines are preferably statically positioned. In combination, a beam adjustment system is used to perform energy adjustments on circulating charged particles in a synchrotron previously accelerated to a starting energy with a traditional accelerator of the synchrotron or related devices, such as a cyclotron. The beam adjustment system uses a radio-frequency modulated potential difference applied along a longitudinal path of the circulating charged particles to accelerate or decelerate the circulating charged particles. Optionally, the beam adjustment system phase shifts the applied radio-frequency field to accelerate or decelerate the circulating charged particle while spatially longitudinally tightening a grouped bunch of the circulating charged particles. The beam adjustment system facilitates treating multiple layers or depths of the tumor between the slow step of reloading the synchrotron. Optionally, the potential differences across a gap described herein are used to accelerate or decelerate the charged particle after extraction from the synchrotron without use of the radio-frequency modulation. In combination, an imaging system, such as a positron emission tracking system, optionally used to control the beam adjustment system, is used to: dynamically determine a treatment beam position, track a history of treatment beam positions, guide the treatment beam, and/or image a tumor before, during, and/or after treatment with the charged particle beam. In combination, an imaging system translating on a linear path past a patient operates alternatingly with and/or during a gantry rotating a treatment beam around the patient. More particularly, a method for both imaging a tumor and treating the tumor of a patient using positively charged particles includes the steps of: (1) rotating a gantry support and/or gantry, connected to at least a portion of a beam transport system configured to pass a charged particle treatment beam, circumferentially about the patient and a gantry rotation axis; (2) translating a translatable imaging system past the patient on a path parallel to an axis perpendicular to the gantry rotation axis; (3) imaging the tumor using the translatable imaging system; and (4) treating the tumor using the treatment beam. In combination, a method for imaging and treating a tumor of a patient with positively charged particles, comprises the steps of: (1) using a rotatable gantry support to support and rotate a section of a positively charged particle beam transport line about a rotation axis and a tumor of a patient; (2) using a rotatable and optionally extendable secondary support to support, circumferentially position, and laterally position a primary and optional secondary imaging system about the tumor; (3) image the tumor using the primary and optional secondary imaging system as a function of rotation and/or translation of the secondary support; and (4) treat, optionally concurrently, the tumor using the positively charged particles as a function of circumferential position of the section of the charged particle beam about the tumor. In combination, a method and apparatus for imaging a tumor of a patient using positively charged particles and X-rays, comprises the steps of: (1) transporting the positively charged particles from an accelerator to a patient position using a beam transport line, where the beam transport line comprises a positively charged particle beam path and an X-ray beam path; (2) detecting scintillation induced by the positively charged particles using a scintillation detector system; (3) detecting X-rays using an X-ray detector system; (4) positioning a mounting rail through linear extension/retraction to: at a first time and at a first extension position of the mounting rail, position the scintillation detector system opposite the patient position from the exit nozzle and at a second time and at a second extension position of the mounting rail, position the X-ray detector system opposite the patient position from the exit nozzle; (5) generating an image of the tumor using output of the scintillation detector system and the X-ray detector system; and (6) alternating between the step of detecting scintillation and treating the tumor via irradiation of the tumor using the positively charged particles. In combination, a method or apparatus for tomographically imaging a sample, such as a tumor of a patient, using positively charged particles is described. Position, energy, and/or vectors of the positively charged particles are determined using a plurality of scintillators, such as layers of chemically distinct scintillators where each chemically distinct scintillator emits photons of differing wavelengths upon energy transfer from the positively charged particles. Knowledge of position of a given scintillator type and a color of the emitted photon from the scintillator type allows a determination of residual energy of the charged particle energy in a scintillator detector. Optionally, a two-dimensional detector array additionally yields x/y-plane information, coupled with the z-axis energy information, about state of the positively charged particles. State of the positively charged particles as a function of relative sample/particle beam rotation is used in tomographic reconstruction of an image of the sample or the tumor. In another example, a method or apparatus for tomographic imaging of a tumor of a patient using positively charged particles respectively positions a plurality of two-dimensional detector arrays on multiple surfaces of a scintillation material or scintillator. For instance, a first two-dimensional detector array is optically coupled to a first side or surface of a scintillation material, a second two-dimensional detector array is optically coupled to a second side of the scintillation material, and a third two-dimensional detector array is optically coupled to a third side of the scintillation material. Secondary photons emitted from the scintillation material, resultant from energy transfer from the positively charged particles, are detected by the plurality of two-dimensional detector arrays, where each detector array images the scintillation material. Combining signals from the plurality of two-dimensional detector arrays, the path, position, energy, and/or state of the positively charged particle beam as a function of time and/or rotation of the patient relative to the positively charged particle beam is determined and used in tomographic reconstruction of an image of the tumor in the patient or a sample. Particularly, a probabilistic pathway of the positively charged particles through the sample, which is altered by sample constituents, is constrained, which yields a higher resolution, a more accurate and/or a more precise image. In another example, a scintillation material is longitudinally packaged in a circumferentially surrounding sheath, where the sheath has a lower index of refraction than the scintillation material. The scintillation material yields emitted secondary photons upon passage of a charged particle beam, such as a positively charged residual particle beam having transmitted through a sample. The internally generated secondary photons within the sheath are guided to a detector element by the difference in index of refraction between the sheath and the scintillation material, similar to a light pipe or fiber optic. The coated scintillation material or fiber is referred to herein as a scintillation optic. Multiple scintillation optics are assembled to form a two-dimensional scintillation array. The scintillation array is optionally and preferably coupled to a detector or two-dimensional detector array, such as via a coupling optic, an array of focusing optics, and/or a color filter array. In combination, an ion source is coupled to the apparatus. The ion source extraction system facilitates on demand extraction of charged particles at relatively low voltage levels and from a stable ion source. For example, a triode extraction system allows extraction of charged particles, such as protons, from a maintained temperature plasma source, which reduces emittance of the extracted particles and allows use of lower, more maintainable downstream potentials to control an ion beam path of the extracted ions. The reduced emittance facilitates ion beam precision in applications, such as in imaging, tumor imaging, tomographic imaging, and/or cancer treatment. In combination, a state of a charged particle beam is monitored and/or checked, such as against a previously established radiation plan, in a position just prior to the beam entering the patient. In one example, the charged particle beam state is measured after a final manipulation of intensity, energy, shape, and/or position, such as via use of an insert, a range filter, a collimator, an aperture, and/or a compensator. In one case, one or more beam crossing elements, sheets, coatings, or layers, configured to emit photons upon passage therethrough by the charged particle beam, are positioned between the final manipulation apparatus, such as the insert, and prior to entry into the patient. In combination, a patient specific tray insert is inserted into a tray frame to form a beam control tray assembly, the beam control tray assembly is inserted into a slot of a tray receiver assembly, and the tray assembly is positioned relative to a gantry nozzle. Optionally, multiple tray inserts, each used to control a beam state parameter, are inserted into slots of the tray receiver assembly. The beam control tray assembling includes an identifier, such as an electromechanical identifier, of the particular insert type, which is communicated to a main controller, such as via the tray receiver assembly. Optionally and preferably, a hand control pendant is used in loading and/or positioning the tray receiver assembly. In combination, a gantry positions both: (1) a section of a beam transport system, such as a terminal section, used to transport and direct positively charged particles to a tumor and (2) at least one imaging system. In one case, the imaging system is orientated on a same axis as the positively charged particle, such as at a different time through rotation of the gantry. In another case, the imaging system uses at least two crossing beamlines, each beamline coupled to a respective detector, to yield multiple views of the patient. In another case, one or more imaging subsystem yields a two-dimensional image of the patient, such as for position confirmation and/or as part of a set of images used to develop a three-dimensional image of the patient. In combination, multiple linked control stations are used to control position of elements of a beam transport system, nozzle, and/or patient specific beam shaping element relative to a dynamically controlled patient position and/or an imaging surface, element, or system. In combination, a tomography system is optionally used in combination with a charged particle cancer therapy system. The tomography system uses tomography or tomographic imaging, which refers to imaging by sections or sectioning through the use of a penetrating wave, such as a positively charge particle from an injector and/or accelerator. Optionally and preferably, a common injector, accelerator, and beam transport system is used for both charged particle based tomographic imaging and charged particle cancer therapy. In one case, an output nozzle of the beam transport system is positioned with a gantry system while the gantry system and/or a patient support maintains a scintillation plate of the tomography system on the opposite side of the patient from the output nozzle. In another example, a charged particle state determination system, of a cancer therapy system or tomographic imaging system, uses one or more coated layers in conjunction with a scintillation material, scintillation detector and/or a tomographic imaging system at time of tumor and surrounding tissue sample mapping and/or at time of tumor treatment, such as to determine an input vector of the charged particle beam into a patient and/or an output vector of the charged particle beam from the patient. In another example, the charged particle tomography apparatus is used in combination with a charged particle cancer therapy system. For example, tomographic imaging of a cancerous tumor is performed using charged particles generated with an injector, accelerator, and guided with a delivery system. The cancer therapy system uses the same injector, accelerator, and guided delivery system in delivering charged particles to the cancerous tumor. For example, the tomography apparatus and cancer therapy system use a common raster beam method and apparatus for treatment of solid cancers. More particularly, the invention comprises a multi-axis and/or multi-field raster beam charged particle accelerator used in: (1) tomography and (2) cancer therapy. Optionally, the system independently controls patient translation position, patient rotation position, two-dimensional beam trajectory, delivered radiation beam energy, delivered radiation beam intensity, beam velocity, timing of charged particle delivery, and/or distribution of radiation striking healthy tissue. The system operates in conjunction with a negative ion beam source, synchrotron, patient positioning, imaging, and/or targeting method and apparatus to deliver an effective and uniform dose of radiation to a tumor while distributing radiation striking healthy tissue. In combination, a treatment delivery control system (TDCS) or main controller is used to control multiple aspects of the cancer therapy system, including one or more of: an imaging system, such as a CT or PET; a positioner, such as a couch or patient interface module; an injector or injection system; a radio-frequency quadrupole system; a ring accelerator or synchrotron; an extraction system; an irradiation plan; and a display system. The TDCS is preferably a control system for automated cancer therapy once the patient is positioned. The TDCS integrates output of one or more of the below described cancer therapy system elements with inputs of one or more of the below described cancer therapy system elements. More generally, the TDCS controls or manages input and/or output of imaging, an irradiation plan, and charged particle delivery. In combination, one or more trays are inserted into the positively charged particle beam path, such as at or near the exit port of a gantry nozzle in close proximity to the patient. Each tray holds an insert, such as a patient specific insert for controlling the energy, focus depth, and/or shape of the charged particle beam. Examples of inserts include a range shifter, a compensator, an aperture, a ridge filter, and a blank. Optionally and preferably, each tray communicates a held and positioned insert to a main controller of the charged particle cancer therapy system. The trays optionally hold one or more of the imaging sheets configured to emit light upon transmission of the charged particle beam through a corresponding localized position of the one or more imaging sheets. For clarity of presentation and without loss of generality, throughout this document, treatment systems and imaging systems are described relative to a tumor of a patient. However, more generally any sample is imaged with any of the imaging systems described herein and/or any element of the sample is treated with the positively charged particle beam(s) described herein. Charged Particle Beam Therapy Throughout this document, a charged particle beam therapy system, such as a proton beam, hydrogen ion beam, or carbon ion beam, is described. Herein, the charged particle beam therapy system is described using a proton beam. However, the aspects taught and described in terms of a proton beam are not intended to be limiting to that of a proton beam and are illustrative of a charged particle beam system, a positively charged beam system, and/or a multiply charged particle beam system, such as C4+ or C6+. Any of the techniques described herein are equally applicable to any charged particle beam system. Referring now to FIG. 1A, a charged particle beam system 100 is illustrated. The charged particle beam preferably comprises a number of subsystems including any of: a main controller 110; an injection system 120; a synchrotron 130 that typically includes: (1) an accelerator system 131 and (2) an internal or connected extraction system 134; a beam transport system 135; a scanning/targeting/delivery system 140; a nozzle system 146; a patient interface module 150; a display system 160; and/or an imaging system 170. An exemplary method of use of the charged particle beam system 100 is provided. The main controller 110 controls one or more of the subsystems to accurately and precisely deliver protons to a tumor of a patient. For example, the main controller 110 obtains an image, such as a portion of a body and/or of a tumor, from the imaging system 170. The main controller 110 also obtains position and/or timing information from the patient interface module 150. The main controller 110 optionally controls the injection system 120 to inject a proton into a synchrotron 130. The synchrotron typically contains at least an accelerator system 131 and an extraction system 134. The main controller 110 preferably controls the proton beam within the accelerator system, such as by controlling speed, trajectory, and timing of the proton beam. The main controller then controls extraction of a proton beam from the accelerator through the extraction system 134. For example, the controller controls timing, energy, and/or intensity of the extracted beam. The controller 110 also preferably controls targeting of the proton beam through the scanning/targeting/delivery system 140 to the patient interface module 150. One or more components of the patient interface module 150, such as translational and rotational position of the patient, are preferably controlled by the main controller 110. Further, display elements of the display system 160 are preferably controlled via the main controller 110. Displays, such as display screens, are typically provided to one or more operators and/or to one or more patients. In one embodiment, the main controller 110 times the delivery of the proton beam from all systems, such that protons are delivered in an optimal therapeutic manner to the tumor of the patient. Herein, the main controller 110 refers to a single system controlling the charged particle beam system 100, to a single controller controlling a plurality of subsystems controlling the charged particle beam system 100, or to a plurality of individual controllers controlling one or more sub-systems of the charged particle beam system 100. Still referring to FIG. 43, a first input to the semi-automated radiation treatment plan development system 4300, used to generate the radiation treatment plan 4310, is a requirement of dose distribution 4320. Herein, dose distribution comprises one or more parameters, such as a prescribed dosage 4321 to be delivered; an evenness or uniformity of radiation dosage distribution 4322; a goal of reduced overall dosage 4323 delivered to the patient 730; a specification related to minimization or reduction of dosage delivered to critical voxels 4324 of the patient 730, such as to a portion of an eye, brain, nervous system, and/or heart of the patient 730; and/or an extent of, outside a perimeter of the tumor, dosage distribution 4325. The automated radiation treatment plan development system 4300 calculates and/or iterates a best radiation treatment plan using the inputs, such as via a computer implemented algorithm. Each parameter provided to the automated radiation treatment plan development system 4300, optionally and preferably contains a weight or importance. For clarity of presentation and without loss of generality, two cases illustrate. In a first case, a requirement/goal of reduction of dosage or even complete elimination of radiation dosage to the optic nerve of the eye, provided in the minimized dosage to critical voxels 4324 input is given a higher weight than a requirement/goal to minimize dosage to an outer area of the eye, such as the rectus muscle, or an inner volume of the eye, such as the vitreous humor of the eye. This first case is exemplary of one input providing more than one sub-input where each sub-input optionally includes different weighting functions. In a second case, a first weight and/or first sub-weight of a first input is compared with a second weight and/or a second sub-weight of a second input. For instance, a distribution function, probability, or precision of the even radiation dosage distribution 4322 input optionally comprises a lower associated weight than a weight provided for the reduce overall dosage 4323 input to prevent the computer algorithm from increasing radiation dosage in an attempt to yield an entirely uniform dose distribution. Each parameter and/or sub-parameter provided to the automated radiation treatment plan development system 4300, optionally and preferably contains a limit, such as a hard limit, an upper limit, a lower limit, a probability limit, and/or a distribution limit. The limit requirement is optionally used, by the computer algorithm generating the radiation treatment plan 4310, with or without the weighting parameters, described supra. Still referring to FIG. 43, a second input to the semi-automated radiation treatment plan development system 4300, is a patient motion 4330 input. The patient motion 4330 input comprises: a move the patient in one direction 4332 input, a move the patient at a uniform speed 4333 input, a total patient rotation 4334 input, a patient rotation rate 4335 input, and/or a patient tilt 4336 input. For clarity of presentation and without loss of generality, the patient motion inputs are further described, supra, in several cases. Still referring to FIG. 43, in a first case the automated radiation treatment plan development system 4300, provides a guidance input, such as the move the patient in one direction 4332 input, but a further associated directive is if other goals require it or if a better overall score of the radiation treatment plan 4310 is achieved, the guidance input is optionally automatically relaxed. Similarly, the move the patient at a uniform rate 4333 input is also provided with a guidance input, such as a low associated weight that is further relaxable to yield a high score, of the radiation treatment plan 4310, but is only relaxed or implemented an associated fixed or hard limit number of times. Still referring to FIG. 43, in a second case the computer implemented algorithm, in the automated radiation treatment plan development system 4300, optionally generates a sub-score. For instance, a patient comfort score optionally comprises a score combining a metric related to two or more of: the move the patient in one direction 4332 input, the move the patient at a uniform rate 4333 input, the total patient rotation 4334 input, the patient rotation rate 4335 input, and/or the reduce patient tilt 4336 input. The sub-score, which optionally has a preset limit, allows flexibility, in the computer implemented algorithm, to yield on patient movement parameters as a whole, again to result in patient comfort. Still referring to FIG. 43, in a third case the automated radiation treatment plan development system 4300 optionally contains an input used for more than one sub-function. For example, a reduce treatment time 4331 input is optionally used as a patient comfort parameter and also links into the dose distribution 4320 input. Still referring to FIG. 43, a third input to the automated radiation treatment plan development system 4300 comprises output of an imaging system, such as any of the imaging systems described herein. Still referring to FIG. 43, a fourth optional input to the automated radiation treatment plan development system 4300 is structural and/or physical elements present in the treatment room 1222. Again, for clarity of presentation and without loss of generality, two cases illustrate treatment room object information as an input to the automated development of the radiation treatment plan 4310. Still referring to FIG. 43, in a first case the automated radiation treatment plan development system 4300 is optionally provided with a pre-scan of potentially intervening support structures 4422 input, such as a patient support device, a patient couch, and/or a patient support element, where the pre-scan is an image/density/redirection impact of the support structure on the positively charged particle treatment beam. Preferably, the pre-scan is an actual image or tomogram of the support structure using the actual facility synchrotron, a remotely generated actual image, and/or a calculated impact of the intervening structure on the positively charge particle beam. Determination of impact of the support structure on the charged particle beam is further described, infra. Still referring to FIG. 43, in a second case the automated radiation treatment plan development system 4300 is optionally provided with a reduce treatment through a support structure 4344 input. As described supra, an associated weight, guidance, and/or limit is optionally provided with the reduce treatment through the support structure 4344 input and, also as described supra, the support structure input is optionally compromised relative to a more critical parameter, such as the deliver prescribed dosage 4321 input or the minimize dosage to critical voxels 4324 of the patient 730 input. Still referring to FIG. 43, a fifth optional input to the automated radiation treatment plan development system 4300 is a doctor input 4236, such as provided only prior to the auto generation of the radiation treatment plan. Separately, doctor oversight 4230 is optionally provided to the automated radiation treatment plan development system 4300 as plans are being developed, such as an intervention to restrict an action, an intervention to force an action, and/or an intervention to change one of the inputs to the automated radiation treatment plan development system 4300 for a radiation plan for a particular individual. Still referring to FIG. 43, a sixth input to the automated radiation treatment plan development system 4300 comprises information related to collapse and/or shifting of the tumor 720 of the patient 730 during treatment. For instance, the radiation treatment plan 4310 is automatically updated, using the automated radiation treatment plan development system 4300, during treatment using an input of images of the tumor 720 of the patient 730 collected concurrently with treatment using the positively charged particles. For instance, as the tumor 720 reduces in size with treatment, the tumor 720 collapses inward and/or shifts. The auto-updated radiation treatment plan is optionally auto-implemented, such as without the patient moving from a treatment position. Optionally, the automated radiation treatment plan development system 4300 tracks dosage of untreated voxels of the tumor 720 and/or tracks partially irradiated, relative to the prescribed dosage 4321, voxels and dynamically and/or automatically adjusts the radiation treatment plan 4310 to provide the full prescribed dosage to each voxel despite movement of the tumor 720. Similarly, the automated radiation treatment plan development system 4300 tracks dosage of treated voxels of the tumor 720 and adjusts the automatically updated tumor treatment plan to reduce and/or minimize further radiation delivery to the fully treated and shifted tumor voxels while continuing treatment of the partially treated and/or untreated shifted voxels of the tumor 720. Intervening Object As the positively charged particle beam travels along a treatment beam path in the treatment room 1222, in some situations the positively charged particle beam passes through an object, referred to herein as an intervening object, which decelerates and/or redirects the positively charged particles. Herein, predetermining an impact of the intervening object on the positively charged particle beam is described and compensating for the impact is described. Referring now to FIG. 44, a method for determining an impact of an object 4400 on the positively charged particle beam is described. Herein, an intervening object 4410 is any inanimate and/or non-biological object in the treatment room 1222 between an exit surface of the nozzle system 146 and a terminal point of the charged particle beam in the tumor as determined by the Bragg peak. Examples of intervening objects 4410 comprise: a patient couch, a patient support element, an implant, an embedded element in the patient 730, and/or a prosthesis. Parameters defining the intervening object 4410 and/or the physical intervening object 4410 itself is provided to the method for determining an impact of an object 4400. Still referring to FIG. 44, in a first case, the intervening object 4410 is pre-scanned 4420, such as with an X-ray system, a positron emission system, and/or a positively charged particle beam system. For example, a three-dimensional (3D) computed tomography (CT) proton beam image of the intervening object is obtained. In the radiation treatment plan 4310, described supra, a determination is made for each treatment beam, of a set of treatment beam covering relative motion and/or translation of the nozzle system and the patient, whether or not the charged particle beam will traverse the intervening object 4410 and if so, what cross-section of the intervening object 4410 is traversed at each position along a pathway through the intervening object 4410. For each voxel of the intervening object 4410 along the treatment path, a deceleration and/or redirection/scattering of the treatment beam is calculated. By integrating the impact of the intervening object 4410 across the voxels traversed, a total deceleration and/or net direction/scattering change of the positively charged particle beam is predetermined. Subsequently, in a generation of the radiation treatment plan step 4440 or in the auto-generate the radiation treatment plan step 4226, the incident energy of the positively charged particles for each incident treatment vector of the radiation treatment plan 4310 is adjusted to increase the energy of the initial charged particle beam to compensate for the loss of energy or deceleration of the positively charged particle beam resultant from passage through the intervening object. Similarly, in the generation of the radiation treatment plan step 4440 or in the auto-generate the radiation treatment plan step 4226, the incident vector/direction of the positively charged particles for each incident treatment vector of the radiation treatment plan 4310 is adjusted to compensate for redirection of the initial charged particle beam to account for redirection of the treatment beam resultant from passage through the intervening object. Still referring to FIG. 44 and still referring to the first case of pre-scanning the object 4420, two approaches are used to measure the impact of the intervening object 4410 on the positively charged particle beam. In a first approach, the initial energy and direction of a treatment beam mimic traverses an actual treatment path 4424 through the intervening object 4410 and a residual energy and/or altered direction of the treatment beam mimic is measured, such as with the tomography apparatus and/or tomography imaging system described supra. In this first approach, the energy and/or vector of a particular incident treatment beam is adjusted to compensate for a directly measured impact of the intervening object 4410 on the particular incident treatment beam to yield a planned treatment beam in the radiation treatment plan. In a second approach, the 3D CT image of the intervening object 4410 is used to calculate impact to a transformed and/or proposed incident treatment path 4424 through the intervening object 4410, where the proposed incident treatment path is a combination of voxels crossing many layers of the 3D CT image of the intervening object. Similar to the first approach, in the second approach, a residual energy and/or altered direction of the proposed treatment path is adjusted to compensate for the calculated impact, using real image data, of the intervening object 4410 on the proposed incident treatment beam to yield a planned treatment beam in the radiation treatment plan. The first case finds particular utility for standard items, such as a standard implanted item, or for an item readily available in the treatment room, such as a patient support/positioning/movement system element. Still referring to FIG. 44, in a second case, impact of the intervening object 4410 on the positively charged particle treatment beam is pre-calculated 4430 using known physical properties. For example, physical parameters such as material type, material density, and shape of the intervening object 4410 are coded into a 3D model of the intervening object 4410. Similar to the first case, the 3D model of the intervening object 4410 is used to determine a deceleration and/or altered direction of a proposed treatment path and the model data is used to adjust a proposed treatment beam to a planned treatment beam that accounts for the purely calculated impact of the intervening object 4410 on the treatment beam. One method of pre-calculating impact of the intervening object 4410 on a treatment beam is via use of finite element analysis 4432. The second case finds particular utility for compensating for an implanted object, such as a hip replacement, titanium bone support, plate, fastener, or other medically implanted item, especially a custom implant. Still referring to FIG. 44, in a third case, an actual image, such as a 3D CT image, of the intervening object 4410 is combined with model based calculations of impact of the intervening object 4410 on an incident particle beam, such as through use of known physical material properties, chemical properties, physical shape, and/or chemical/physical state of the intervening object. The resulting hybrid measured-calculated impact of the intervening object 4410 on a proposed treatment beam is used to generate an actual treatment beam vector in the radiation treatment plan 4310, which is generated 4440 and/or auto-generated 4226. Automated Adaptive Treatment Referring now to FIG. 45, a system for automatically updating the radiation treatment plan 4500 and preferably automatically updating and implementing the radiation treatment plan is illustrated. In a first task 4510, an initial radiation treatment plan is provided, such as the auto-generated radiation treatment plan 4226, described supra. The first task is a startup task of an iterative loop of tasks and/or recurring set of tasks, described herein as comprising tasks two to four. In a second task 4520, the tumor 720 is treated using the positively charged particles delivered from the synchrotron 130. In a third task 4530, changes in the tumor shape and/or changes in the tumor position relative to surrounding constituents of the patient 730 are observed, such as via any of the imaging systems described herein. The imaging optionally occurs simultaneously, concurrently, periodically, and/or intermittently with the second task while the patient remains positioned by the patient positioning system. The main controller 110 uses images from the imaging system(s) and the provided and/or current radiation treatment plan to determine if the treatment plan is to be followed or modified. Upon detected relative movement of the tumor 720 relative to the other elements of the patient 730 and/or change in a shape of the tumor 730, a fourth task 4540 of updating the treatment plan is optionally and preferably automatically implemented and/or use of the radiation treatment plan development system 4300, described supra, is implemented. The process of tasks two to four is optionally and preferably repeated n times where n is a positive integer of greater than 1, 2, 5, 10, 20, 50, or 100 and/or until a treatment session of the tumor 720 ends and the patient 730 departs the treatment room 1222. Automated Treatment Referring now to FIG. 46, an automated cancer therapy treatment system 4600 is illustrated. In the automated cancer therapy treatment system 4600, a majority of tasks are implemented according to a computer based algorithm and/or an intelligent system. Optionally and preferably, a medical professional oversees the automated cancer therapy treatment system 4600 and stops or alters the treatment upon detection of an error but fundamentally observes the process of computer algorithm guided implementation of the system using electromechanical elements, such as any of the hardware and/or software described herein. Optionally and preferably, each sub-system and/or sub-task is automated. Optionally, one or more of the sub-systems and/or sub-tasks are performed by a medical professional. For instance, the patient 730 is optionally initially positioned in the patient positioning system by the medical professional and/or a tray insert 510 is loaded into a tray assembly 400 by the medical professional. Optional and preferably automated, such as computer algorithm implemented, sub-tasks include one or more and preferably all of: receiving the treatment plan input 4300, such as a prescription, guidelines, patient motion guidelines 4330, dose distribution guidelines 4320, intervening object 4310 information, and/or images of the tumor 720; using the treatment plan input 4300 to auto-generate a radiation treatment plan 4226; auto-positioning 4222 the patient 730; auto-imaging 4224 the tumor 720; implementing medical profession oversight 4238 instructions; auto-implementing the radiation treatment plan 4520/delivering the positively charged particles to the tumor 720; auto-reposition the patient 4521 for subsequent radiation delivery; auto-rotate a nozzle position 4522 of the nozzle system 146 relative to the patient 730; auto-translate a nozzle position 4523 of the nozzle system 146 relative to the patient 730; auto-verify a clear treatment path using an imaging system, such as to observe presence of a metal object or unforeseen dense object via an X-ray image; auto-verify a clear treatment path using fiducial indicators 4524; auto control a state of the positively charge particle beam 4525, such as energy, intensity, position (x,y,z), duration, and/or direction; auto-control a particle beam path 4526, such as to a selected beamline and/or to a selected nozzle; auto implement positioning a tray insert 510 and/or tray assembly 400; auto-update a tumor image 4610; auto-observe tumor movement 4530; and/or generate an auto-modified radiation treatment plan 4540/new treatment plan. Still yet another embodiment includes any combination and/or permutation of any of the elements described herein. The main controller, a localized communication apparatus, and/or a system for communication of information optionally comprises one or more subsystems stored on a client. The client is a computing platform configured to act as a client device or other computing device, such as a computer, personal computer, a digital media device, and/or a personal digital assistant. The client comprises a processor that is optionally coupled to one or more internal or external input device, such as a mouse, a keyboard, a display device, a voice recognition system, a motion recognition system, or the like. The processor is also communicatively coupled to an output device, such as a display screen or data link to display or send data and/or processed information, respectively. In one embodiment, the communication apparatus is the processor. In another embodiment, the communication apparatus is a set of instructions stored in memory that is carried out by the processor. The client includes a computer-readable storage medium, such as memory. The memory includes, but is not limited to, an electronic, optical, magnetic, or another storage or transmission data storage medium capable of coupling to a processor, such as a processor in communication with a touch-sensitive input device linked to computer-readable instructions. Other examples of suitable media include, for example, a flash drive, a CD-ROM, read only memory (ROM), random access memory (RAM), an application-specific integrated circuit (ASIC), a DVD, magnetic disk, an optical disk, and/or a memory chip. The processor executes a set of computer-executable program code instructions stored in the memory. The instructions may comprise code from any computer-programming language, including, for example, C originally of Bell Laboratories, C++, C#, Visual Basic® (Microsoft, Redmond, Wash.), Matlab® (MathWorks, Natick, Mass.), Java® (Oracle Corporation, Redwood City, Calif.), and JavaScript® (Oracle Corporation, Redwood City, Calif.). Herein, any number, such as 1, 2, 3, 4, 5, is optionally more than the number, less than the number, or within 1, 2, 5, 10, 20, or 50 percent of the number. Herein, an element and/or object is optionally manually and/or mechanically moved, such as along a guiding element, with a motor, and/or under control of the main controller. The particular implementations shown and described are illustrative of the invention and its best mode and are not intended to otherwise limit the scope of the present invention in any way. Indeed, for the sake of brevity, conventional manufacturing, connection, preparation, and other functional aspects of the system may not be described in detail. Furthermore, the connecting lines shown in the various figures are intended to represent exemplary functional relationships and/or physical couplings between the various elements. Many alternative or additional functional relationships or physical connections may be present in a practical system. In the foregoing description, the invention has been described with reference to specific exemplary embodiments; however, it will be appreciated that various modifications and changes may be made without departing from the scope of the present invention as set forth herein. The description and figures are to be regarded in an illustrative manner, rather than a restrictive one and all such modifications are intended to be included within the scope of the present invention. Accordingly, the scope of the invention should be determined by the generic embodiments described herein and their legal equivalents rather than by merely the specific examples described above. For example, the steps recited in any method or process embodiment may be executed in any order and are not limited to the explicit order presented in the specific examples. Additionally, the components and/or elements recited in any apparatus embodiment may be assembled or otherwise operationally configured in a variety of permutations to produce substantially the same result as the present invention and are accordingly not limited to the specific configuration recited in the specific examples. Benefits, other advantages and solutions to problems have been described above with regard to particular embodiments; however, any benefit, advantage, solution to problems or any element that may cause any particular benefit, advantage or solution to occur or to become more pronounced are not to be construed as critical, required or essential features or components. As used herein, the terms “comprises”, “comprising”, or any variation thereof, are intended to reference a non-exclusive inclusion, such that a process, method, article, composition or apparatus that comprises a list of elements does not include only those elements recited, but may also include other elements not expressly listed or inherent to such process, method, article, composition or apparatus. Other combinations and/or modifications of the above-described structures, arrangements, applications, proportions, elements, materials or components used in the practice of the present invention, in addition to those not specifically recited, may be varied or otherwise particularly adapted to specific environments, manufacturing specifications, design parameters or other operating requirements without departing from the general principles of the same. Although the invention has been described herein with reference to certain preferred embodiments, one skilled in the art will readily appreciate that other applications may be substituted for those set forth herein without departing from the spirit and scope of the present invention. Accordingly, the invention should only be limited by the Claims included below. Still referring to FIG. 2C and FIG. 2D, optionally and preferably geometries of the gating electrode 204 and/or the extraction electrode 206 are used to focus the extracted ions along the initial ion beam path 262. Still referring to FIG. 2C and FIG. 2D, the lower emittance of the electron cyclotron resonance triode extraction system is optionally and preferably coupled with a downbeam or downstream radio-frequency quadrupole, used to focus the beam, and/or a synchrotron, used to accelerate the beam. Still referring to FIG. 2C and FIG. 2D, the lower emittance of the electron cyclotron resonance triode extraction system is maintained through the synchrotron 130 and to the tumor of the patient resulting in a more accurate, precise, smaller, and/or tighter treatment voxel of the charged particle beam or charged particle pulse striking the tumor. Still referring to FIG. 2C and FIG. 2D, the lower emittance of the electron cyclotron resonance triode extraction system reduces total beam spread through the synchrotron 130 and the tumor to one or more imaging elements, such as an optical imaging sheet or scintillation material emitting photons upon passage of the charged particle beam or striking of the charged particle beam, respectively. The lower emittance of the charged particle beam, optionally and preferably maintained through the accelerator system 134 and beam transport system yields a tighter, more accurate, more precise, and/or smaller particle beam or particle burst diameter at the imaging surfaces and/or imaging elements, which facilitates more accurate and precise tumor imaging, such as for subsequent tumor treatment or to adjust, while the patient waits in a treatment position, the charged particle treatment beam position. Any feature or features of any of the above provided examples are optionally and preferably combined with any feature described in other examples provided, supra, or herein. Ion Extraction from Accelerator Referring now to FIG. 3, both: (1) an exemplary proton beam extraction system 300 from the synchrotron 130 and (2) a charged particle beam intensity control system 305 are illustrated. For clarity, FIG. 3 removes elements represented in FIG. 1C, such as the turning magnets, which allows for greater clarity of presentation of the proton beam path as a function of time. Generally, protons are extracted from the synchrotron 130 by slowing the protons. As described, supra, the protons were initially accelerated in a circulating path, which is maintained with a plurality of main bending magnets 132. The circulating path is referred to herein as an original central beamline 264. The protons repeatedly cycle around a central point in the synchrotron 136. The proton path traverses through a radio frequency (RF) cavity system 310. To initiate extraction, an RF field is applied across a first blade 312 and a second blade 314, in the RF cavity system 310. The first blade 312 and second blade 314 are referred to herein as a first pair of blades. In the proton extraction process, an RF voltage is applied across the first pair of blades, where the first blade 312 of the first pair of blades is on one side of the circulating proton beam path 264 and the second blade 314 of the first pair of blades is on an opposite side of the circulating proton beam path 264. The applied RF field applies energy to the circulating charged-particle beam. The applied RF field alters the orbiting or circulating beam path slightly of the protons from the original central beamline 264 to an altered circulating beam path 265. Upon a second pass of the protons through the RF cavity system, the RF field further moves the protons off of the original proton beamline 264. For example, if the original beamline is considered as a circular path, then the altered beamline is slightly elliptical. The frequency of the applied RF field is timed to apply outward or inward movement to a given band of protons circulating in the synchrotron accelerator. Orbits of the protons are slightly more off axis compared to the original circulating beam path 264. Successive passes of the protons through the RF cavity system are forced further and further from the original central beamline 264 by altering the direction and/or intensity of the RF field with each successive pass of the proton beam through the RF field. Timing of application of the RF field and/or frequency of the RF field is related to the circulating charged particles circulation pathlength in the synchrotron 130 and the velocity of the charged particles so that the applied RF field has a period, with a peak-to-peak time period, equal to a period of time of beam circulation in the synchrotron 130 about the center 136 or an integer multiple of the time period of beam circulation about the center 136 of the synchrotron 130. Alternatively, the time period of beam circulation about the center 136 of the synchrotron 130 is an integer multiple of the RF period time. The RF period is optionally used to calculated the velocity of the charged particles, which relates directly to the energy of the circulating charged particles. The RF voltage is frequency modulated at a frequency about equal to the period of one proton cycling around the synchrotron for one revolution or at a frequency than is an integral multiplier of the period of one proton cycling about the synchrotron. The applied RF frequency modulated voltage excites a betatron oscillation. For example, the oscillation is a sine wave motion of the protons. The process of timing the RF field to a given proton beam within the RF cavity system is repeated thousands of times with each successive pass of the protons being moved approximately one micrometer further off of the original central beamline 264. For clarity, the approximately 1000 changing beam paths with each successive path of a given band of protons through the RF field are illustrated as the altered beam path 265. The RF time period is process is known, thus energy of the charged particles at time of hitting the extraction material or material 330, described infra, is known. With a sufficient sine wave betatron amplitude, the altered circulating beam path 265 touches and/or traverses a material 330, such as a foil or a sheet of foil. The foil is preferably a lightweight material, such as beryllium, a lithium hydride, a carbon sheet, or a material having low nuclear charge components. Herein, a material of low nuclear charge is a material composed of atoms consisting essentially of atoms having six or fewer protons. The foil is preferably about 10 to 150 microns thick, is more preferably about 30 to 100 microns thick, and is still more preferably about 40 to 60 microns thick. In one example, the foil is beryllium with a thickness of about 50 microns. When the protons traverse through the foil, energy of the protons is lost and the speed of the protons is reduced. Typically, a current is also generated, described infra. Protons moving at the slower speed travel in the synchrotron with a reduced radius of curvature 266 compared to either the original central beamline 264 or the altered circulating path 265. The reduced radius of curvature 266 path is also referred to herein as a path having a smaller diameter of trajectory or a path having protons with reduced energy. The reduced radius of curvature 266 is typically about two millimeters less than a radius of curvature of the last pass of the protons along the altered proton beam path 265. The thickness of the material 330 is optionally adjusted to create a change in the radius of curvature, such as about ½, 1, 2, 3, or 4 mm less than the last pass of the protons 265 or original radius of curvature 264. The reduction in velocity of the charged particles transmitting through the material 330 is calculable, such as by using the pathlength of the betatron oscillating charged particle beam through the material 330 and/or using the density of the material 330. Protons moving with the smaller radius of curvature travel between a second pair of blades. In one case, the second pair of blades is physically distinct and/or is separated from the first pair of blades. In a second case, one of the first pair of blades is also a member of the second pair of blades. For example, the second pair of blades is the second blade 314 and a third blade 316 in the RF cavity system 310. A high voltage DC signal, such as about 1 to 5 kV, is then applied across the second pair of blades, which directs the protons out of the synchrotron through an extraction magnet 137, such as a Lambertson extraction magnet, into a transport path 268. Control of acceleration of the charged particle beam path in the synchrotron with the accelerator and/or applied fields of the turning magnets in combination with the above described extraction system allows for control of the intensity of the extracted proton beam, where intensity is a proton flux per unit time or the number of protons extracted as a function of time. For example, when a current is measured beyond a threshold, the RF field modulation in the RF cavity system is terminated or reinitiated to establish a subsequent cycle of proton beam extraction. This process is repeated to yield many cycles of proton beam extraction from the synchrotron accelerator. In another embodiment, instead of moving the charged particles to the material 330, the material 330 is mechanically moved to the circulating charged particles. Particularly, the material 330 is mechanically or electromechanically translated into the path of the circulating charged particles to induce the extraction process, described supra. In this case, the velocity or energy of the circulating charged particle beam is calculable using the pathlength of the beam path about the center 136 of the synchrotron 130 and from the force applied by the bending magnets 132. In either case, because the extraction system does not depend on any change in magnetic field properties, it allows the synchrotron to continue to operate in acceleration or deceleration mode during the extraction process. Stated differently, the extraction process does not interfere with synchrotron acceleration. In stark contrast, traditional extraction systems introduce a new magnetic field, such as via a hexapole, during the extraction process. More particularly, traditional synchrotrons have a magnet, such as a hexapole magnet, that is off during an acceleration stage. During the extraction phase, the hexapole magnetic field is introduced to the circulating path of the synchrotron. The introduction of the magnetic field necessitates two distinct modes, an acceleration mode and an extraction mode, which are mutually exclusive in time. The herein described system allows for acceleration and/or deceleration of the proton during the extraction step and tumor treatment without the use of a newly introduced magnetic field, such as by a hexapole magnet. Charged Particle Beam Intensity Control Control of applied field, such as a radio-frequency (RF) field, frequency and magnitude in the RF cavity system 310 allows for intensity control of the extracted proton beam, where intensity is extracted proton flux per unit time or the number of protons extracted as a function of time. Still referring FIG. 3, the intensity control system 305 is further described. In this example, an intensity control feedback loop is added to the extraction system, described supra. When protons in the proton beam hit the material 330 electrons are given off from the material 330 resulting in a current. The resulting current is converted to a voltage and is used as part of an ion beam intensity monitoring system or as part of an ion beam feedback loop for controlling beam intensity. The voltage is optionally measured and sent to the main controller 110 or to an intensity controller subsystem 340, which is preferably in communication or under the direction of the main controller 110. More particularly, when protons in the charged particle beam path pass through the material 330, some of the protons lose a small fraction of their energy, such as about one-tenth of a percent, which results in a secondary electron. That is, protons in the charged particle beam push some electrons when passing through material 330 giving the electrons enough energy to cause secondary emission. The resulting electron flow results in a current or signal that is proportional to the number of protons going through the target or extraction material 330. The resulting current is preferably converted to voltage and amplified. The resulting signal is referred to as a measured intensity signal. The amplified signal or measured intensity signal resulting from the protons passing through the material 330 is optionally used in monitoring the intensity of the extracted protons and is preferably used in controlling the intensity of the extracted protons. For example, the measured intensity signal is compared to a goal signal, which is predetermined in an irradiation of the tumor plan. The difference between the measured intensity signal and the planned for goal signal is calculated. The difference is used as a control to the RF generator. Hence, the measured flow of current resulting from the protons passing through the material 330 is used as a control in the RF generator to increase or decrease the number of protons undergoing betatron oscillation and striking the material 330. Hence, the voltage determined off of the material 330 is used as a measure of the orbital path and is used as a feedback control to control the RF cavity system. In one example, the intensity controller subsystem 340 preferably additionally receives input from: (1) a detector 350, which provides a reading of the actual intensity of the proton beam and/or (2) an irradiation plan 360. The irradiation plan provides the desired intensity of the proton beam for each x, y, energy, and/or rotational position of the patient/tumor as a function of time. Thus, the intensity controller 340 receives the desired intensity from the irradiation plan 350, the actual intensity from the detector 350 and/or a measure of intensity from the material 330, and adjusts the amplitude and/or the duration of application of the applied radio-frequency field in the RF cavity system 310 to yield an intensity of the proton beam that matches the desired intensity from the irradiation plan 360. As described, supra, the protons striking the material 330 is a step in the extraction of the protons from the synchrotron 130. Hence, the measured intensity signal is used to change the number of protons per unit time being extracted, which is referred to as intensity of the proton beam. The intensity of the proton beam is thus under algorithm control. Further, the intensity of the proton beam is controlled separately from the velocity of the protons in the synchrotron 130. Hence, intensity of the protons extracted and the energy of the protons extracted are independently variable. Still further, the intensity of the extracted protons is controllably variable while scanning the charged particles beam in the tumor from one voxel to an adjacent voxel as a separate hexapole and separated time period from acceleration and/or treatment is not required, as described supra. For example, protons initially move at an equilibrium trajectory in the synchrotron 130. An RF field is used to excite or move the protons into a betatron oscillation. In one case, the frequency of the protons orbit is about 10 MHz. In one example, in about one millisecond or after about 10,000 orbits, the first protons hit an outer edge of the target material 130. The specific frequency is dependent upon the period of the orbit. Upon hitting the material 130, the protons push electrons through the foil to produce a current. The current is converted to voltage and amplified to yield a measured intensity signal. The measured intensity signal is used as a feedback input to control the applied RF magnitude or RF field. An energy beam sensor, described infra, is optionally used as a feedback control to the RF field frequency or RF field of the RF field extraction system 310 to dynamically control, modify, and/or alter the delivered charge particle beam energy, such as in a continuous pencil beam scanning system operating to treat tumor voxels without alternating between an extraction phase and a treatment phase. Preferably, the measured intensity signal is compared to a target signal and a measure of the difference between the measured intensity signal and target signal is used to adjust the applied RF field in the RF cavity system 310 in the extraction system to control the intensity of the protons in the extraction step. Stated again, the signal resulting from the protons striking and/or passing through the material 130 is used as an input in RF field modulation. An increase in the magnitude of the RF modulation results in protons hitting the foil or material 130 sooner. By increasing the RF, more protons are pushed into the foil, which results in an increased intensity, or more protons per unit time, of protons extracted from the synchrotron 130. In another example, a detector 350 external to the synchrotron 130 is used to determine the flux of protons extracted from the synchrotron and a signal from the external detector is used to alter the RF field, RF intensity, RF amplitude, and/or RF modulation in the RF cavity system 310. Here the external detector generates an external signal, which is used in a manner similar to the measured intensity signal, described in the preceding paragraphs. Preferably, an algorithm or irradiation plan 360 is used as an input to the intensity controller 340, which controls the RF field modulation by directing the RF signal in the betatron oscillation generation in the RF cavity system 310. The irradiation plan 360 preferably includes the desired intensity of the charged particle beam as a function of time and/or energy of the charged particle beam as a function of time, for each patient rotation position, and/or for each x-, y-position of the charged particle beam. In yet another example, when a current from material 330 resulting from protons passing through or hitting material is measured beyond a threshold, the RF field modulation in the RF cavity system is terminated or reinitiated to establish a subsequent cycle of proton beam extraction. This process is repeated to yield many cycles of proton beam extraction from the synchrotron accelerator. In still yet another embodiment, intensity modulation of the extracted proton beam is controlled by the main controller 110. The main controller 110 optionally and/or additionally controls timing of extraction of the charged particle beam and energy of the extracted proton beam. The benefits of the system include a multi-dimensional scanning system. Particularly, the system allows independence in: (1) energy of the protons extracted and (2) intensity of the protons extracted. That is, energy of the protons extracted is controlled by an energy control system and an intensity control system controls the intensity of the extracted protons. The energy control system and intensity control system are optionally independently controlled. Preferably, the main controller 110 controls the energy control system and the main controller 110 simultaneously controls the intensity control system to yield an extracted proton beam with controlled energy and controlled intensity where the controlled energy and controlled intensity are independently variable and/or continually available as a separate extraction phase and acceleration phase are not required, as described supra. Thus the irradiation spot hitting the tumor is under independent control of: time; energy; intensity; x-axis position, where the x-axis represents horizontal movement of the proton beam relative to the patient, and y-axis position, where the y-axis represents vertical movement of the proton beam relative to the patient. In addition, the patient is optionally independently translated and/or rotated relative to a translational axis of the proton beam at the same time. Beam Transport The beam transport system 135 is used to move the charged particles from the accelerator to the patient, such as via a nozzle in a gantry, described infra. Charged Particle Energy The beam transport system 135 optionally includes means for determining an energy of the charged particles in the charged particle beam. For example, an energy of the charged particle beam is determined via calculation, such as via equation 1, using knowledge of a magnet geometry and applied magnetic field to determine mass and/or energy. Referring now to equation 1, for a known magnet geometry, charge, q, and magnetic field, B, the Larmor radius, ρL, or magnet bend radius is defined as: ρ L = v ⊥ Ω c = 2 ⁢ E m qB ( eq . ⁢ 1 ) where: ν⊥ is the ion velocity perpendicular to the magnetic field, Ωc is the cyclotron frequency, q is the charge of the ion, B is the magnetic field, m is the mass of the charge particle, and E is the charged particle energy. Solving for the charged particle energy yields equation 2. E = ( ρ L ⁢ qB ) 2 2 ⁢ m ( eq . ⁢ 2 ) Thus, an energy of the charged particle in the charged particle beam in the beam transport system 135 is calculable from the know magnet geometry, known or measured magnetic field, charged particle mass, charged particle charge, and the known magnet bend radius, which is proportional to and/or equivalent to the Larmor radius. Nozzle After extraction from the synchrotron 130 and transport of the charged particle beam along the proton beam path 268 in the beam transport system 135, the charged particle beam exits through the nozzle system 146. In one example, the nozzle system includes a nozzle foil covering an end of the nozzle system 146 or a cross-sectional area within the nozzle system forming a vacuum seal. The nozzle system includes a nozzle that expands in x/y-cross-sectional area along the z-axis of the proton beam path 268 to allow the proton beam 268 to be scanned along the x-axis and y-axis by the vertical control element and horizontal control element, respectively. The nozzle foil is preferably mechanically supported by the outer edges of an exit port of the nozzle or nozzle system 146. An example of a nozzle foil is a sheet of about 0.1 inch thick aluminum foil. Generally, the nozzle foil separates atmosphere pressures on the patient side of the nozzle foil from the low pressure region, such as about 10−5 to 10−7 torr region, on the synchrotron 130 side of the nozzle foil. The low pressure region is maintained to reduce scattering of the circulating charged particle beam in the synchrotron. Herein, the exit foil of the nozzle is optionally the first sheet 760 of the charged particle beam state determination system 750, described infra. Charged Particle Control Referring now to FIG. 4A, FIG. 4B, FIG. 5, FIG. 6A, and FIG. 6B, a charged particle beam control system is described where one or more patient specific beam control assemblies are removably inserted into the charged particle beam path proximate the nozzle of the charged particle cancer therapy system 100, where the patient specific beam control assemblies adjust the beam energy, diameter, cross-sectional shape, focal point, and/or beam state of the charged particle beam to properly couple energy of the charged particle beam to the individual's specific tumor. Beam Control Tray Referring now to FIG. 4A and FIG. 4B, a beam control tray assembly 400 is illustrated in a top view and side view, respectively. The beam control tray assembly 400 optionally comprises any of a tray frame 410, a tray aperture 412, a tray handle 420, a tray connector/communicator 430, and means for holding a patient specific tray insert 510, described infra. Generally, the beam control tray assembly 400 is used to: (1) hold the patient specific tray insert 510 in a rigid location relative to the beam control tray 400, (2) electronically identify the held patient specific tray insert 510 to the main controller 110, and (3) removably insert the patient specific tray insert 510 into an accurate and precise fixed location relative to the charged particle beam, such as the proton beam path 268 at the nozzle of the charged particle cancer therapy system 100. For clarity of presentation and without loss of generality, the means for holding the patient specific tray insert 510 in the tray frame 410 of the beam control tray assembly 400 is illustrated as a set of recessed set screws 415. However, the means for holding the patient specific tray insert 510 relative to the rest of the beam control tray assembly 400 is optionally any mechanical and/or electromechanical positioning element, such as a latch, clamp, fastener, clip, slide, strap, or the like. Generally, the means for holding the patient specific tray insert 510 in the beam control tray 400 fixes the tray insert and tray frame relative to one another even when rotated along and/or around multiple axes, such as when attached to a charged particle cancer therapy system 100, nozzle system 146, dynamic gantry nozzle, or gantry nozzle, which is an optional element of the nozzle system 146, that moves in three-dimensional space relative to a fixed point in the beamline, proton beam path 268, and/or a given patient position. As illustrated in FIG. 4A and FIG. 4B, the recessed set screws 415 fix the patient specific tray insert 510 into the aperture 412 of the tray frame 410. The tray frame 410 is illustrated as circumferentially surrounding the patient specific tray insert 510, which aids in structural stability of the beam control tray assembly 400. However, generally the tray frame 410 is of any geometry that forms a stable beam control tray assembly 400. Still referring to FIG. 4A and now referring to FIG. 5 and FIG. 6A, the optional tray handle 420 is used to manually insert/retract the beam control tray assembly 400 into a receiving element of the gantry nozzle, nozzle system 146, or dynamic gantry nozzle. While the beam control tray assembly 400 is optionally inserted into the charged particle beam path 268 at any point after extraction from the synchrotron 130, the beam control tray assembly 400 is preferably inserted into the positively charged particle beam proximate the nozzle system 146 or dynamic gantry nozzle as control of the beam shape is preferably done with little space for the beam shape to defocus before striking the tumor. Optionally, insertion and/or retraction of the beam control tray assembly 400 is semi-automated, such as in a manner of a digital-video disk player receiving a digital-video disk, with a selected auto-load and/or a selected auto-unload feature. Patient Specific Tray Insert Referring again to FIG. 5, a system of assembling trays 500 is described. The beam control tray assembly 400 optionally and preferably has interchangeable patient specific tray inserts 510, such as a range shifter insert 511, a patient specific ridge filter insert 512, an aperture insert 513, a compensator insert 514, or a blank insert 515. As described, supra, any of the range shifter insert 511, the patient specific ridge filter insert 512, the aperture insert 513, the compensator insert 514, or the blank insert 515 after insertion into the tray frame 410 are inserted as the beam control tray assembly 400 into the positively charged particle beam path 268, such as proximate the nozzle system 146 or dynamic gantry nozzle. Still referring to FIG. 5, the patient specific tray inserts 510 are further described. The patient specific tray inserts comprise a combination of any of: (1) a standardized beam control insert and (2) a patient specific beam control insert. For example, the range shifter insert or 511 or compensator insert 514 used to control the depth of penetration of the charged particle beam into the patient is optionally: (a) a standard thickness of a beam slowing material, such as a first thickness of Lucite, an acrylic, a clear plastic, and/or a thermoplastic material, (b) one member of a set of members of varying thicknesses and/or densities where each member of the set of members slows the charged particles in the beam path by a known amount, or (c) is a material with a density and thickness designed to slow the charged particles by a customized amount for the individual patient being treated, based on the depth of the individual's tumor in the tissue, the thickness of intervening tissue, and/or the density of intervening bone/tissue. Similarly, the ridge filter insert 512 used to change the focal point or shape of the beam as a function of depth is optionally: (1) selected from a set of ridge filters where different members of the set of ridge filters yield different focal depths or (2) customized for treatment of the individual's tumor based on position of the tumor in the tissue of the individual. Similarly, the aperture insert is: (1) optionally selected from a set of aperture shapes or (2) is a customized individual aperture insert 513 designed for the specific shape of the individual's tumor. The blank insert 515 is an open slot, but serves the purpose of identifying slot occupancy, as described infra. Slot Occupancy/Identification Referring again to FIG. 4A, FIG. 4B, and FIG. 5, occupancy and identification of the particular patient specific tray insert 510 into the beam control tray assembly 400 is described. Generally, the beam control tray assembly 400 optionally contains means for identifying, to the main controller 110 and/or a treatment delivery control system described infra, the specific patient tray insert 510 and its location in the charged particle beam path 268. First, the particular tray insert is optionally labeled and/or communicated to the beam control tray assembly 400 or directly to the main controller 110. Second, the beam control tray assembly 400 optionally communicates the tray type and/or tray insert to the main controller 110. In various embodiments, communication of the particular tray insert to the main controller 110 is performed: (1) directly from the tray insert, (2) from the tray insert 510 to the tray assembly 400 and subsequently to the main controller 110, and/or (3) directly from the tray assembly 400. Generally, communication is performed wirelessly and/or via an established electromechanical link. Identification is optionally performed using a radio-frequency identification label, use of a barcode, or the like, and/or via operator input. Examples are provided to further clarify identification of the patient specific tray insert 510 in a given beam control tray assembly 400 to the main controller. In a first example, one or more of the patient specific tray inserts 510, such as the range shifter insert 511, the patient specific ridge filter insert 512, the aperture insert 513, the compensator insert 514, or the blank insert 515 include an identifier 520 and/or and a first electromechanical identifier plug 530. The identifier 520 is optionally a label, a radio-frequency identification tag, a barcode, a 2-dimensional bar-code, a matrix-code, or the like. The first electromechanical identifier plug 530 optionally includes memory programmed with the particular patient specific tray insert information and a connector used to communicate the information to the beam control tray assembly 400 and/or to the main controller 110. As illustrated in FIG. 5, the first electromechanical identifier plug 530 affixed to the patient specific tray insert 510 plugs into a second electromechanical identifier plug, such as the tray connector/communicator 430, of the beam control tray assembly 400, which is described infra. In a second example, the beam control tray assembly 400 uses the second electromechanical identifier plug to send occupancy, position, and/or identification information related to the type of tray insert or the patient specific tray insert 510 associated with the beam control tray assembly to the main controller 110. For example, a first tray assembly is configured with a first tray insert and a second tray assembly is configured with a second tray insert. The first tray assembly sends information to the main controller 110 that the first tray assembly holds the first tray insert, such as a range shifter, and the second tray assembly sends information to the main controller 110 that the second tray assembly holds the second tray insert, such as an aperture. The second electromechanical identifier plug optionally contains programmable memory for the operator to input the specific tray insert type, a selection switch for the operator to select the tray insert type, and/or an electromechanical connection to the main controller. The second electromechanical identifier plug associated with the beam control tray assembly 400 is optionally used without use of the first electromechanical identifier plug 530 associated with the tray insert 510. In a third example, one type of tray connector/communicator 430 is used for each type of patient specific tray insert 510. For example, a first connector/communicator type is used for holding a range shifter insert 511, while a second, third, fourth, and fifth connector/communicator type is used for trays respectively holding a patient specific ridge filter insert 512, an aperture insert 513, a compensator insert 514, or a blank insert 515. In one case, the tray communicates tray type with the main controller. In a second case, the tray communicates patient specific tray insert information with the main controller, such as an aperture identifier custom built for the individual patient being treated. Tray Insertion/Coupling Referring now to FIG. 6A and FIG. 6B a beam control insertion process 600 is described. The beam control insertion process 600 comprises: (1) insertion of the beam control tray assembly 400 and the associated patient specific tray insert 510 into the charged particle beam path 268 and/or dynamic gantry nozzle 610, such as into a tray assembly receiver 620 and (2) an optional partial or total retraction of beam of the tray assembly receiver 620 into the dynamic gantry nozzle 610. Referring now to FIG. 6A, insertion of one or more of the beam control tray assemblies 400 and the associated patient specific tray inserts 510 into the dynamic gantry nozzle 610 is further described. In FIG. 6A, three beam control tray assemblies, of a possible n tray assemblies, are illustrated, a first tray assembly 402, a second tray assembly 404, and a third tray assembly 406, where n is a positive integer of 1, 2, 3, 4, 5 or more. As illustrated, the first tray assembly 402 slides into a first receiving slot 403, the second tray assembly 404 slides into a second receiving slot 405, and the third tray assembly 406 slides into a third receiving slot 407. Generally, any tray optionally inserts into any slot or tray types are limited to particular slots through use of a mechanical, physical, positional, and/or steric constraints, such as a first tray type configured for a first insert type having a first size and a second tray type configured for a second insert type having a second distinct size at least ten percent different from the first size. Still referring to FIG. 6A, identification of individual tray inserts inserted into individual receiving slots is further described. As illustrated, sliding the first tray assembly 402 into the first receiving slot 403 connects the associated electromechanical connector/communicator 430 of the first tray assembly 402 to a first receptor 626. The electromechanical connector/communicator 430 of the first tray assembly communicates tray insert information of the first beam control tray assembly to the main controller 110 via the first receptor 626. Similarly, sliding the second tray assembly 404 into the second receiving slot 405 connects the associated electromechanical connector/communicator 430 of the second tray assembly 404 into a second receptor 627, which links communication of the associated electromechanical connector/communicator 430 with the main controller 110 via the second receptor 627, while a third receptor 628 connects to the electromechanical connected placed into the third slot 407. The non-wireless/direct connection is preferred due to the high radiation levels within the treatment room and the high shielding of the treatment room, which both hinder wireless communication. The connection of the communicator and the receptor is optionally of any configuration and/or orientation. Tray Receiver Assembly Retraction Referring again to FIG. 6A and FIG. 6B, retraction of the tray receiver assembly 620 relative to a nozzle end 612 of the dynamic gantry nozzle 610 is described. The tray receiver assembly 620 comprises a framework to hold one or more of the beam control tray assemblies 400 in one or more slots, such as through use of a first tray receiver assembly side 622 through which the beam control tray assemblies 400 are inserted and/or through use of a second tray receiver assembly side 624 used as a backstop, as illustrated holding the plugin receptors configured to receive associated tray connector/communicators 430, such as the first, second, and third receptors 626, 627, 628. Optionally, the tray receiver assembly 620 retracts partially or completely into the dynamic gantry nozzle 610 using a retraction mechanism 660 configured to alternately retract and extend the tray receiver assembly 620 relative to a nozzle end 612 of the gantry nozzle 610, such as along a first retraction track 662 and a second retraction track 664 using one or more motors and computer control. Optionally the tray receiver assembly 620 is partially or fully retracted when moving the gantry, nozzle, and/or gantry nozzle 610 to avoid physical constraints of movement, such as potential collision with another object in the patient treatment room. For clarity of presentation and without loss of generality, several examples of loading patient specific tray inserts into tray assemblies with subsequent insertion into an positively charged particle beam path proximate a gantry nozzle 610 are provided. In a first example, a single beam control tray assembly 400 is used to control the charged particle beam 268 in the charged particle cancer therapy system 100. In this example, a patient specific range shifter insert 511, which is custom fabricated for a patient, is loaded into a patient specific tray insert 510 to form a first tray assembly 402, where the first tray assembly 402 is loaded into the third receptor 628, which is fully retracted into the gantry nozzle 610. In a second example, two beam control assemblies 400 are used to control the charged particle beam 268 in the charged particle cancer therapy system 100. In this example, a patient specific ridge filter 512 is loaded into a first tray assembly 402, which is loaded into the second receptor 627 and a patient specific aperture 513 is loaded into a second tray assembly 404, which is loaded into the first receptor 626 and the two associated tray connector/communicators 430 using the first receptor 626 and second receptor 627 communicate to the main controller 110 the patient specific tray inserts 510. The tray receiver assembly 620 is subsequently retracted one slot so that the patient specific ridge filter 512 and the patient specific aperture reside outside of and at the nozzle end 612 of the gantry nozzle 610. In a third example, three beam control tray assemblies 400 are used, such as a range shifter 511 in a first tray inserted into the first receiving slot 403, a compensator in a second tray inserted into the second receiving slot 405, and an aperture in a third tray inserted into the third receiving slot 407. Generally, any patient specific tray insert 510 is inserted into a tray frame 410 to form a beam control tray assembly 400 inserted into any slot of the tray receiver assembly 620 and the tray assembly is not retracted or retracted any distance into the gantry nozzle 610. Tomography/Beam State In one embodiment, the charged particle tomography apparatus is used to image a tumor in a patient. As current beam position determination/verification is used in both tomography and cancer therapy treatment, for clarity of presentation and without limitation beam state determination is also addressed in this section. However, beam state determination is optionally used separately and without tomography. In another example, the charged particle tomography apparatus is used in combination with a charged particle cancer therapy system using common elements. For example, tomographic imaging of a cancerous tumor is performed using charged particles generated with an injector, accelerator, and guided with a delivery system that are part of the cancer therapy system, described supra. In various examples, the tomography imaging system is optionally simultaneously operational with a charged particle cancer therapy system using common elements, allows tomographic imaging with rotation of the patient, is operational on a patient in an upright, semi-upright, and/or horizontal position, is simultaneously operational with X-ray imaging, and/or allows use of adaptive charged particle cancer therapy. Further, the common tomography and cancer therapy apparatus elements are optionally operational in a multi-axis and/or multi-field raster beam mode. In conventional medical X-ray tomography, a sectional image through a body is made by moving one or both of an X-ray source and the X-ray film in opposite directions during the exposure. By modifying the direction and extent of the movement, operators can select different focal planes, which contain the structures of interest. More modern variations of tomography involve gathering projection data from multiple directions by moving the X-ray source and feeding the data into a tomographic reconstruction software algorithm processed by a computer. Herein, in stark contrast to known methods, the radiation source is a charged particle, such as a proton ion beam or a carbon ion beam. A proton beam is used herein to describe the tomography system, but the description applies to a heavier ion beam, such as a carbon ion beam. Further, in stark contrast to known techniques, herein the radiation source is preferably stationary while the patient is rotated. Referring now to FIG. 7, an example of a tomography apparatus is described and an example of a beam state determination is described. In this example, the tomography system 700 uses elements in common with the charged particle beam system 100, including elements of one or more of the injection system 120, the accelerator 130, a positively charged particle beam transport path 268 within a beam transport housing 320 in the beam transport system 135, the targeting/delivery system 140, the patient interface module 150, the display system 160, and/or the imaging system 170, such as the X-ray imaging system. The scintillation material is optionally one or more scintillation plates, such as a scintillating plastic, used to measure energy, intensity, and/or position of the charged particle beam. For instance, a scintillation material 710 or scintillation plate is positioned behind the patient 730 relative to the targeting/delivery system 140 elements, which is optionally used to measure intensity and/or position of the charged particle beam after transmitting through the patient. Optionally, a second scintillation plate or a charged particle induced photon emitting sheet, described infra, is positioned prior to the patient 730 relative to the targeting/delivery system 140 elements, which is optionally used to measure incident intensity and/or position of the charged particle beam prior to transmitting through the patient. The charged particle beam system 100 as described has proven operation at up to and including 330 MeV, which is sufficient to send protons through the body and into contact with the scintillation material. Particularly, 250 MeV to 330 MeV are used to pass the beam through a standard sized patient with a standard sized pathlength, such as through the chest. The intensity or count of protons hitting the plate as a function of position is used to create an image. The velocity or energy of the proton hitting the scintillation plate is also used in creation of an image of the tumor 720 and/or an image of the patient 730. The patient 730 is rotated about the y-axis and a new image is collected. Preferably, a new image is collected with about every one degree of rotation of the patient resulting in about 360 images that are combined into a tomogram using tomographic reconstruction software. The tomographic reconstruction software uses overlapping rotationally varied images in the reconstruction. Optionally, a new image is collected at about every 2, 3, 4, 5, 10, 15, 30, or 45 degrees of rotation of the patient. Herein, the scintillation material 710 or scintillator is any material that emits a photon when struck by a positively charged particle or when a positively charged particle transfers energy to the scintillation material sufficient to cause emission of light. Optionally, the scintillation material emits the photon after a delay, such as in fluorescence or phosphorescence. However, preferably, the scintillator has a fast fifty percent quench time, such as less than 0.0001, 0.001, 0.01, 0.1, 1, 10, 100, or 1,000 milliseconds, so that the light emission goes dark, falls off, or terminates quickly. Preferred scintillation materials include sodium iodide, potassium iodide, cesium iodide, an iodide salt, and/or a doped iodide salt. Additional examples of the scintillation materials include, but are not limited to: an organic crystal, a plastic, a glass, an organic liquid, a luminophor, and/or an inorganic material or inorganic crystal, such as barium fluoride, BaF2; calcium fluoride, CaF2, doped calcium fluoride, sodium iodide, NaI; doped sodium iodide, sodium iodide doped with thallium, NaI(Tl); cadmium tungstate, CdWO4; bismuth germanate; cadmium tungstate, CdWO4; calcium tungstate, CaWO4; cesium iodide, CsI; doped cesium iodide; cesium iodide doped with thallium, CsI(Tl); cesium iodide doped with sodium CsI(Na); potassium iodide, KI; doped potassium iodide, gadolinium oxysulfide, Gd2O2S; lanthanum bromide doped with cerium, LaBr3(Ce); lanthanum chloride, LaCl3; cesium doped lanthanum chloride, LaCl3(Ce); lead tungstate, PbWO4; LSO or lutetium oxyorthosilicate (Lu2SiO5); LYSO, Lu1.8Y0.2SiO5(Ce); yttrium aluminum garnet, YAG(Ce); zinc sulfide, ZnS(Ag); and zinc tungstate, ZnWO4. In one embodiment, a tomogram or an individual tomogram section image is collected at about the same time as cancer therapy occurs using the charged particle beam system 100. For example, a tomogram is collected and cancer therapy is subsequently performed: without the patient moving from the positioning systems, such as in a semi-vertical partial immobilization system, a sitting partial immobilization system, or the a laying position. In a second example, an individual tomogram slice is collected using a first cycle of the accelerator 130 and using a following cycle of the accelerator 130, the tumor 720 is irradiated, such as within about 1, 2, 5, 10, 15 or 30 seconds. In a third case, about 2, 3, 4, or 5 tomogram slices are collected using 1, 2, 3, 4, or more rotation positions of the patient 730 within about 5, 10, 15, 30, or 60 seconds of subsequent tumor irradiation therapy. In another embodiment, the independent control of the tomographic imaging process and X-ray collection process allows simultaneous single and/or multi-field collection of X-ray images and tomographic images easing interpretation of multiple images. Indeed, the X-ray and tomographic images are optionally overlaid to from a hybrid X-ray/proton beam tomographic image as the patient 730 is optionally in the same position for each image. In still another embodiment, the tomogram is collected with the patient 730 in the about the same position as when the patient's tumor is treated using subsequent irradiation therapy. For some tumors, the patient being positioned in the same upright or semi-upright position allows the tumor 720 to be separated from surrounding organs or tissue of the patient 730 better than in a laying position. Positioning of the scintillation material 710 behind the patient 730 allows the tomographic imaging to occur while the patient is in the same upright or semi-upright position. The use of common elements in the tomographic imaging and in the charged particle cancer therapy allows benefits of the cancer therapy, described supra, to optionally be used with the tomographic imaging, such as proton beam x-axis control, proton beam y-axis control, control of proton beam energy, control of proton beam intensity, timing control of beam delivery to the patient, rotation control of the patient, and control of patient translation all in a raster beam mode of proton energy delivery. The use of a single proton or cation beamline for both imaging and treatment facilitates eases patient setup, reduces alignment uncertainties, reduces beam state uncertainties, and eases quality assurance. In yet still another embodiment, initially a three-dimensional tomographic proton based reference image is collected, such as with hundreds of individual rotation images of the tumor 720 and patient 730. Subsequently, just prior to proton treatment of the cancer, just a few 2-dimensional control tomographic images of the patient are collected, such as with a stationary patient or at just a few rotation positions, such as an image straight on to the patient, with the patient rotated about 45 degrees each way, and/or the patient rotated about 90 degrees each way about the y-axis. The individual control images are compared with the 3-dimensional reference image. An adaptive proton therapy is subsequently performed where: (1) the proton cancer therapy is not used for a given position based on the differences between the 3-dimensional reference image and one or more of the 2-dimensional control images and/or (2) the proton cancer therapy is modified in real time based on the differences between the 3-dimensional reference image and one or more of the 2-dimensional control images. Charged Particle State Determination/Verification/Photonic Monitoring Still referring to FIG. 7, the tomography system 700 is optionally used with a charged particle beam state determination system 750, optionally used as a charged particle verification system. The charged particle state determination system 750 optionally measures, determines, and/or verifies one of more of: (1) position of the charged particle beam, such as the treatment beam 269, (2) direction of the treatment beam 269, (3) intensity of the treatment beam 269, (4) energy of the treatment beam 269, (5) position, direction, intensity, and/or energy of the charged particle beam, such as a residual charged particle beam 267 after passing through a sample or the patient 730, and (6) a history of the charged particle beam. For clarity of presentation and without loss of generality, a description of the charged particle beam state determination system 750 is described and illustrated separately in FIG. 8 and FIG. 9A; however, as described herein elements of the charged particle beam state determination system 750 are optionally and preferably integrated into the nozzle system 146 and/or the tomography system 700 of the charged particle treatment system 100. More particularly, any element of the charged particle beam state determination system 750 is integrated into the nozzle system 146, the dynamic gantry nozzle 610, and/or tomography system 700, such as a surface of the scintillation material 710 or a surface of a scintillation detector, plate, or system. The nozzle system 146 or the dynamic gantry nozzle 610 provides an outlet of the charged particle beam from the vacuum tube initiating at the injection system 120 and passing through the synchrotron 130 and beam transport system 135. Any plate, sheet, fluorophore, or detector of the charged particle beam state determination system is optionally integrated into the nozzle system 146. For example, an exit foil of the nozzle 610 is optionally a first sheet 760 of the charged particle beam state determination system 750 and a first coating 762 is optionally coated onto the exit foil, as illustrated in FIG. 7. Similarly, optionally a surface of the scintillation material 710 is a support surface for a fourth coating 792, as illustrated in FIG. 7. The charged particle beam state determination system 750 is further described, infra. Referring now to FIG. 7, FIG. 8, and FIG. 9A, four sheets, a first sheet 760, a second sheet 770, a third sheet 780, and a fourth sheet 790 are used to illustrated detection sheets and/or photon emitting sheets upon transmittance of a charged particle beam. Each sheet is optionally coated with a photon emitter, such as a fluorophore, such as the first sheet 760 is optionally coated with a first coating 762. Without loss of generality and for clarity of presentation, the four sheets are each illustrated as units, where the light emitting layer is not illustrated. Thus, for example, the second sheet 770 optionally refers to a support sheet, a light emitting sheet, and/or a support sheet coated by a light emitting element. The four sheets are representative of n sheets, where n is a positive integer. Referring now to FIG. 7 and FIG. 8, the charged particle beam state verification system 750 is a system that allows for monitoring of the actual charged particle beam position in real-time without destruction of the charged particle beam. The charged particle beam state verification system 750 preferably includes a first position element or first beam verification layer, which is also referred to herein as a coating, luminescent, fluorescent, phosphorescent, radiance, or viewing layer. The first position element optionally and preferably includes a coating or thin layer substantially in contact with a sheet, such as an inside surface of the nozzle foil, where the inside surface is on the synchrotron side of the nozzle foil. Less preferably, the verification layer or coating layer is substantially in contact with an outer surface of the nozzle foil, where the outer surface is on the patient treatment side of the nozzle foil. Preferably, the nozzle foil provides a substrate surface for coating by the coating layer. Optionally, a binding layer is located between the coating layer and the nozzle foil, substrate, or support sheet. Optionally, the position element is placed anywhere in the charged particle beam path. Optionally, more than one position element on more than one sheet, respectively, is used in the charged particle beam path and is used to determine a state property of the charged particle beam, as described infra. Still referring to FIG. 7 and FIG. 8, the coating, referred to as a fluorophore, yields a measurable spectroscopic response, spatially viewable by a detector or camera, as a result of transmission by the proton beam. The coating is preferably a phosphor, but is optionally any material that is viewable or imaged by a detector where the material changes spectroscopically as a result of the charged particle beam hitting or transmitting through the coating or coating layer. A detector or camera views secondary photons emitted from the coating layer and determines a position of a treatment beam 269, which is also referred to as a current position of the charged particle beam or final treatment vector of the charged particle beam, by the spectroscopic differences resulting from protons and/or charged particle beam passing through the coating layer. For example, the camera views a surface of the coating surface as the proton beam or positively charged cation beam is being scanned by the first axis control 143, vertical control, and the second axis control 144, horizontal control, beam position control elements during treatment of the tumor 720. The camera views the current position of the charged particle beam or treatment beam 269 as measured by spectroscopic response. The coating layer is preferably a phosphor or luminescent material that glows and/or emits photons for a short period of time, such as less than 5 seconds for a 50% intensity, as a result of excitation by the charged particle beam. The detector observes the temperature change and/or observe photons emitted from the charged particle beam traversed spot. Optionally, a plurality of cameras or detectors are used, where each detector views all or a portion of the coating layer. For example, two detectors are used where a first detector views a first half of the coating layer and the second detector views a second half of the coating layer. Preferably, at least a portion of the detector is mounted into the nozzle system to view the proton beam position after passing through the first axis and second axis controllers 143, 144. Preferably, the coating layer is positioned in the proton beam path 268 in a position prior to the protons striking the patient 730. Referring now to FIG. 1 and FIG. 7, the main controller 110, connected to the camera or detector output, optionally and preferably compares the final proton beam position or position of the treatment beam 269 with the planned proton beam position and/or a calibration reference to determine if the actual proton beam position or position of the treatment beam 269 is within tolerance. The charged particle beam state determination system 750 preferably is used in one or more phases, such as a calibration phase, a mapping phase, a beam position verification phase, a treatment phase, and a treatment plan modification phase. The calibration phase is used to correlate, as a function of x-, y-position of the glowing response the actual x-, y-position of the proton beam at the patient interface. During the treatment phase, the charged particle beam position is monitored and compared to the calibration and/or treatment plan to verify accurate proton delivery to the tumor 720 and/or as a charged particle beam shutoff safety indicator. Referring now to FIG. 10, the position verification system 179 and/or the treatment delivery control system 112, upon determination of a tumor shift, an unpredicted tumor distortion upon treatment, and/or a treatment anomaly optionally generates and or provides a recommended treatment change 1070. The treatment change 1070 is optionally sent out while the patient 730 is still in the treatment position, such as to a proximate physician or over the internet to a remote physician, for physician approval 1072, receipt of which allows continuation of the now modified and approved treatment plan.
050383701	abstract	The invention relates to an arrangement for generating an X-ray or gamma beam with small cross-section and variable direction, having an X-ray or gamma emitter, from the focus of which a bundle of rays emerges, and a diaphragm arrangement, which cuts out a beam from the bundle of rays and comprises a hollow-cylindrical first diaphragm body which is rotatable about its axis of symmetry and has two mutually offset helical slits on the circumference. In this arrangement, an X-ray beam with at least approximately square cross-section is cut out on a relatively long hollow-cylindrical body with small diameter by the slits winding around the diaphragm body in at least one turn each and being shaped in such a way that at least one straight line runs through the slits towards the focus, the position of which line can be varied by turning the diaphragm body.
	summary
043740839	summary	The invention relates to a nuclear reactor with a liquid coolant, especially water, in a cooling loop to which hydrogen is added, and having a volume control surge tank for the coolant and a high-pressure pump which feeds coolant taken from the cooling loop back into the cooling loop after purification. In the book "VGB-Kernkradtwerks-Seminar 1970", especially on page 41 thereof, a volume control system for a pressurized-water reactor is described that is used, among other things, for feeding-in chemicals. Part of this system, through which a portion of the reactor cooling water continuously flows, is also charged with hydrogen. This is to counteract the radiolytic decomposition of the coolant in the vicinity of the core. In the known system, the hydrogen is added into the volume control surge tank, which is part of the volume control system, and is therefore present above the liquid level as a gas cushion. The partial H.sub.2 pressure in the gas cushion is set in accordance with the desired hydrogen concentration in the cooling water. As the hereinafore-mentioned reference literature shows, the volume control system contains high-pressure pumps because the cooling system must be fed back into the primary loop of the pressurized-water reactor in which, as is well-known, pressures of, for instance, 160 bar prevail. It is accordingly an object of the invention to provide a nuclear reactor having a liquid coolant which overcomes the hereinafore-mentioned disadvantages of the heretofore known devices of this general type, and to modify the known volume control system with the objective of reducing the hydrogen components present in the volume control surge tank outside the volume of liquid, so that the danger of oxyhydrogen gas explosions if leaks occur is completely eliminated. With the foregoing and other objects in view, there is provided, in accordance with the invention, a nuclear reactor with a liquid (especially water) coolant loop having a volume control surge tank for the coolant disposed in the loop, and a high pressure pump disposed in the loop for feeding coolant taken from the loop back into the loop after purification, comprising a line bypassing the volume control surge tank and having an end connected to the suction side of the high-pressure pump, and means for introducing hydrogen into a liquid-filled section of the loop on the suction side of the high-pressure pump. With the use of the invention, the hydrogen to be added is directly fed into the liquid. Thus, it is no longer necessary to place very stringent requirements on the tightness of the volume control surge tank, since a flammable mixture can no longer be produced in the event of leaks. In accordance with another feature of the invention, the hydrogen introducing means feeds hydrogen gas to the bypass line. In accordance with a further feature of the invention, the hydrogen introducing means mixes hydrogen gas with coolant bled off from the volume control surge tank. In accordance with an additional feature of the invention, there are provided means for mixing hydrogen and coolant, disposed up-stream of the hydrogen introducing means, in flow direction of the coolant. The mixing means or section is a section of line with built-in components which causes deflection of the liquid flowing through it and thereby effects thorough mixing of added gas components. In accordance with an added feature of the invention, there is provided a gas separator disposed upstream of the mixing means, in the flow direction of the coolant. This prevents the occurrence of large gas bubbles due to excess hydrogen on the suction side of the high-pressure pump. In accordance with still another feature of the invention, there is provided a hydrogen delivery source connected to the hydrogen introducing means, and means connected to the gas separator and the hydrogen introducing means, for controlling the delivery rate from the hydrogen delivery source. The hydrogen source can be commercially available gas bottles with a suitable regulating device which is, for instance, a reducing valve. In accordance with still a further feature of the invention, there is provided a gas outlet line connected from the gas separator to the hydrogen introducing means, and a compressor preferably a diaphragm compressor, disposed in the gas outlet line. In this way, the hydrogen collected in the gas separator is again transported back into the liquid-filled line section with the feeding point or hydrogen introducing means. Further addition of hydrogen can then be dispensed with until the control device connected to the gas separator determines a demand for hydrogen to be added. It is assumed here that the gas which is separated in the gas separator behind the hydrogen feed-in point is substantially all hydrogen. The hydrogen content can be determined with measuring devices such as a process chromatograph. Other gas analyzers can also be used; the hydrogen content need not necessarily be determined continuously but can be performed at certain time intervals. In accordance with still an additional feature of the invention, the hydrogen introducing means includes a ceramic filter cartridge. Thereby, a fine distribution is achieved, which promotes the dissolution of the hydrogen in the coolant. In accordance with a concomitant feature of the invention, there is provided an exhaust gas system connected to the volume control surge tank above the coolant level. The gas space of the volume control surge tank, which in the invention is no longer required for hydrogen enrichment, should be connected to the exhaust gas system in such a manner that, contrary to known practice, the hydrogen content in the gas space remains reliably below 4%. Since the danger of oxyhydrogen gas explosions exists only above 4%, the connection to the exhaust gas system is used here to keep the amount of hydrogen small, while in the known system, a pure hydrogen cushion is present for the purpose of charging with gas. 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 nuclear reactor having a liquid coolant, 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.
	summary
	description	This application is a National Stage Application of International Application Number PCT/EP2006/000324, filed Jan. 16, 2006; which claims the benefit of U.S. Provisional Application No. 60/644,182, filed Jan. 14, 2005, in its entirety. The present invention relates to an universal method for the large scale production of high-purity carrier free or non carrier added radioisotopes by applying a number of “unit operations” which are derived from physics and material science and hitherto not used for isotope production. A required number of said unit operations is combined, selected and optimised individually for each radioisotope production scheme. The use of said unit operations allows a batch wise operation or a fully automated continuous production scheme. The radioisotopes produced by the inventive method are especially suitable for producing radioisotope-labelled bioconjugates as well as particles, in particular nanoparticles and microparticles. Radioisotopes are widely used in the fields of life science, research and medicine, for example, in nuclear medicine, diagnosis, radiotherapy, biochemical analysis, as well as diagnostic and therapeutic pharmaceuticals. One such important application for radioisotopes is the diagnosis and therapy of diseases, such as cancer. For example, there has been considerable progress during the last two decades in the use of radio-labelled tumor-selective monoclonal antibodies in the diagnosis and therapy of cancer. The concept of localizing the cytotoxic radionuclide to the cancer cell is an important supplement to conventional forms of radiotherapy. In theory the intimate contract between a radioactive antibody conjugate and a target cell enables the absorbed radiation dose to be concentrated at the site of abnormality with minimal injury to the normal surrounding cells and tissues [Bruland O S. Cancer therapy with radiolabelled antibodies. An overview. Acta Oncol. 1995; 34(8):1085-94]. Furthermore, the use of monoclonal antibodies to deliver radioisotopes directly to tumor cells has become a promising strategy to enhance the antitumor effects of native antibodies. Since the alpha- and beta-particles emitted during the decay of radioisotopes differ in significant ways, proper selection of isotope and antibody combinations is crucial to making radioimmunotherapy a standard therapeutic modality. Because of the short path length (50-80 microm) and high linear energy transfer (approximately 100 keV/microm) of alpha-emitting radioisotopes, targeted alpha-particle therapy offers the potential for more specific tumor cell killing with less damage to surrounding normal tissues than beta-emitters. These properties make targeted alpha-particle therapy ideal for the elimination of minimal residual or micrometastatic disease. Radioimmunotherapy using alpha-emitters such as (213)Bi, (211)At, and (225)Ac has shown activity in several in vitro and in vivo experimental models as well as in clinical trials. Further advances will require investigation of more potent isotopes, new sources and methods of isotope production, improved chelation techniques, better methods for pharmacokinetic and dosimetric modeling, and new methods of isotope delivery such as pretargeting. [Mulford D A, Scheinberg D A, Jurcic J G. The promise of targeted alpha-particle therapy. J Nucl Med. 2005 January; 46 Suppl 1:199 S-204S.] In addition, radioimmunotherapy (RIT) combines the advantages of targeted radiation therapy and specific immunotherapy using monoclonal antibodies. RIT can be used either to target tumor cells or to specifically suppress immunocompetent host cells in the setting of allogeneic transplantation. The choice of radionuclide used for RIT depends on its distinct radiation characteristics and the type of malignancy or cells targeted. In general, beta-emitters with their lower energy and longer path length are more suitable to target bulky, solid tumors whereas alpha-emitters with their high linear energy transfer and short path length are better suited to target hematopoietic cells (normal or malignant). Different approaches of RIT such as the use of stable radioimmunoconjugates or of pretargeting strategies are available. [Bethge W A, Sandmaier B M. Targeted cancer therapy using radiolabeled monoclonal antibodies. Technol Cancer Res Treat. 2005 August; 4(4):393-405. Also the method SIRT (selective internal radiation therapy) or radioembolization has been developed which is similar to chemoembolization but uses radioactive microspheres (microscopic particles or beads). Thereby, radioisotopes are incorporated directly into the microspheres in order to deliver radiation directly to its destination, e.g. the tumor. The loaded spheres/beads are e.g. injected through a catheter into the blood vessel supplying the tumor. The spheres/beads become lodged within the tumor vessels where they deliver local radiation that causes tumor death. This technique allows for a higher dose of radiation to be used to kill the tumor without subjecting adjacent healthy tissue to harmful levels of radiation. Radioembolization has been described utilizing, for example, 90Y (Herba M J, Thirlwell M P. Radioembolization for hepatic metastases. Semin Oncol. 2002 April; 29(2):152-9.) or 188Re (Wunderlich G, Pinkert J, Stintz M, Kotzerke J. Labeling and biodistribution of different particle materials for radioembolization therapy with 188Re. Appl Radiat Isot. 2005 May; 62(5):745-50.) However, the presently used methods in radioisotope production have reached their limits and there is a strong need for improved methods. This applies in particular to the isotopic purity, the specific activity and the range of available radionuclides. With the growing complexity of positron emission tomography (PET)/single photon emission computed tomography (SPECT) imaging and the developments in systemic radionuclide therapy there is a growing need for radioisotope preparations with higher radiochemical and radionuclic purity that has not been achievable before. Especially important for the new applications is the specific activity of the radiotracer. Furthermore, an implementation of the break-through in development of the drug target delivery systems of new methods of cancer therapy is limited due to the lack of availability of the existing radionuclides with optimal decay characteristics for such an application. An object of the present invention is, thus, to provide a method for the large scale production of high-purity radioisotopes, especially of carrier free or non carrier added radioisotopes. Another object of the present invention is, thus, to provide uses of these radioisotopes. The invention relates to a general method for industrial scale production of radioisotope preparations for life science research, medical application and industry. In particular it opens up for mass production of a number of rare isotopes that hitherto have not been available on the market and now are much in demand. By combining a number of physics unit operations with radiochemical unit operations the method allows to extract and refine any useful radioisotope from a suitable activated material in a non destructive and reusable way that generates a minimum of waste and almost no liquid waste. According to the method of the present invention target material activated by any method can be used as raw material. A number of the isotopes of interest are abundantly produced by the high energy nuclear reactions that occur as by product in present and future high energy particle accelerators, experiments and other accelerator driven systems. In those facilities the method of the present invention permits to harvest the radioisotopes from their various waste products, their molten metal target and cooling media and spent beam absorbers or if needed from dedicated target stations sharing the primary particle beam. According to the method of the present invention extraction of radionuclides from the irradiated material and their subsequent concentration and purification into monoisotopic samples is achieved by application of a number of innovative “unit operations” (see below, units 1-14) derived from physics and material science and hitherto not used for isotope production. The required number of these unit operations of the present invention are combined, selected, put in the required order and optimised individually for each radioisotope production scheme. They allow a batch wise operation or a fully automated continuous production scheme. In the following a list is given of these unit operations that also can be further combined if needed with more conventional radiochemical methods in order to obtain a given product: Unit 1: Activation (i.e. irradiation with charged particles, neutrons, electrons or gamma-rays) of target materials that allow pyrochemical or pyrometallurgical treatment to produce the radioisotopes of interest or their predecessors. Unit 2: Transport of the element in question to the surface of the target material is accomplished by means of high temperature diffusion in the solid or liquid target matrix. Unit 3: Separation of the element in question from the bulk target material can be achieved by high temperature desorption from the target surface under vacuum or in inert atmosphere (e.g. He, Ar, . . . ). Unit 4: Separation of the element in question from the bulk target material can be achieved by removing the target material by high temperature sublimation under vacuum or in inert atmosphere if the element in question is less volatile than the target material. Unit 5: Separation of the element in question from the bulk target material can be achieved by adsorption on suitable substrates located in the flow of a liquid metal target and coolant medium. Unit 6: Desorption of the element in question from the bulk target material can be assisted by means of the chemical evaporation technique, i.e. the addition of chemical reactive gases that form in-situ more volatile compounds of the element in question. Unit 7: Transport of the element or chemical compound in question to further purification steps is accomplished by molecular flow at high temperature or by a gas flow. Unit 8: Condensation or adsorption on a surface compatible with the purity requirement of an accelerator ion-source. Unit 9: Conditioning for ionisation in the ion sources by addition of suitable chemicals that either allow pyrochemical reduction to the elementary state or oxidation/molecule formation on the other hand and controlling the mass separation process i.e. mass marking. Unit 10: Introduction of the sample into an oven from where the sample is fed into the ion source by raising the oven temperature in a controlled way. Unit 11: Use of various types of ion-sources optimised for an isotope of the element in question, e.g. surface ionisation, resonant laser ionisation or plasma ionisation. Unit 12: Acceleration of the radioactive ion-beam extracted from the ion source with a dc or ac acceleration voltage. Unit 13: Separation of the ion beam in a suitable mass selective device, e.g. a magnetic sector field, a Wien-filter or a radio-frequency multipole. Unit 14: Use is made of the momentum imparted to the mass separated nuclides in order to collect them by implantation into a suitably prepared chemical substrate, e.g. nanoparticles or microparticles, macromolecules, microspheres, macroaggregates, ion exchange resins or other matrices used in chromatographic systems. Application unit: Application of the obtained isotopes in research and medicine, for diagnosis and/or therapy of diseases, such as in vivo and in vitro applications, e.g. RIT, biodistribution studies, PET imaging, SPECT, gamma-spectrometry, TAT, radioembolization, Auger-therapy etc. Unit operation 1 is also called the “production” unit operation. Unit operations 2-14 are also called the “separation” unit operations. Although the method of the invention (units 1-14) allows harvesting the radioisotopes independent on the mode of activation the synergy with present and future high energy particle accelerators, experiments and other accelerator driven systems is obvious. A number of the isotopes of interest are abundantly produced by the high energy nuclear reactions that occur as by-product in various locations: 1. Target of the type where a circulating molten metal is used as combined target and heat transfer medium. In a bypass line of this metal flow the radio isotopes of interest can be continuously extracted. 2. Any sufficiently irradiated structure disposed of as waste. 3. Dedicated targets and ion-source units irradiated in the primary particle beam or in its spent beam absorber. Finally the method of the present invention lends itself to build a radioisotope factory in which the radioisotopes are produced on-line in a continuous process where dedicated target and mass separator stations share the primary beam. Improvements and Advantages The mass separating step of the method according to the invention fulfils the newly formulated higher quality standards by producing mono isotopic samples without any stable isotope of the element in question. This form features the highest possible and achievable specific activity of a radionuclide, also called “carrier free”. Almost all useable nuclides in the chart of nuclides can be produced so that radionuclides that are better adapted to their applications can be selected in amounts that also allow widespread use of the upcoming methods for radiotherapy. The method is independent of the nuclear reaction used to produce the radioactivity. The method allows a cost efficient extraction of the wanted nuclei from a number of by products available in present and future accelerator projects and to facilitate the control and disposal of their radioactive waste inventory. The inclusion of ion-beam formation and acceleration as production stages facilitates the process of labelling of the pharmaceutical end product and production of new isotope generators. This method uses rather non destructive dry techniques that often allow reusing the target and mainly produces solid waste products with much less liquid waste as in the present production that proceeds via dissolution of the targets. The radioisotope labelled bioconjugates preferably can be used in radio-immunotherapy of diseases, such as cancer, e.g., in targeted alpha therapy (TAT). The method which is provided by the present invention preferably comprises the following steps: (a) Activation of a target by a particle beam, (b) Separation of the isotope from the irradiated target, (c) Ionisation of the separated isotope, (d) Extraction from the ion source and acceleration of the ion beam, (e) Mass-separation, (f) Collection of the isotope.wherein step (a) comprises unit operation 1,wherein step (b) comprises unit operation 2, 3, 4 and/or 5,wherein step (c) comprises unit operation 11,wherein step (d) comprises unit operation 12,wherein step (e) comprises unit operation 13, andwherein step (f) comprises unit operation 14. Thus, one preferred combination of the unit operations utilizes units 1 and 2 (or 3 or 4 or 5) and 11-14. A combination of units 1, 2, 3 and 11-14 is preferred, such as for the production of carrier-free radioisotopes of the rare earth elements. A combination of units 1, 2, 3, 7 and 11-14 is preferred, such as for the on- or off-line extraction of radioisotopes from a high power liquid metal target, for the production of radioisotopes relevant for targeted alpha therapy (TAT) via continuous or batch-mode extraction from actinide targets, for the on-line production of carrier-free 204-210At as well as for the production of carrier-free radioisotopes of the rare earth elements. Furthermore, a combination of units 1, 2, 3, 7, 8, 10 and 11-14 is preferred, such as for the on-line production of carrier-free 204-210At. Also the combination of units 1, 2, 3, 7 and 8 is suitable, such as for the on-line production of carrier-free 211At or 204-210At as well as for the production of carrier-free radioisotopes of the rare earth elements. Further preferred combinations are the combinations of units 1, 2, 4 (or 5), 8, 10 and 11-14; units 1, 7, 8, 10 and 11-14; units 1, 2, 3, 10, and 11-14; units 1, 2, 3, 6, 7 and 11-14. The combinations of units 1, 9, 10, 11, 12 and 13 as well as units 1, 2, 3, 7, 9, 11, 12 and 13 are also preferred, such as for the fission production of neutron-rich lanthanide and tin isotopes. Furthermore, for the fission production of isotopes, such as neutron-rich lanthanide and tin isotopes, a method comprising the following steps is preferred: (a) Activation of a fission target by a particle beam, (b) Separation of the isotope from the irradiated target, and optionally, (c) uonisation of the separated isotope, optionally, (d) Extraction from the ion source and acceleration of the ion beam, optionally, (e) Mass-separation, optionally, (f) Collection of the isotope,wherein step (a) comprises unit operation 1,wherein step (b) comprises unit operation 2, 3, 4 and/or 5,wherein step (c) comprises unit operation 11,wherein step (d) comprises unit operation 12,wherein step (e) comprises unit operation 13, andwherein step (f) comprises unit operation 14. Unit operations 10 to 13 of the method of the present invention can also preferably be combined for the mass-separation of radioisotopes that were created and separated in any other way (e.g. commercially available radioisotopes) and hence to increase the specific activity of the resulting radioisotope preparation. Furthermore, unit operations 10 to 14 can also preferably be used to implant radioisotopes that were created and separated in any other way (e.g. commercially available radioisotopes) into nanoparticles, macromolecules, microspheres, macroaggregates, ion exchange resins or other matrices used in chromatographic systems. In this case, unit 13 is optional if the specific activity of the original radioisotope preparation has already sufficient specific activity and radioisotopic purity for the application. The so marked substrates may either be used directly for in vitro or in vivo applications (e.g. nanoparticles, microspheres, . . . for radioembolization therapy, see e.g. Wunderlich et al. Labeling and biodistribution of different particle materials for radioembolization therapy with 188Re. Appl Radiat Isot. 2005 May; 62(5):745-50.) or for subsequent chemical steps (e.g. ion exchange resins or other matrices used in chromatographic systems) or biochemical steps. Certain elements can be brought into a chemical form which is already volatile at room temperature and can thus be conveniently injected in gaseous form into an ion source. For metallic elements this method is known under the name MIVOC (metal ions from volatile compounds). E.g. iron can be introduced as ferrocene Fe(C5H5)2, zinc as dimethylzinc C2H6Zn, germanium as tetraethylgermanium Ge(C2H5)4, molybdenum as molybdenumhexacarbonyl Mo(CO)6, etc. For all these cases an oven is not absolutely necessary and unit operation 10 can be replaced by unit operation 9. The isotopes obtained by the method according to the invention are preferably 225Ac, 224Ra, 223Ra, 213Bi, 211At, 152Tb, 149Tb, 44Sc, 153Sm, 82Sr or 82Rb. The production of the following isotopes is also preferred: 28Mg, 26Al, 32Si, 32P, 33P, 42Ar, 42K, 43K, 45Ca, 47Ca, 44Sc, 44mSc, 46Sc, 47Sc, 44Ti, 52Mn, 54Mn, 56Mn, 52Fe, 55Fe, 59Fe, 55Co, 56Co, 57Co, 58Co, 62Cu, 64Cu, 67Cu, 62Zn, 68Ga, 68Ge, 72As, 72Se, 73Se, 75Se, 75Br, 76Br, 77Br, 75Kr, 76Kr, 77Kr, 81Rb, 82Rb, 82Sr, 83Sr, 85Sr, 89Sr, 85Y, 86Y, 87Y, 88Y, 89Zr, 90Nb, 97Ru, 103Pd, 103Cd, 111Ag, 113Sn, 117mSn, 119Sb, 121mTe, 121I, 122I, 123I, 124I, 125I, 126I, 130I, 121Xe, 122Xe, 123Xe, 125Xe, 127Xe, 129mXe, 131mXe, 131m,gXe, 134Ce/La, 137Ce, 139Ce, 141Ce, 143Pr, 138N/Pr, 140Nd/Pr, 147Nd, 149Pm, 142Sm/Pm, 153Sm, 155Eu, 147Gd, 148Gd, 149Gd, 149Tb, 152Tb, 155Tb, 161Tb, 157Dy, 159Dy, 166Ho, 165Er, 169Er, 165Tm, 167Tm, 169Yb, 177Yb, 172Lu, 177Lu, 172Hf, 175Hf, 178Ta, 178W, 188W, 186Re, 188Re, 192Ir, 195Au, 198Au, 194Hg, 194Hg, 197Hg, 201Tl, 202Tl, 211Pb, 212Pb, 212Bi, 213Bi, 204At, 205At, 206At, 207At, 208At, 209At, 210At, 211At, 220Rn, 221Rn, 220Fr, 221Fr, 223Ra, 224Ra, 225Ra, 225Ac, 227Ac, 227Th or 228Th. Preferably, radioisotopes in carrier-free or non-carrier added form are produced by the method of the present invention. Preferred is a method according to the invention, wherein the target that is activated by a particle beam is a metal or alloy or another high temperature compound (preferably carbide, oxide, etc). Preferred targets suitable in the present invention are Ta foil, Hg, Pb, Bi, Pb/Bi alloy, Ti, Th, U, Nb, Mo, Hf, W, ThCx, UCx ThO2 or an isotopically enriched target material, such as 152Gd, 144Sm or others. Preferably, the target is heated during or after the activation step (unit 1). In one embodiment, the target is heated above 2,000° C. However, the temperature depends on the target material and the element to be released. In one embodiment, the target is kept in a molten state, in particular elements like Hg, Pb or Bi. In other embodiments, the target is kept solid, in particular refractory elements like Nb, Mo, Hf. Ta, W or refractory compounds like ThCx, UCx. Preferably, the particles in the particle beam used to activate the target are charged or neutral particles, protons, electrons, neutrons, photons. Preferably, the particle beam has an energy in the range of a few or several ten MeV to several GeV. In few cases it is necessary to restrict the particle energy to a more narrow range to avoid production of disturbing contaminations, e.g. an alpha energy <30 MeV is preferred for the production of 211At via 209Bi(alpha,2n). Preferably, the particle beam is provided by a particle accelerator, such as cyclotron, LINAC, synchrotron. Preferably, the separation of the isotopes from the irradiated target is carried out by bringing the target to high temperature, e.g. solid targets to 60-95% of their melting point, under vacuum, e.g. in the order of 10−5 mbar or better, or suitable gas atmosphere. A preferred suitable gas atmosphere is a noble gas (He, Ne, Ar, . . . ) that is not reacting with the hot target. Occasionally reactive gases like O2, CF4, . . . are added in an amount not deleterious for the target but sufficiently high to favour the release of the wanted isotopes, e.g. at a partial pressure in the order of 10−4 mbar. Step (b) preferably comprises the transport of the isotope of interest to the surface of the target material by means of high temperature diffusion, and/or the separation of the isotope of interest from the bulk target material by high temperature desorption from the target surface under vacuum or in inert atmosphere, and/or the separation of the isotope of interest from the bulk target material by removing the target material by high temperature sublimation under vacuum or in inert atmosphere, and/or the separation of the isotope of interest from the bulk target material by adsorption on suitable substrates located in the flow of a liquid metal target and coolant medium, and/or the desorption of the isotope of interest from the bulk target material by means of chemical evaporation. Between steps (b) and (c) the isotope of interest is preferably transported by molecular flow at high temperature or by a gas flow. Between steps (b) and (c) the isotope of interest is preferably condensed or adsorbed on a surface compatible with the purity requirement of an accelerator ion source. The isotope of interest is preferably conditioned for ionisation in the ion source by adding chemicals that allow pyrochemical reduction to the elementary state, oxidation or molecule formation. The mass separation process is preferably controlled by mass marking. Before step (c) the isotope of interest is preferably introduced into an oven from where the sample is fed into the ion source. Preferably, the ionisation in step (c) is surface ionisation, laser ionisation or plasma ionisation. Elements or compounds with low ionization potential, i.e. elements of the chemical groups 1 and 3 (including many lanthanides) and heavier elements of the group 2, are most easily ionized by surface ionization. Resonant laser ionisation provides an efficient and selective ionization mode for most metallic elements. Plasma ionisation is intrinsically less selective, but compatible with practically all elements and compounds. Preferably, the mass separation step is an on-line or off-line mass separation. On-line mass separation is preferred for short-lived isotopes where a longer delay would cause unacceptable decay losses. Off-line mass separation is preferred for longer-lived isotopes where a delay is less important and in cases where technical reasons prevent a direct coupling of the production target to an on-line mass separator. Step (f) preferably comprises that the isotope of interest is collected by implantation into a prepared chemical substrate. Preferably, a further purification step follows the collection of the isotope in step (f). Preferably, all steps (a) to (f) are repeated or the irradiated target material of step (a) is reused. The steps can be repeated, one time, two times, three times or as often as necessary to obtain the required purity. The radioisotopes produced by the method of the present invention are preferably used for producing radioisotope-labelled bioconjugates or radioisotope-labelled nanoparticles, microspheres or macroaggregates. Preferred bioconjugates are immuno-conjugates, antibodies, antibody fragments, such Fv, Fab, scFv, heavy and light chains, chimeric antibodies or antibody fragments, humanized antibodies or antibody fragments proteins, peptides, nucleic acids, such as RNA, DNA and modifications thereof, such as PNA, and oligonucleotides or fragments of any of them. Bioconjugates are any wildtype or recombinant protein (such as monoclonal antibodies, their fragments, human serum albumin (HSA)) as well as microspheres or macro-aggregates made from said proteins, peptides and/or oligonucleotides. Bioconjugates further comprise nanoparticles, microspheres or macroaggregates that are conjugated with or covalently or noncovalently attached to said immuno-conjugates, antibodies, proteins, peptides, nucleic acids, oligonucleotides or fragments thereof. Bioconjugates can carry linker molecules or tags for molecular recognition, purification and/or handling purposes, such as avidin, streptavidin, biotin, protein A or G, fluorophores, dyes, chromophores. However, such linker molecules and tags are well known to the person of skill in the art. Preferably, the bioconjugates further comprise chelating groups, such as derivatives of DTPA or DOTA, with or without linking molecules for the labelling with the isotopes. The radioisotope-labelled bioconjugates can preferably used for diagnostic procedures or therapeutic protocols, such as SPECT, quantitative PET imaging for individual in vivo dosimetry, RIT, TAT, Auger-therapy or radioembolization. The radioisotopes produced by the method of the present invention, preferably 204At, 205At, 206At, 207At, 708At, 209At or 210At, can be used for in vitro or in vivo biodistribution studies or dosimetry via PET, gamma-spectrometry or SPECT. The mass-separated ion-beam is preferably implanted into an implantation substrate (unit operation 14). The implantation energy is preferably selected in order to adjust the implantation depth. By selecting the implantation energy, the implantation depth can be adjusted that alpha-recoils can either be ejected and emanate (implantation energy typically <100 keV leads to a low implantation depth), thus representing an open source, or that alpha-recoils cannot leave the matrix (implantation energy typically >150 keV leads to a deeper implantation depth), hence representing a closed source. The implantation is preferably performed through a thin cover layer into the implantation substrate. Thus the source can be transported as “closed”. The end user can easily remove the cover layer by dissolving, evaporating, burning, mechanically removing, etc. to obtain an open source with well-defined depth profile. The implantation substrate is preferably a salt layer, a water-soluble substance, such as sugars, a thin ice layer of frozen water or another liquid or a solid matrix, such as a metal foil. The separation from the salt layer containing the radioisotopes preferably comprises subsequent dissolving in a small volume of water or the eluting agent, and/or as such direct injection into the chromatographic system. The separation from the thin ice layer containing the radioisotopes preferably comprises subsequent melting by heating, with any suitable method (Ohmic heating, infrared heating, radio-frequency heating, . . . ). The separation from the solid matrix, such as a metal foil, preferably requires additional chemical separation from the matrix material. Instead of a soluble matrix, the ion beam can also be implanted into any other solid matrix, e.g. a metal foil. In this case one needs additionally a chemical separation of the desired isotope from the matrix material that usually disturbs the chromatographic process. It is furthermore preferred that conventional radio-chemical and radio-chromatographical processes are performed, such as precipitation, electrochemical separations, extraction, cation exchange chromatography, anion exchange chromatography, extraction chromatography, thermo chromatography, gas chromatography. The separation from the implantation substrate preferably comprises thermal release from a refractory matrix. A particularly simple and efficient separation from the implantation substrate can be achieved by thermal release from a refractory matrix. The present invention further provides a method for direct radioisotope-labelling of bioconjugates, comprising (i) performing the method for the production of high-purity isotopes according to the invention as described above, (ii) obtaining the product fraction containing the radioisotope of interest in a small volume, and (ii) direct radioisotope-labelling of bioconjugates and/or direct injection into a chromatographic system for further purification, wherein the bioconjugates are as defined above. The bioconjugates further preferably comprise nanoparticles, microspheres or macroaggregates that are conjugated with or covalently or noncovalently attached to said immuno-conjugates, antibodies, proteins, peptides, nucleic acids, oligonucleotides or fragments thereof. The radioisotope-labelled bioconjugates obtained by the bioconjugate-labelling method (see above) are preferably used in radio-immunotherapy (RIT) of diseases, such as cancer. Said radioisotope-labelled bioconjugates are preferably used for diagnostic procedures, such as SPECT, quantitative PET imaging for individual in vivo dosimetry, or for therapeutic protocols, such as RIT, TAT or Auger-therapy. Further preferred implantation substrates are nanoparticles, macromolecules, microspheres, macroaggregates, ion exchange resins or other matrices used in chromatographic systems. The present invention further provides a method for direct labelling of nanoparticles, macro-molecules, micro-spheres, macro-aggregates, ion exchange resins or other matrices used in chromatographic systems, comprising (i) performing the method for the production of high-purity isotopes according to the invention as described above, (ii) direct implanting of the radioactive ion beam into said nanoparticles, macro-molecules, micro-spheres, macro-aggregates, ion exchange resins or other matrices used in chromatographic systems. Preferably, step (ii) of the above method is carried out on-line. Alternatively, after the standard purification steps of step (i) the product is again injected into an ion source, ionized, accelerated and then step (ii) is performed. Furthermore, unit operations 10 to 14 can also preferably be used to implant radioisotopes that were created and separated in any other way (e.g. commercially available radioisotopes) into nanoparticles, macromolecules, microspheres, macroaggregates, ion exchange resins or other matrices used in chromatographic systems. In this case, unit 13 is optional if the specific activity of the original radioisotope preparation has already sufficient specific activity and radioisotopic purity for the application. The so marked substrates may either be used directly for in vitro or in vivo applications (e.g. nanoparticles, microspheres, . . . for radioembolization therapy, see e.g. Wunderlich et al. Labeling and biodistribution of different particle materials for radioembolization therapy with 188Re. Appl Radiat Isot. 2005 May; 62(5):745-50.) or for subsequent chemical steps (e.g. ion exchange resins or other matrices used in chromatographic systems) or biochemical steps. Therefore, the present invention further provides a method for direct labelling of nanoparticles, macro-molecules, micro-spheres, macro-aggregates, ion exchange resins or other matrices used in chromatographic systems, comprising the following steps: (a) Obtaining a sample of an isotope, such as a commercially available isotope, (b) Introduction of said isotope into an oven from where said sample is fed into an ion source, (c) Ionisation of said isotope, (d) Extraction from the ion source and acceleration of the ion beam, (e) optionally, Mass-separation, (f) Collection of the isotope by direct implanting of the radioactive ion beam into said nanoparticles, macro-molecules, micro-spheres, macro-aggregates, ion exchange resins or other matrices used in chromatographic systems.wherein step (b) comprises unit operation 10,wherein step (c) comprises unit operation 11,wherein step (d) comprises unit operation 12,wherein step (e) comprises unit operation 13, andwherein step (f) comprises unit operation 14. The invention further provides a device for performing the method for the production of high-purity isotopes according to the invention, as described above. The invention further provides the use of said device as a dry-isotope generator, in particular dry 62Zn/62Cu, 228Th/224Ra, 224Ra/212Pb/212Bi, 228Th/212Pb/212Bi, 225Ac/213Bi, 227Ac/227Th/223Ra, 44Ti/44Sc generator. The invention further provides a device for performing the method for direct radioisotope-labelling of bioconjugates, as described above. The invention further provides a device for performing the method for direct labelling of nanoparticles, macro-molecules, micro-spheres, macro-aggregates, ion exchange resins or other matrices used in chromatographic systems, as described above. The present invention also provides a method for the large scale production of high-purity carrier-free or non carrier added radioisotopes comprising the following steps: (a) Activation of a target by a particle beam, (b) Separation of the isotope from the irradiated target, (c) Ionisation of the separated isotope, (d) Extraction from the ion source and acceleration of the ion beam, (e) Mass-separation, (f) Collection of the isotope, wherein the isotopes are produced by on- or off-line extraction of radioisotopes from a high power liquid metal target. The present invention also provides a method for the large scale production of high-purity carrier-free or non carrier added radioisotopes comprising the following steps: (a) Activation of a target by a particle beam, (b) Separation of the isotope from the irradiated target, (c) Ionisation of the separated isotope, (d) Extraction from the ion source and acceleration of the ion beam, (e) Mass-separation, (f) Collection of the isotope, wherein the isotope are produced via continuous or batch-mode extraction from targets. The present invention also provides a method for the large scale production of high-purity carrier-free radioisotope 211At comprising the following steps: (a) Activation of a target by a particle beam, (b) Separation of the isotope from the irradiated target, (c) Ionisation of the separated isotope, (d) Extraction from the ion source and acceleration of the ion beam, (e) Mass-separation, (f) Collection of the isotope, wherein the isotope is produced on-line, and wherein the produced isotope is the carrier-free radioisotope 211At. The present invention also provides a method for the large scale production of high-purity carrier-free radioisotopes comprising the following steps: (a) Activation of a target by a particle beam, (b) Separation of the isotope from the irradiated target, (c) Ionisation of the separated isotope, (d) Extraction from the ion source and acceleration of the ion beam, (e) Mass-separation, (f) Collection of the isotope, wherein the isotopes are produced on-line, and wherein the produced isotopes are the carrier-free radioisotopes 204-210At. The present invention also provides a method for the large scale production of high-purity carrier-free radioisotopes of the rare earth elements comprising the following steps: (a) Activation of a target by a particle beam, (b) Separation of the isotope from the irradiated target, (c) Ionisation of the separated isotope, (d) Extraction from the ion source and acceleration of the ion beam, (e) Mass-separation, (f) Collection of the isotope, wherein the produced isotopes are carrier free radioisotopes of the rare earth elements The present invention also provides a method for the large scale production of high-purity carrier-free or non carrier added neutron-rich lanthanide and tin isotopes comprising the following steps: (a) Activation of a fission target by a particle beam, (b) Separation of the isotope from the irradiated target, and optionally, (c) Ionisation of the separated isotope, optionally, (d) Extraction from the ion source and acceleration of the ion beam, optionally, (e) Mass-separation, optionally, (f) Collection of the isotope, wherein the neutron-rich lanthanide and tin isotopes are produced by fission. Unit operations 10 to 13 of the method of the present invention can also preferably be combined for the mass-separation of radioisotopes that were created and separated in any other way (e.g. commercially available radioisotopes) and hence to increase the specific activity of the resulting radioisotope preparation. A preferred method for such mass-separation of isotopes comprises the following steps: (a) Obtaining a sample of an isotope, such as a commercially available isotope, (b) Introduction of said isotope into an oven from where said sample is fed into an ion source, (c) Ionisation of said isotope, (d) Extraction from the ion source and acceleration of the ion beam, (e) Mass-separation.wherein step (b) comprises unit operation 10,wherein step (c) comprises unit operation 11,wherein step (d) comprises unit operation 12, andwherein step (e) comprises unit operation 13. Other preferable aspects of the invention will become apparent from the detailed description of preferred embodiments and aspects thereof. The features of the present invention disclosed in the specification, the preferred embodiments and aspects, the examples, the claims and/or in the accompanying figures, may, both separately, and in any combination thereof, be material for realizing the invention in various forms thereof. For the purposes of the present invention, all references as cited herein are incorporated by reference in their entireties. Definitions: The following terms and abbreviations are used throughout the description and examples: First of all, the terms “radioisotopes” and “radionuclides” are used interchangeably throughout the description. “Spallation” means a nuclear reaction occurring for incident particle energies >100 MeV. The method of the present invention preferably uses high energy particles (>100 MeV). Because when beams with lower energy are used reduced production cross-sections and also some production of products close-by to the target nuclides can occur. However, the method of the present invention also uses high energy particles with an energy lower than 100 MeV, such as 80 or 90 MeV. The energy limits used throughout the description, embodiments, aspects, examples and claims of the present invention are not to be considered as sharp but rather indicative, allowing an application of lower-energy beams during the separation (such as unit operations 2-14) and during production (such as unit operation 1). A preparation of a given radioisotope is “carrier free”, when it is free from other isotopes (both stable and radioactive) of the element in question. However, the term “carrier free” also comprises preparations, where the wanted radioisotope is absolutely dominating the total activity and radiotoxicity over radioisotopes of the same element and where stable isobars of the same element that would cause significant differences in the application to that of a pure radioisotope are not be present. A preparation of a given radioisotope is “non carrier added”, when special attention has been paid to procedures, equipment and material in order to minimize the introduction of other isotopes (both stable and radioactive) of the element in question in the same chemical form or as a species enabling isotopic exchange reactions. In the method of the present invention no stable or radioactive isotopes of the same element are added on purpose, though some amount may be intrinsically present due to the production process. The “target” is that part of a radioisotope production system which is exposed to the beam inducing nuclear reactions in it. The target “matrix” is more specifically the inner part of the target where the wanted nuclear reactions occur. The target “matrix” does not contain the surrounding target container, etc. “Effusion” defines diffusion in open space (e.g. under vacuum). Similar to the diffusion in solids or liquids “effusion” is a random walk process described by similar mathematical concepts. Effusing isotopes are those, which have already left the target matrix, i.e. have already desorbed. “Release” requires the diffusion to the surface of the matrix plus desorption. In case of an “on-line” mode, the part of a device performing the separation (such as unit operations 2-14) of the method of the invention is directly connected to the part of the device performing the production (such as unit operation 1) and operates simultaneously to the production. Whereas in case of an “off-line” mode, the separation starts after a stop of the production or batch-wise by removing target material from the irradiation region before separation. “ADS” (Accelerator Driven Systems) are subcritical nuclear reactors where the neutrons necessary to maintain a continuous chain reaction are supplied by an (accelerator driven) spallation neutron source (or by breakup of deuteron beams). “MEGAPIE” is a demonstrator experiment for a megawatt liquid metal target at the Paul Scherrer Institute. In the “ISOL” (Isotope Separation On-Line) method thick targets are bombarded with a primary beam to produce nuclear reaction products. The latter are first stopped in the target matrix, then diffuse out of it, desorb from its surface, get to an ion source where they are ionised, extracted, slightly accelerated and mass-separated. “RIT” (Radio-Immuno Therapy) is an immunotherapy where the agents (monoclonal antibodies, etc.) are conjugated with radioisotopes. The decay of the latter destroys or harms preferentially the environment, i.e. the cancer cells or any other illness related unit in the body. “TAT” (Targeted Alpha Therapy) is a RIT using alpha emitting radioisotopes. “PET” (Positron Emission Tomography): A radioactive tracer isotope which decays by emitting a positron, chemically incorporated into a molecule, is injected into the living subject (usually into blood circulation). There is a waiting period while the molecule becomes concentrated in tissues of interest, then the subject is placed in the imaging scanner. The isotope decays, emitting a positron. After traveling up to a few millimeters the positron annihilates with an electron, producing a pair of annihilation photons (511 keV) moving in opposite directions. These are detected when they reach a scintillator material in the scanning device, creating a burst of light which is detected by photomultiplier tubes. The technique depends on coincident detection of the pair of photons; photons which do not arrive in pairs (i.e., within a few nanoseconds) are ignored. By measuring where the annihilation photons end up, their origin in the body can be plotted, allowing the chemical uptake or activity of certain parts of the body to be determined. The scanner uses the pair-detection events to map the density of the isotope in the body, in the form of slice images separated by some millimeters. The resulting map shows the tissues in which the molecular probe has become concentrated, and is read by a nuclear medicine physician or radiologist, to interpret the result in terms of the patient's diagnosis and treatment. “SPECT” (Single Photon Emission Computed Tomography) is a nuclear medicine tomographic imaging technique using gamma rays. The technique results in a set of image slices through a patient, showing the distribution of a radiopharmaceutical. Firstly a patient is injected with a gamma-emitting radiopharmaceutical. Then a series of projection images are acquired using a gamma camera. The acquisition involves the gamma camera rotating around the patient acquiring images at various positions. The number of images and the rotation angle covered varies depending on the type of investigation required. The preferred embodiments and preferred aspects as well as the examples of the present invention shall now be further described with reference to the accompanying figures without being limited thereto. The following embodiments utilize the previously defined unit operations of the method of the present invention, i.e. preferred combinations, selections, sequences and/or optimizations thereof. However, the person of skill in the art will be able to define und utilize other suitable combinations, selections, sequences and/or optimizations of these unit operations depending on the desired radioisotope(s) to be produced. For better understanding, the respective utilized unit operations are marked in brackets, e.g. {unit 1}. 1. Application: High power liquid metal targets are presently being built, planned or proposed for a series of facilities: spallation neutron sources, ADS (accelerator driven systems), as neutron converter for high power ISOL facilities, as meson production target for “superbeams”, neutrino factories or muon collider. As a by-product, in the liquid metal target large amounts of radioisotopes are produced by spallation, fragmentation and high energy fission. Generally this radioactivity production is rather considered as a problem since the buildup to a high radioactivity inventory poses tight constraints on the safety of the facility. The inventors provide here a series of methods to continuously extract a good fraction of the produced activity. This serves two purposes: a reduction of the radioactive inventory in the hot target area and the liquid metal loop as a safety measure, and an exploitation of the retrieved radioisotopes for life sciences. FIGS. 1A and 1B illustrate the different ways to extract radioisotopes from a liquid metal target, either continuously (FIG. 1A) or batch-wise (FIG. 1B). The detailed steps are discussed in the following. 2. Method: A molten metal target (Hg, Pb, Bi or alloys containing at least one of these elements) is irradiated with high energy particles of >100 MeV energy {unit 1}. With an intermediate energy of few 100 MeV mainly close spallation products (evaporation of 10-30 nucleons) as well as little fission and fragmentation products are generated. With increasing energy of the incident beam (around 1 GeV and above) also deep spallation products (evaporation of 30-60 nucleons) and more fission and fragmentation products are generated. Hence nearly all radioisotopes ranging from 3H up to two elements beyond the target element are generated and can be extracted. Depending on the chemical nature of the elements to be extracted, different variants have to be applied for the extraction: A. Noble Gases Noble gases will diffuse to the surface of the liquid target material {unit 2} and be released from it into the target enclosure. The effusing {unit 3} radioisotopes can then be transported by vacuum diffusion or by a flow of inert gas (He, Ar, . . . ) {unit 7} to a plasma ion source where they are ionized {unit 11}. The ions are extracted from the ion source, accelerated to typically several tens of keV {unit 12} and separated in a magnetic sector field according to the mass/charge ratio {unit 13}. The ions are implanted into e.g. a metallic catcher {unit 14}. Alternatively the ions are directly implanted into nanoparticles, etc. {unit 14} for labelling of the latter. For purification, a cold trap can be placed between the target and ion source to retain elements and molecules, which are less volatile than the noble gases of interest. Thus, method A utilizes a combination of units 1, 2, 3, 7, 11, 12, 13 and 14. B. Halogens, Mercury, Thallium {units 1 and 2 as in A} The halogens and mercury are relatively volatile and are released {unit 3} at the typical operation temperature of targets made from Pb, Bi or alloys containing these elements, e.g. Pb/Bi (this method is not applicable for Hg targets which are operated at lower temperatures). At an enhanced temperature (>600° C.) also thallium is released. These elements will adsorb easily on the walls of the target enclosure if the latter are kept at room temperature. The inventors provide therefore to heat the walls of the target enclosure, and insert a dedicated catcher, which is held at lower temperature {unit 8}. In case of combination with the on-line extraction of noble gases for mass separation, the cold trap will act as catcher for halogens, Hg and Tl. Thus, method B utilizes a combination of units 1, 2, 3 and 8. Variant with on-line mass separation: The effusing {unit 3} radioisotopes of halogens, Hg and optionally Tl can be transported {unit 7} together with the noble gases by vacuum diffusion or by a flow of inert gas (He, Ar, . . . ) to an ion source where they are ionized {unit 11}. The ions are extracted from the ion source, accelerated to typically several tens of keV {unit 12} and separated in a magnetic sector field according to the mass/charge ratio {unit 13}. The ions are implanted into e.g. a metallic catcher {unit 14}. Alternatively the ions are directly implanted into nanoparticles, etc. {unit 14} for labelling of the latter. Thus, this variant of method B utilizes a combination of units 1, 2, 3, 7, 11, 12, 13 and 14. C. Elements with Lower Volatility than the Target Elements {units 1 and 2 as in A} Part of the liquid target material is removed from the area where the beam interacts with it. If the target material is circulated, this can be done e.g. continuously via a side loop. This will help to harvest some radioisotopes, but will not contribute much to a reduction of the overall radioactive inventory in the target area. Therefore, the system is instead made to operate in a push-pull-mode between two batches of liquid target material. While the second batch has come in operation, the first one is available for extraction of the interesting nuclei or for general reduction of its inventory. The recovery of the wanted species can be done in one of the following non-destructive ways that leave the Hg intact and ready for immediate reuse: 1) Dry Distillation {Unit 4} The target material is removed by evaporation under vacuum or inert atmosphere leaving the less volatile elements in the residue. The wanted nuclei can be recovered from the residue with a variety of methods depending on the element. 2) Liquid-Liquid Extraction Liquid Hg can be mixed with a suitable solvent, e.g. citric acid. Shaking the mixture for a certain time, e.g. half an hour, allows to transfer a good fraction of the radiolanthanides (valence 3 elements) to the solvent. The solvent is easily separated from the mercury, which will due to its high density and surface tension rapidly coagulate at the bottom of the recipient. 3) Harvesting by Selective Adsorption {Unit 5} The liquid target metal can be brought in contact with a surface which strongly adsorbs the lanthanides and transition metals that are known to have the lowest solubility, at least in Hg. This can be stable impurities added or dissolved from the steel plumbing like Ni, Mn and Cr that segregate out as oxides floating on the surface of Hg. They act as scavengers for the radioisotopes of the other transition metals and the rare earths so that they can be recovered by simple wiping them of the Hg surface. In all cases the solvent or residue containing the radioisotopes is either used as stock solution for any conventional radiochemical separation method or evaporated to dryness {unit 8} and inserted into an oven {unit 10} connected to an ion source (surface, laser or plasma ionization). The oven is heated to allow the radioisotopes effuse to the ion source {unit 11} where they are ionized. The ions are extracted from the ion source, accelerated to typically several tens of keV {unit 12} and separated in a magnetic sector field according to the mass/charge ratio {unit 13}. The ions are implanted into e.g. a suitable catcher {unit 14} that facilitates the labeling of the radiopharmaceutical. Alternatively the ions are directly implanted into nanoparticles, etc. {unit 14} for labelling of the latter. Thus, method C utilizes a combination of units 1, 2, (4 or 5), 8, 10, 11, 12, 13 and 14. Particular Advantages Include: The inventors describe for the first time the details of implementation of an extraction plant for radioactive isotopes from irradiated liquid metal targets. The inventors provide for each class of elements the preferred method of extraction. The inventors have performed a demonstration for the on-line production of mass-separated noble gas beams from a Pb target as well as from a Pb/Bi target irradiated with 1.4 GeV protons (“proof-of-principle”). The inventors have performed a demonstration of the on-line production of mass-separated mercury isotopes from a Pb/Bi target irradiated with 1.4 GeV protons (“proof-of-principle”). Particularly strong undissociable bonds to nanoparticles can be obtained by the ion-implantation labelling. The continuous, automated production without manual operation steps ideally suited for industrial production is demonstrated. The inventors provide this method to keep the radioactive inventory in the target area and the liquid metal loop small, an important factor in the safety of high power facilities. The inventors provide a new, simple way to obtain a 62Zn/62Cu generator. In summary, the methods provided within this embodiment comprise the following features: This universal method works basically for all radionuclides between 3H and two elements beyond the target element. In particular the following radionuclides have dedicated relevance: 28Mg, 26Al, 32Si, 32P, 33P, 42Ar, 42K, 43K, 45Ca, 47Ca, 44Sc, 44mSc, 46Sc, 47Sc, 44Ti, 52Mn, 54Mn, 56Mn, 52Fe, 55Fe, 59Fe, 55Co, 56Co, 62Cu, 64Cu, 67Cu, 62Zn, 68Ga, 68Ge, 72As, 72Se, 73Se, 75Se, 75Br, 76Br, 77Br, 75Kr, 76Kr, 77Kr, 81Rb, 82Rb, 82Sr, 83Sr, 85Sr, 89Sr, 85Y, 86Y, 87Y, 88Y, 89Zr, 90Nb, 97Ru, 103Pd, 103Cd, 111Ag, 113Sn, 117mSn, 119Sb, 121mTe, 121I, 122I, 123I, 124I, 125I, 126I, 130I, 121Xe, 122Xe, 123Xe, 125Xe, 127Xe, 129mXe, 131mXe, 131m,gXe, 134Ce/La, 137Ce, 139Ce, 141Ce, 143Pr, 138N/Pr, 140Nd/Pr, 147Nd, 149Pm, 142Sm/Pm, 153Sm, 155Eu, 147Gd, 148Gd, 149Gd, 149Tb, 152Tb, 155Tb, 161Tb, 157Dy, 159Dy, 166Ho, 165Er, 169Er, 165Tm, 167Tm, 169Yb, 177Yb, 172Lu, 177Lu, 172Hf, 175Hf, 178Ta, 178W, 188W, 186Re, 188Re, 192Ir, 195Au, 198Au, 194Hg, 194Hg, 197Hg, 201Tl, 202Tl, and 202Tl. A liquid metal target made from pure Hg, Pb, Bi or an alloy containing at least one of these elements is used. For producing the elements Ba and lighter additionally to the target materials mentioned above a liquid target made from pure lanthanides or an alloy containing at least one lanthanide element can be used. For producing the elements Sb and lighter additionally to the target materials mentioned above a liquid target made from pure tins or an alloy containing tin can be used. For producing the elements As and lighter additionally to the target materials mentioned above a liquid target made from pure germanium or an alloy containing germanium can be used. Of particular interest here is the possibility to separate a pure 62Zn beam, which can be implanted into a suitable matrix and serve as generator for the daughter isotope 62Cu. Since there are no other long-lived zinc isotopes decaying to radioactive copper isotopes, such a generator can even be produced without ionization and mass separation, by just catching the zinc fraction released from a liquid germanium target kept at a suitable temperature (>1000° C.). After few hours most of the short-lived zinc isotopes (mainly 63Zn) have decayed and from now for the next 1-2 days a 62Cu/65Cu mixture with >10% 62Cu content is obtained by extracting repeatedly the Cu fraction by conventional radiochemical separation methods. During irradiation the target is kept above the melting point. The temperature is controlled by heating/cooling the target vessel and/or heating/cooling the target material when the latter is flowing in a circuit. The liquid target material can be standing as a bath in a container, be a free-standing jet or a flow enclosed on one or more sides by a wall. The incident beam with >100 MeV energy is provided by a particle accelerator (cyclotron, LINAC, synchrotron, etc.). The incident proton beam can be replaced by energetic light ions (d, 3He, 4He, . . . ), heavy ions, neutrons, electrons or photons. The proton beam can enter the target enclosure via a window or via a differentially pumped section. The target material can be kept in motion by pumping, mechanical shaking, electromagnetic agitation, etc. to assure a better temperature homogeneity and thus allow for higher beam currents without the risk of local overheating. A chimney or baffles can be used to condense evaporating target material before it reaches the catcher or ion source. The radioisotopes will diffuse to the surface of the liquid target material. Radioisotopes of elements with higher volatility than the target material can be released from the target surface into the target enclosure. The effusing radioisotopes can then be transported by vacuum diffusion or by a flow of inert gas (He, Ar, . . . ) to an ion source where they are ionized. The target is connected to the ion source in a way that no other escape path is available for the radioisotopes. Optionally the flow of effusing volatile radioisotopes can be directed towards the ion source with a turbomolecular pump. The entire target enclosure and all surfaces which the released radioisotopes can encounter, except the catcher, is kept at a sufficiently high temperature to avoid a condensation of halogens, mercury and thallium at places other than the catcher. The ions are extracted from the ion source, accelerated to typically several tens of keV and separated in a magnetic sector field according to the mass/charge ratio. The ions are implanted into e.g. a metallic catcher. Alternatively the ions are directly implanted into nanoparticles, etc. for labelling of the latter. For purification, a cold trap can be placed between the target and ion source to retain elements and molecules, which are less volatile than the noble gases of interest. Radioisotopes can be extracted on-line without disturbing the target irradiation if part of the liquid target material is removed from the area where the beam interacts with it. If the target material is circulated, this can be done e.g. continuously via a side loop. To reduce the overall radioactive inventory in the target area, the system can be made to operate in a push-pull-mode between two batches of liquid target material. While the second batch has come in operation, the first one is available for extraction of the interesting nuclei or for general reduction of its inventory. The recovery of the wanted species can be done in one of the following non-destructive ways that leave the Hg intact and ready for immediate reuse: A) Dry distillation: The target material is removed by evaporation under vacuum or inert atmosphere leaving the less volatile elements in the residue. The wanted nuclei can be recovered from the residue with a variety of methods depending on the element. B) Liquid-liquid extraction: Liquid Hg can be mixed with a suitable solvent, e.g. citric acid. Shaking the mixture for a certain time, e.g. half an hour, allows to transfer a good fraction of the radiolanthanides (valence 3 elements) to the solvent. The solvent is easily separated from the mercury, which will due to its high density and surface tension rapidly coagulate at the bottom of the recipient. C) Harvesting by selective adsorption: The liquid target metal can be brought in contact with a surface which strongly adsorbs the lanthanides and transition metals that are known to have the lowest solubility, at least in Hg. This can be stable impurities added or dissolved from the steel plumbing like Ni, Mn and Cr that segregate out as oxides floating on the surface of Hg. They act as scavengers for the radioisotopes of the other transition metals and the rare earths so that they can be recovered by simple wiping them of the Hg surface. In all cases (A, B or C) the solvent or residue containing the radioisotopes is either used as stock solution for any conventional radiochemical separation method or evaporated to dryness and inserted into an oven connected to an ion source (surface, laser or plasma ionization). The oven is heated. The effusing radioisotopes can then be transported by vacuum diffusion or by a flow of inert gas (He, Ar, . . . ) to an ion source where they are ionized. The oven is connected to the ion source in a way that no other escape path is available for the radioisotopes. The inert gas can be replaced by any other gas if the latter is compatible with the integrity of the target, the enclosure and the catcher surface. Several chambers with catchers can be attached to the target chamber and connected/disconnected from the latter without interruption of the irradiation for a significant time. Variant: instead of on-line separation, the irradiation can be performed at a reduced target temperature. The target is then heated afterwards when needed to release the elements of interest. The target, oven, walls, ion source, etc. are heated by any suitable mean (Ohmic heating, electron bombardment, radio-frequency, infrared heating, laser heating, energy loss of the incident beam, etc.) or any combination of these methods. The effusing radioisotopes can be transported by a flow of inert gas (He, Ar, . . . ) to the ion source instead of being transported by vacuum diffusion. The mass separation can be performed with any mass-selective device, e.g. a Wien-filter, a radio-frequency quadrupole, etc. instead of the magnetic sector field. Often several isotopes of the same element, or isobars with comparable masses are produced in the same system. In this case a mass-selective device is of advantage, which allows to collect simultaneously several masses. The mass-separated ion beam is implanted into a salt layer. The salt layer containing the radioisotopes is subsequently dissolved in a small volume of water or the eluting agent. The salt cover of the backings can be replaced by many other water-soluble substances (sugar, . . . ) or by a thin ice layer (frozen water or other liquid). Instead of dissolving, the latter is subsequently melted by heating with any suitable method (Ohmic heating, infrared heating, radio-frequency heating, . . . ). Instead of a soluble matrix, the ion beam can also be implanted into any other solid matrix, e.g. a metal foil. In this case one needs additionally a chemical separation of the desired isotope from the matrix material that usually disturbs the chromatographic process. In all cases (i.e. elution from the catcher, dissolving of salt, etc. layer, melting of ice layer) the product fraction is usually obtained in a small volume and can be directly used for the labelling procedure of bio-conjugates or be directly injected into a chromatographic system for further purification. A particularly simple separation that allows to obtain many of the described elements in gaseous form can be achieved by thermal release from a refractory matrix. Any of the classical radio-chemical and radio-chromatographical processes (precipitation, electrochemical separations, extraction, cation exchange chromatography, anion exchange chromatography, extraction chromatography, thermo chromatography, gas chromatography, etc.) suitable for the separation of astatine can be applied for the separation of the desired product from isobars and pseudo-isobars (stemming from molecular sidebands like oxides or fluorides appearing at the same mass settings), from daughter products generated by the radioactive decay of the collected radioisotopes during collection and processing and from other impurities. Ligands used for the chemical separation process are eventually remaining with the product fraction and need to be eliminated before further labelling procedures. Evaporation is the most suitable way for many cases. Nano- or micro-particles, macro-molecules, micro-spheres, macro-aggregates, ion exchange resins or other matrices used in chromatographic systems can be labelled directly by implanting the radioactive ion beam into them. For cases where the radioisotopic purity is already sufficient or for implantation into ion exchange resins or other matrices used in chromatographic systems, this can be done directly on-line. Else, after the standard purification steps (radio-chromatographic separation of isobars) the product is again injected into an ion source, ionized, accelerated and implanted. The so obtained products are carrier-free and isotopically pure. The process can be operated with all the technological steps of the chain as described. However, one can reduce freely the number of steps in many cases to adapt to the required purity of the respective application. The inventors provide the separation of the noble gas isotopes 75,76,77Kr as a new production method of their respective decay daughters 75,76,77Br. The inventors provide the separation of the noble gas isotopes 121,122,123,125Xe as a new production method of their respective decay daughters 121,122,123,125I. 1. Application: The alpha emitters 212Bi, 213Bi, 223Ra, 224Ra and 225Ac and the in vivo generator isotope Pb are promising candidates for targeted alpha therapy. 2. Method: The inventors provide the following new methods: A. Spallation production of 225Ac A target made from metallic 232Th or a compound or alloy containing 232Th is irradiated by high energy (>50 MeV) particles {unit 1}. Alternatively a target made from natural uranium or 238U partially or fully depleted in 235U or a compound or alloy containing these isotopes is irradiated by high energy (>80 MeV) particles {unit 1}. 225Ac is produced by the spallation reaction 232Th(p,2p6n) or 238U(p,4p10n) respectively. After a suitable cooling period to let short-lived isotopes decay, Ac is separated from the target and the mixture of spallation and fission products by a conventional radiochemical separation method. The resulting Ac fraction contains a mixture of 225Ac and 227Ac with an activity ratio of the order of 100 to 1000 in favor of 225Ac. Optionally, the isotopic purity of 225Ac can be further enhanced by evaporating the Ac fraction to dryness {unit 8} and inserting it into an oven {unit 10} connected to an ion source (surface, laser or plasma ionization) {unit 11}. The oven is heated to allow the radioisotopes effuse {unit 7} to the ion source where they are ionized. The ions are extracted from the ion source, accelerated to typically several tens of keV {unit 12} and separated in a magnetic sector field according to the mass/charge ratio {unit 13}. The ions are implanted into e.g. a suitable catcher {unit 14} that facilitates the labeling of the radiopharmaceutical or directly into a column of a 225Ac/213Bi generator {unit 14} Alternatively the ions are directly implanted into nanoparticles, etc. {unit 14} for labelling of the latter. 227Ac can be collected simultaneously and serve as generator for 223Ra production. Thus, method A utilizes a combination of units 1, 7, 8, 10, 11, 12, 13 and 14. B. Spallation Production and Dry, Non-Target-Destructive Extraction of 225Ac Method A has still the drawback that the target is destroyed during the Ac extraction process and that liquid chemical waste is produced. The following variant omits these problems: A target made from metallic 232Th or a compound or alloy containing 232Th is irradiated by high energy (>50 MeV) particles {unit 1}. The Th foils/fibers/spheres/foam/etc. can be mixed with spacers made from a refractory metal (Ta, W, Re, Ir, . . . ) which maintain the geometric arrangement during heating. The target is heated to sufficiently high temperature (80-100% of the melting temperature) to make Ac diffuse {unit 2} to the surface from where it can desorb {unit 3}. As in method 1, chemical and mass separations {units 10-14} can be used to achieve the desired isotopic purity. The target can be used continuously over longer time or batch-wise (irradiating/extracting/irradiating/ . . . ) for several times. Thus, method B utilizes a combination of units 1, 2, 3, 10, 11, 12, 13 and 14. C. Production of Isotopically Pure 223,224,225Ra Samples A target made from metallic 232Th or a compound or alloy containing 232Th is irradiated by high energy (>50 MeV) particles {unit 1}. Alternatively, a target made from natural uranium or 238U partially or fully depleted in 235U or a compound or alloy containing these isotopes is irradiated by high energy (>80 MeV) particles {unit 1}. The isotopes 223,224,225Ra are produced by the spallation reaction 232Th(p,3p5-7n) or 238U(p,5p9-11n) respectively. The target is heated to sufficiently high temperature (70-100% of the melting temperature) to make Ra diffuse to the surface {unit 2} from where it can desorb {unit 3}. Ra desorption is favored {unit 6} by addition of halogens or a volatile halogenated compound. The Ra isotopes are escaping from the target material and transported in vacuum or under gas flow {unit 7} to the ion source {unit 11}, where they are ionised into single positively charged ions using any kind of ionisation principles (surface ionisation, resonant laser ionisation or plasma ionisation). The ions are extracted from the ion source, accelerated to typically several tens of keV {unit 12} and separated in a magnetic sector field according to the mass/charge ratios into isobars {13}. The ions are implanted into e.g. a suitable catcher {unit 14} that facilitates the labeling of the radiopharmaceutical or directly into a column {unit 14} of a 224Ra/212Bi generator or 225Ra/213Bi generator respectively. Alternatively the ions are directly implanted into nanoparticles, etc. {unit 14} for labelling of the latter. After mass separation 223Ra, 224Ra and 225Ra can be collected simultaneously for different applications. Thus, method C utilizes a combination of units 1, 2, 3, 6, 7, 11, 12, 13 and 14. D. Indirect On-Line Production of Pure 212,213Bi Samples In a variant of method C the target and/or ion source is kept at a lower temperature {units 1-3, 7, 11-13 as before}. Thus very pure beams of francium isotopes can be produced. The mass-separated 220Fr beam is collected {unit 14} and decays to a pure 212Bi sample. Simultaneously the mass-separated 221Fr beam can be collected {unit 14}, which decays to a pure 213Bi sample. Thus, method D utilizes a combination of units 1, 2, 3, 7, 11, 12, 13 and 14. E. Indirect On-Line Production of Pure 212,213Bi Samples In a variant of method C the target is connected via a cold trap to a plasma ion source {units 1-3, 7, 11-13 as before}. Thus very pure beams of radon isotopes can be produced. The mass-separated 220Rn beam is collected {unit 14} and decays to a pure 212Pb/212Bi sample. Simultaneously the mass-separated 221Rn beam can be collected {unit 14}, which decays to a pure 213Bi sample. Thus, method E utilizes a combination of units 1, 2, 3, 7, 11, 12, 13 and 14. A New, Dry 225Ac/213Bi generator Making use of the fact that actinides form rather stable carbides, 225Ac can be bound in a graphite matrix. Heating the latter to temperatures around 1400-2000° C., Ac will remain in the matrix, while the decay daughters 221Fr, 217At and 213Bi are easily released {units 2,3}. They can be condensed {unit 8} on a cooler surface which acts as catcher of the 213Bi product. Instead of a graphite matrix, 225Ac can also be adsorbed, implanted or alloyed onto/into a suitable metallic matrix. A New, Dry 228Th/224Ra Generator Also 228Th can be bound in a graphite or metallic matrix. Heating this matrix to temperatures around 1600-2200° C., Th will remain in the matrix, while the decay daughter Ra is released {units 2,3}. It can be condensed {unit 8} on a cooler surface and extracted. A New, Dry 224Ra/212Pb/212Bi Generator 224Ra is bound in a porous matrix of e.g. a fatty acid salt, a metal hydroxide or oxide. Emanation of the decay daughter 220Rn occurs at room temperature and can be accelerated by heating the matrix {units 2,3}. The emanating 220Rn is condensed {unit 8} on a cold surface which acts as catcher of the 212Pb/212Bi product or collected electrostatically from the gas phase. A longer-lived generator can be obtained by replacing the 224Ra with 228Th or by a combination with the methods 7. and 8. (i.e. the dry 228Th/224Ra generator and the dry 224Ra/212Pb/212Bi generator), by keeping the 228Th generator at a temperature where 224Ra is not released, but 220Rn emanates. A New, Dry 227Ac/227Th/213Ra Generator 225Ac can be bound in a graphite matrix which will also bind the decay daughter 227Th. Heating the matrix to temperatures around 1600-2000° C., Ac and Th will remain in the matrix, while the decay daughters 223Fr and 223Ra are easily released {units 2,3}. They can be condensed {unit 8} on a cooler surface which acts as catcher of the 223Ra product. Instead of a graphite matrix, 225Ac can also be adsorbed, implanted or alloyed onto/into a suitable metallic matrix. Particular Advantages Include: The inventors provide a new general method of 225Ac production. The inventor's production methods can start from natural or depleted uranium and natural thorium targets. These are cheaper and easier to handle than the normally necessary 226Ra, 228Th, 229Th, etc. The inventors provide to collect mass-separated Fr or Rn isotopes, which decay then to isotopically pure Bi or Pb samples. The inventors have performed a demonstration for the on-line production of isotopically pure 212Bi samples as decay product of mass-separated 220Fr ion beams (“proof-of-principle”). The inventors provide new types of dry generators, which avoid wet chemical waste and surpass the activity limitations of conventional ion exchange generators, which are subject to radiation damage. Selecting the implantation energy one can choose freely between a radioactive source, which is “closed” or “open” for the release of daughter recoils. Particularly strong undissociable bonds to nanoparticles can be obtained by the ion-implantation labelling. The continuous, automated production without manual operation steps, ideally suited for industrial production, is demonstrated. In summary, the methods provided within this embodiment comprise the following features: This approach works for the isotopes 212Pb, 212Bi, 213Bi, 223Ra, 224Ra, 225Ra and 225Ac which are alpha emitters or decay parents of alpha emitters. A target made from metallic 232Th or a compound or alloy containing 232Th is irradiated by medium or high energy (>50 MeV) particles. A target made from natural uranium or 238U partially or fully depleted in 235U or a compound or alloy containing these isotopes is irradiated by medium or high energy (>80 MeV) particles. Some of the target materials can be in form of foils, wires, powder, foam, etc. The wanted products close the target are produced by spallation by a medium or high energy (>80 MeV) particle beam provided by a particle accelerator (cyclotron, LINAC, synchrotron, etc.). Combined with conventional radiochemical separation from the target 225Ac samples are obtained containing 0.1-1% relative activity 227Ac. Non target-destructive extraction of 225Ac samples with 0.1-1% 227Ac are obtained by dry high-temperature separation of the nuclear reaction products from the target material combined with conventional radio chemistry. During or after the irradiation the target is kept at a temperature of >1200° C. The entire target enclosure and all surfaces which the released Ac can encounter, except a catcher, is kept at a sufficiently high temperature to avoid condensation of Ac at places other than the cooled Ac catcher. This non-destructive batch-mode operation has the advantage that the same target unit can be used many times and the amount of liquid waste is reduced. Monoisotopic 225Ac samples are obtained by removing the 227Ac from the purified Ac batch using mass separation. The mass separation can be performed with any mass-selective device, e.g. a Wien-filter, a radio-frequency quadrupole, etc. instead of the magnetic sector field. The Ac containing oven for feeding the ion source and ion source are heated by any suitable mean (Ohmic heating, electron bombardment, radio-frequency, infrared heating, laser heating, energy loss of the incident beam, etc.) or any combination of these methods. The effusing radioisotopes can be transported by a flow of inert gas (He, Ar, . . . ) to the ion source instead of the transport by vacuum diffusion. The target is connected to the ion source in a way that no other escape path is available for the radioisotopes. The desorption and transport of Ac to the catcher or the ion source can be accelerated by chemical evaporation, adding a small amount of suitable agent (halogens or volatile halogenated compounds). Surface or plasma ionisation of Fr, Ra and Ac can be used as well as resonant laser ionisation with laser light generated from dye lasers, Ti:sapphire lasers or any other type of wavelength tunable light sources (OPO, . . . ) which are pumped by solid state lasers (Nd:YAG, Nd:YLF, Nd:YVO, diode, . . . ) or gas lasers (copper vapour lasers, etc.). The wanted 223,224,225Ra isotopes are produced in a continuous on-line or discontinuous but still fully automated method in which the target is connected directly to the ion source of a mass separator. Pure 212Pb and 212,213Bi samples too are produced in a continuous on-line fully automated method in which the target is connected directly to the ion source of a mass separator and the ion source type and target temperature are selected and or adjusted to make beams of their Fr or Rn precursors. The availability of the wanted alpha emitters in ion beam form allows to label nanoparticles, other substrates or chemical compounds that facilitates the labeling of bioconjugates. By selecting the implantation energy, the implantation depth can be adjusted that alpha-recoils can either be ejected and emanate (implantation energy typically <100 keV leads to a low implantation depth), thus representing an open source, or that alpha-recoils cannot leave the matrix (implantation energy typically >150 keV leads to a deeper implantation depth), hence representing a closed source. Implantation can be performed through a suitable thin cover layer into the collection matrix. Thus the source can be transported as “closed”. The end user can easily remove the cover layer by dissolving, evaporating, burning, mechanically removing, etc. to obtain an open source with well-defined depth profile. A number of new dry isotope-generators can be made by either incorporating the purified precursor isotopes chemically or directly by ion implantation in a suitable substrate. New, dry forms of isotope generators 228Th/224Ra, 224Ra/212Pb/212Bi, 228Th/212Pb/212Bi 225Ac/213Bi and 227Ac/227Th/223Ra are described. They are all based on the fact that the mother isotope(s) is/are bound in the matrix while the daughter isotopes can emanate at the given temperature and are collected on a suitable catcher. 1. Application: 211At is a very promising isotope for targeted alpha therapy (TAT), but it does not emit positrons. Therefore this isotope is not useful for imaging via PET (positron emission tomography). The provided astatine isotopes 204-207At allow for the first time to use this powerful technique for diagnostics in vitro (development of new At-labelled compounds) and in vivo (individual dosimetry to adapt the dose of a 211At-TAT). Moreover the gamma emission with high branching ratios of 204-210At allows to use the latter as convenient radiotracers for biodistribution studies or even for in vitro or in vivo dosimetry with SPECT (single photon emission computerized tomography). 2. Method: 1a) On-Line Production of a Cocktail of Different At Isotopes A molten Bi target is irradiated with protons of >140 MeV energy {unit 1}. 210-xAt isotopes are produced by 209Bi(p,pi−xn) double charge-exchange reactions as well as by secondary 209Bi(alpha,xn) and 209Bi(3He,xn) reactions with the alpha and 3He produced in (p,alpha) and (p,3He) reactions respectively. Astatine is released {units 2,3} and is transported either under vacuum or in inert gas {unit 7} to a suitable catcher {unit 8} surface, e.g. silver. No polonium is released for temperatures below 500° C. The catcher is mounted in a way to be easily changeable once the desired amount of At has been collected on it. This method will produce a mixture of astatine isotopes which can be used as gamma-emitting radiotracers, e.g. for biodistribution studies. Thus, method 1a) utilizes a combination of units 1, 2, 3, 7 and 8. 1b) Off-Line Separation of Individual At Isotopes. Optionally isotopically pure samples can be produced by inserting the catcher containing the At isotopes (produced according to 1a, {units 1-3, 7, 8}) into an oven {unit 10} attached or integrated into an ion source. Heating the catcher will release the At which is transported either under vacuum or in inert gas flow {unit 7} to a plasma ion source {unit 11} where At is single positively ionized. The ions are extracted from the ion source, accelerated to typically several tens of keV {unit 12} and separated in a magnetic sector field according to the mass/charge ratio {unit 13}. The ions are collected {unit 14} on backings covered with a thin film of salt which is later dissolved and used to label a bio-conjugate, or a refractory material {unit 14} from where the At can be released in gaseous form by dry distillation. Alternatively the ions are directly implanted into nanoparticles, etc. {unit 14} for labelling of the latter. Thus, method 1b) utilizes a combination of units 1, 2, 3, 7, 8, 10, 11, 12, 13 and 14. 2) On-Line Production and Separation of Individual At Isotopes A molten Bi target is irradiated with protons of >140 MeV energy {unit 1}. 210-xAt isotopes are produced by 209Bi(p,pi−xn) double charge-exchange reactions as well as by secondary 209Bi(alpha,xn) and 209Bi(3He,xn) reactions with the alpha and 3He produced in (p,alpha) and (p,3He) reactions respectively. Astatine is released {units 2,3} and is transported either under vacuum or in inert gas flow {unit 7} to a plasma ion source {unit 11} where At is single positively ionized. The ions are extracted from the ion source, accelerated to typically several ten keV {unit 12} and separated in a magnetic sector field according to the mass/charge ratio {unit 13}. The ions are collected on backings {unit 14} covered preferably with a thin film of salt which is later dissolved and used to label a bio-conjugate. Alternatively the ions are directly implanted into nanoparticles, etc. {unit 14} for labelling of the latter. Thus, method 2) utilizes a combination of units 1, 2, 3, 7, 11, 12, 13 and 14. Particular Advantages Include: No astatine isotope has so far been used for PET imaging. The inventors provide astatine isotopes for PET imaging and as convenient gamma emitters for biodistribution studies and/or in vitro or in vivo dosimetry. Due to the higher branching ratio for gamma emission compared to 211At, the ratio “signal to radiotoxicity” is improved by a big factor (orders of magnitude). The inventors have performed a demonstration of the on-line production of mass-separated astatine beams from a Pb/Bi target irradiated with 1.4 GeV protons (“proof-of-principle”). The inventor's method allows to collect the At isotopes parasitically from any liquid Bi containing target irradiated with high energy particles, e.g. from Pb/Bi targets used in spallation neutron sources, ADS, etc. Particularly strong undissociable bonds to nanoparticles can be obtained by the ion-implantation labelling. The continuous, automated production without manual operation steps ideally suited for industrial production is demonstrated. In summary, the methods provided within this embodiment comprise the following features: A pure Bi metallic target or a Bi containing alloy is used as target. During irradiation the target is kept in a temperature range between the melting point and ca. 500° C. The temperature is controlled by heating/cooling the target vessel and/or heating/cooling the target material when the latter is flowing in a circuit. The liquid target material can be standing as a bath in a container, be a free-standing jet or a flow enclosed on one or more sides by a wall. Variant: the target temperature can exceed 500° C. if the then also released Po is separated from the At in a subsequent standard radiochemical separation or left in the radioisotope product if it is not considered as disturbing for the application. The incident proton beam with >140 MeV energy is provided by a particle accelerator (cyclotron, LINAC, synchrotron, etc.). The proton beam can enter the target enclosure via a window or via a differentially pumped section. The proton beam can be replaced by a beam of light or heavy ions (d, 3He, alpha, etc.). The target material can be kept in motion by pumping, mechanical shaking, electromagnetic agitation, etc. to assure a better temperature homogeneity and thus allow for higher beam currents without the risk of local overheating. The inert gas can be replaced by any other gas if the latter is compatible with the integrity of the target, the enclosure and the catcher surface. A chimney or baffles can be used to condense evaporating target material before it reaches the catcher or ion source. The entire target enclosure and all surfaces which the released astatine can encounter, except the catcher, is kept at a sufficiently high temperature to avoid a condensation of astatine at places other than the catcher. Any of the known catchers can be used: e.g. Ag, silica gel or cooled surfaces of plastic, quartz, etc. or water or other solvents. Several chambers with catchers can be attached to the target chamber and connected/disconnected from the latter without interruption of the irradiation for a significant time. Variant: instead of on-line separation, the irradiation can be performed at a reduced target temperature. The target is then heated afterwards when needed to release the astatine. The target (and ion source respectively) are heated by any suitable mean (Ohmic heating, electron bombardment, radio-frequency, infrared heating, laser heating, energy loss of the incident beam, etc.) or any combination of these methods. The target is connected to the ion source in a way that no other escape path is available for the radioisotopes. The effusing radioisotopes can be transported by a flow of inert gas (He, Ar, . . . ) to the ion source instead of being transported by vacuum diffusion. The mass separation can be performed with any mass-selective device, e.g. a Wien-filter, a radio-frequency quadrupole, etc. instead of the magnetic sector field. The mass-separated ion beam is implanted into a salt layer. The salt layer containing the radioisotopes is subsequently dissolved in a small volume of water or the eluting agent. The salt cover of the backings can be replaced by many other water-soluble substances (sugar, . . . ) or by a thin ice layer (frozen water or other liquid). Instead of dissolving, the latter is subsequently melted by heating with any suitable method (Ohmic heating, infrared heating, radio-frequency heating, . . . ). Instead of a soluble matrix, the ion beam can also be implanted into any other solid matrix, e.g. a metal foil. In this case one needs additionally a chemical separation of the desired isotope from the matrix material that usually disturbs the chromatographic process. In all cases (i.e. elution from the catcher, dissolving of salt, etc. layer, melting of ice layer) the product fraction is usually obtained in a small volume and can be directly used for the labelling procedure of bio-conjugates or be directly injected into a chromatographic system for further purification. A particularly simple separation that allows to obtain At in gaseous form can be achieved by thermal release from a refractory matrix. Any of the classical radio-chemical and radio-chromatographical processes (precipitation, electrochemical separations, extraction, cation exchange chromatography, anion exchange chromatography, extraction chromatography, thermo chromatography, gas chromatography, etc.) suitable for the separation of astatine can be applied for the separation of the desired product from isobars and pseudo-isobars (stemming from molecular sidebands like oxides or fluorides appearing at the same mass settings), from daughter products generated by the radioactive decay of the collected radioisotopes during collection and processing and from other impurities. Ligands used for the chemical separation process are eventually remaining with the product fraction and need to be eliminated before further labelling procedures. Evaporation is the most suitable way for many cases. Nanoparticles, macro-molecules, microspheres, macroaggregates, ion exchange resins or other matrices used in chromatographic systems can be labelled directly by implanting the radioactive ion beam into them. For cases where the radioisotopic purity is already sufficient or for implantation into ion exchange resins or other matrices used in chromatographic systems, this can be done directly on-line. Else, after the standard purification steps (radio-chromatographic separation of isobars) the product is again injected into an ion source, ionized, accelerated and implanted. The so obtained products are carrier-free and isotopically pure. The process can be operated with all the technological steps of the chain as described. However, one can reduce freely the number of steps in many cases to adapt to the required purity of the respective application. 1. Method: The universal method will be outlined in the following preferred examples: First Variant Production of Carrier-Free 149Tb and Use for In Vivo Application The radionuclide 149Tb (T1/2=4.118 h) has a 16.7% branching ratio for alpha-emission. It is the most promising lanthanide isotope for targeted alpha therapy (TAT). A very high specific activity is crucial for the success of TAT. The isotope (149Tb as example) is generated (among many other isotopes) by irradiating a Ta-foil target kept at a temperature above 2000° C. with energetic particles, e.g. high energy protons (E>100 MeV) {unit 1}. The generated rare earth isotopes are escaping {units 2,3} from the target material and transported in vacuum {unit 7} to the ion source, where they are ionised into single positively charged ions using any kind of ionisation principles (surface ionisation, resonant laser ionisation or plasma ionisation) {unit 11}. The ions are extracted from the ion source, accelerated to typically several tens of keV {unit 12} and separated in a magnetic sector field according to the mass/charge ratios into isobars {unit 13}. The isobars are collected {unit 14} on backings covered preferably with a thin film of a salt. Thus, the first variant utilizes a combination of units 1, 2, 3, 7, 11, 12, 13 and 14. In the following the “application unit” is utilized: By dissolving the layer in a small volume of clean water (typically 50-100 μl) the carrier free isotope solution is transferred to the top of a column for isobaric separation by means of any of the radio-chromatography processes. The carrier free 149Tb in mass separated form is obtained in a volume of ˜200 μl, if alpha-hydroxisobutyric acid (alpha-HIBA) and cation exchange resin is used for the chromatography. The ligand used for the chromatographic separation is removed by evaporation and the remaining 149Tb is dissolved in a suitable small volume of 50 mM HCl-solution. This solution is directly used for the labelling procedure of bio-conjugates. Bio-conjugates in this context are any protein (monoclonal antibodies, their fragments, HAS=Human serum albumin, microspheres or macro-aggregates made from HAS, other protein molecules, peptides and oligonucleotides that are conjugated with chelating groups through or without linking molecules. The labelling procedure is fast (less than 10 minutes at room temperature) and quantitative. The obtained labelled bio-conjugate does not need any further purification, as it is usually needed in other protocols. The labelled bio-conjugate can be directly injected into patients for diagnostic procedures or therapeutic protocols. The radio-bio-conjugates obtained in this way are used for diagnostic imaging procedures as SPECT (single photon emission computerized tomography), quantitative PET (positron emission tomography) imaging for individual in vivo dosimetry, or for targeted beta (using beta emitting isotopes), targeted alpha therapy (TAT) (using the alpha emitting 149Tb) or the Auger-therapy (using the Auger electron emitters). Second Variant Production of high-purity 82Sr for 82Sr/82Rb generators The radionuclide 82Sr (T1/2=25.3 d) decays via the EC process generating the short-lived positron emitting daughter nuclide 82Rb with 76.4 s half-life. This short-lived isotope is used in nuclear cardiology as myocardial perfusion tracer using positron emission tomography (PET). For this purpose special dedicated 82Sr/82Rb generator systems have been developed, where the main generator column can be replaced frequently. Today the 82Sr is produced either via spallation reaction using Nb or Mo as target material or by the 85Rb(p,4n)82Sr process using metallic Rb targets that are exposed to intense proton beams with en energy >70 MeV. The drawback of the existing technology is that 85Sr (T1/2=64.8 d) is unavoidable generated in an amount that is 3-5 times higher. Thus, the obtained 82Sr preparation is “contaminated” by a factor 3 to 5 larger amount of a longer lived Sr isotope, that generates 514 keV gamma radiation in its EC decay. This large 85Sr contribution causes larger shielding efforts for the transport and reduces the shelf-time of the generator in the routine clinical use. In addition the production process is accompanied with relatively large quantities of liquid radioactive waste. The inventors provide a non-destructive technique, that allows to produce the 82Sr without generating liquid radioactive waste and optional in isotopically clean form without the large 85Sr contamination. Version A Non-target-destructive off-line production of 82Sr without mass separation configure a target {unit 1} consisting of 0.2-1 mm thick plates or foils or wires made from Zr or related alloys keep the target in vacuum or inert atmosphere irradiate with high energy particles (E>76 MeV, preferable protons) to generate the 82Sr {unit 1}, that is initially homogenously distributed in the target matrix heat the target under vacuum (or inert gas conditions) up to 1100-1300° C. during 10-60 min the 82Sr diffuses {unit 2} to the target surface, evaporates {unit 3} into the vacuum and becomes adsorbed at another metal surface used as catcher {unit 8} foil (metals of the group 5, 6, 7 and 8, preferable Ta, Nb or W), kept at a temperature below 1200 cool the system down and remove the catcher foil, extract the Sr in a conventional chemical way As option one can use an inert gas flow to transport {unit 7} the Sr from the target unit into a catcher cavity, where the Sr is adsorbed at any cold surface provided for further chemical treatment. The original target can be reused for the next irradiation cycle. Thus, version A of the second variant utilizes a combination of units 1, 2, 3, 7 and 8. Version B Non-target-destructive production of high purity 82Sr with mass separation in off-line or on-line mode target unit according to version A {unit 1} the Sr released {units 2,3} from the target material is here ionised using a suitable ion source (e.g. surface ionisation) {unit 11} single positively charged Sr+ ions are extracted from the target ion source unit {unit 12} The extracted ions can be collected on a catcher {unit 14} before or after passing through a mass-selective device {unit 13}. The process can be operated on-line (irradiation and separation simultaneously) or off-line (long irradiation and time to time a short mass separation) High purity 82Sr is obtained Thus, version B of the second variant utilizes a combination of units 1, 2, 3, 11, 12, 13 and 14. 3. Variant: New Type of 44Ti/44Sc Generator The radionuclide 44Sc (T1/2=3.9 h) is a suitable positron emitter (beta+ branching ratio of 94.34%) and has as such a great future potential in nuclear medical functional imaging using positron emission tomography (PET). Presently new approaches for systemic radionuclide therapy are under development, that are based on bio-selective molecules, liposomes or nanoparticles, that are used as carrier vehicle to transport therapeutic radionuclides into tumor cells or tumor tissue. The quantitative information of the bio-distribution will be accessed using PET imaging based on positron emitting radionuclides of elements that are homologues of the element used for the therapy. Thus, metallic position emitting radionuclides with a half-life of few hours are most suitable to perform this kind of studies and are demanded. 44Sc is most suitable for this kind of studies, but by far not available today and not in the required quantity. 44Sc can be made available from the decay of the mother isotope 44Ti (half-life 60 years). The inventors provide a new type of 44Ti/44Sc generator principle. Due to the long half-life the well known principle used in the 99Mo/99mTc generator cannot be applied here. Ti in form of pure metal or alloy will be irradiated {unit 1} with medium or high energy (E>20 MeV) charged (e.g. protons) or neutral particles to generate in a non-selective way the radionuclide 44Ti inside the target matrix. In addition other isotopes are formed, mainly of the elements Sc, Ca, K and Ar. After a certain cooling period (to let the short-lived isotopes decay) the Ti-target will be annealed at a temperature ≧1000° C., in order to release most of the remaining radioactivity except the 44Ti. The inventors have studied extensively the transport processes of tracer elements inside the Ti-matrix and determined the corresponding diffusion coefficients {unit 2}. In this systematic studies the inventors learnt, that the tracer elements are released from Ti-matrix in the following order: Sc>Ca>K>Ar. The diffusion {unit 2} of Sc is fastest and already at relatively low temperatures one can separate Sc from a thick Ti-matrix within relatively short time. The adsorption enthalpy of Sc at the Ti surface is low, consequently Sc is evaporated {unit 3} from Ti at relatively low temperatures. On the other hand the adsorption enthalpy of Sc on the surface of most noble or refractory metals (i.e. Ta, W, Re, Pt, Au, . . . ) is high. Consequently Sc is adsorbed {unit 8} to those surfaces at the same temperature where it is released from the Ti-matrix. The annealing procedure with the transport of the Sc from the Ti target to the adsorbing surface can be performed in vacuum or in an inert gas atmosphere {unit 7}. The 44Sc adsorbed at the metal surfaces is then removed by any of the known techniques (dissolution, electrochemical, desorption) and conditioned for the use in tracer molecule labelling. The process can be repeated without limitations, since the half-life of the 44Ti is very long and the Ti matrix does not change its behaviour. Thus, the third variant utilizes a combination of units 1, 2, 3, 7 and 8. Particular Advantages Include: The inventors have shown for the first time by an in vivo experiment that 149Th can be successfully applied for TAT (“proof of principle”). The quality of the radioisotope product generated according to the inventor's method allows the application of primary-labelled bioconjugates without further treatment or purification (usually required in alternative production routes). Many of the listed radioisotopes of interest which will become available with the inventor's method are provided for the first time for application in life science research and medical application. In most of the provided variants the target is reusable many times, which is not the case in standard methods where the irradiated target is destroyed by dissolution, giving large amounts of liquid waste. For the first time the inventors provide the direct labelling of nanoparticles, macro-molecules, etc. by radioactive ion implantation. This yields in many cases a stronger undissociable bond than presently used methods. For the first time the inventors provide the production of mass-separated 82Sr which can serve for improved 82Sr/82Rb generators and have shown the parameter range where Sr is released (“proof of principle”). For the first time the inventors provide a dry 44Ti/44Sc generator and have shown the parameter range where Sc is released (“proof of principle”). The inventors provide that the so produced 44Sc is used as PET isotope e.g. as representative tracer for quantitative biodistribution studies of radiolanthanide labelled bio-selective molecules, lyposomes, nanoparticles, etc. The method is ideally suited for large scale industrial production since it has been demonstrated that it is continuous and automated with few manual operation steps. In summary, the methods provided within this embodiment comprise the following features: This approach works for 28Mg, 42K, 43K, 45Ca, 47Ca, 81Rb, 82Sr, 83Sr, 85Sr, 89Sr, 172Hf, 175Hf and all radionuclides of the rare earth elements out of which the following have dedicated relevance: 44Sc, 44mSc, 46Sc, 47Sc, 85Y, 86Y, 87Y, 88Y, 134Ce/La, 137Ce, 139Ce, 141Ce, 143Pr, 138Nd/Pr, 140Nd/pr, 147Nd, 149Pm, 142Sm/Pm, 153Sm, 155Eu, 147Gd, 149Gd, 149Tb, 152Tb, 155Th, 161Tb, 157Dy, 159Dy, 166Ho, 165Er, 169Er, 165Tm, 167Tm, 169Yb, 177Yb, 172Lu, 177Lu. The Ta target can be replaced by Hf, W, Re, Ir or alloys or compounds containing any of these metals. This material can be used in pure form or mixed with other materials. For producing radioisotopes of the lighter elements Mg, K, Ca, Sc, Rb, Sr, Y also targets made from Zr, Nb, Mo, Ru, Rh or alloys or compounds containing these elements can be used, in addition to the ones mentioned above. For producing radioisotopes of the lighter elements Mg, K, Ca, Sc also targets made from Ti or V or alloys or compounds containing these elements can be used, in addition to the ones mentioned above. For producing 28Mg also targets made from Si or alloys or compounds containing Si can be used, in addition to the ones mentioned above. The target can be replaced by the distillation residue from previously irradiated targets of Hg, Pb, Bi or an alloy containing any of these elements. The target material can be in form of foils, wires, powder, foam, etc. The target material, the target enclosure, the ion source and all surfaces the effusing radioisotopes might interact with, are held at high temperature. “High” means of the order of 60-90% or preferably 60-95% of the melting point of the material. The target is connected to the ion source in a way that no other escape path is available for the radioisotopes. The incident proton beam can be replaced by energetic light ions (d, 3He, 4He, . . . ), heavy ions, neutrons, electrons or photons. The target and ion source are heated by any suitable mean (Ohmic heating, electron bombardment, radio-frequency, infrared heating, laser heating, energy loss of the incident beam, etc.) or any combination of these methods. Variant: The target can also be kept at lower temperatures during irradiation, then being heated off-line for the release of the radioisotopes. After release of a sufficient amount of the radioisotopes the target is again irradiated, heated, irradiated, . . . (batch-mode operation). Variant: In cases where the isotope produced during irradiation is long-lived and decays to a daughter radioisotope of interest, the once irradiated target can be used as a dry generator by heating it to a temperature where the daughter radioisotope is released while the long-lived radioisotope remains in the target matrix. A particular application of this method provides a new type of dry 44Ti/44Sc generator. dry 44Ti/44Sc generator: Ti in form of pure metal or alloy is irradiated with medium or high energy (E>20 MeV) charged (e.g. protons) or neutral particles to generate in a non-selective way the radionuclide 44Ti inside the target matrix. In addition other isotopes are formed, mainly of the elements Sc, Ca, K and Ar. After a certain cooling period and waiting period (to let the short-lived isotopes decay) the Ti-target will be annealed at a temperature ≧1000° C., in order to release most of the remaining radioactivity except the 44Ti. The diffusion of Sc is fastest and already at relatively low temperatures one can separate Sc from a thick Ti-matrix within relatively short time. The adsorption enthalpy of Sc at the Ti surface is low, consequently Sc is evaporated from Ti at relatively low temperatures. On the other hand the adsorption enthalpy of Sc on the surface of most noble or refractory metals (i.e. Ta, W, Re, Pt, Au, . . . ) is high. Consequently Sc is adsorbed to those surfaces at the same temperature where it is released from the Ti-matrix. The annealing procedure with the transport of the Sc from the Ti target to the adsorbing surface can be performed in vacuum or in an inert gas atmosphere. The 44Sc adsorbed at the metal surfaces is then removed by any of the known techniques (dissolution, electrochemical, desorption) and conditioned for the use in tracer molecule labelling. The process can be repeated without limitations, since the half-life of the 44Ti is very long and the Ti matrix does not change its behaviour. In this particular process isotopically pure 44Sc is obtained even without mass separation since no other Sc isotope is produced as daughter of a long-lived mother isotope remaining in the Ti matrix. The effusing radioisotopes can be transported by a flow of inert gas (He, Ar, . . . ) to the ion source instead of the transport by vacuum diffusion. The target is connected to the ion source in a way that no other escape path is available for the radioisotopes. The desorption and transport to the ion source can be accelerated by chemical evaporation, adding a small amount of suitable agent (halogens or volatile halogenated compounds). Resonant laser ionisation can be performed with laser light generated from dye lasers, Ti:sapphire lasers or any other type of wavelength tunable light sources (OPO, . . . ) which are pumped by solid state lasers (Nd:YAG, Nd:YLF, Nd:YVO, diode, . . . ) or gas lasers (copper vapour lasers, etc.). Resonant laser ionization is particularly efficient if several or all thermally populated low-lying atomic states of the element to be ionized are simultaneously resonantly excited. This applies e.g. to the element terbium where several of the atomic states 4f9 6s2 6Ho15/2, 4f8(7F)5d6s2 8G13/2, 4f8(7F)5d6s2 8G15/2, 4f8(7F)5d6s2 8G11/2 have to be resonantly excited simultaneously with separate laser beams to the corresponding excited states and from there (via an optional intermediate step) to the continuum or to an autoionizing state. The mass separation can be performed with any mass-selective device, e.g. a Wien-filter, a radio-frequency quadrupole, etc. instead of the magnetic sector field. Often several isotopes of the same element, or isobars with comparable masses are produced in the same system. In this case a mass-selective device is of advantage, which allows to collect simultaneously several masses. The mass separation process is of particular importance if another radioisotope of the element in question is produced in such quantities that it causes a high radiation dose rate (problems of handling), e.g. in the case of 82Sr which is disturbed by the co-production of 85Sr. The mass-separated ion-beam is implanted into a salt layer. The salt layer containing the radioisotopes is subsequently dissolved in a small volume of water or the eluting agent and as such directly injected into the chromatographic system. The salt cover of the backings can be replaced by many other water-soluble substances (sugar, . . . ) or by a thin ice layer (frozen water or other liquid). Instead of dissolving, the latter is subsequently melted by heating with any suitable method (Ohmic heating, infrared-heating, radio-frequency heating, . . . ). Instead of a soluble matrix, the ion beam can also be implanted into any other solid matrix, e.g. a metal foil. In this case one needs additionally a chemical separation of the desired isotope from the matrix material that usually disturbs the chromatographic process. Any of the conventional radio-chemical and radio-chromatographical processes (precipitation, electrochemical separations, extraction, cation exchange chromatography, anion exchange chromatography, extraction chromatography, thermo chromatography, gas chromatography, etc.) suitable for the separation of rare-earth elements can be applied for the separation of the desired product from isobars and pseudo-isobars (stemming from molecular sidebands like oxides or fluorides appearing at the same mass settings), from daughter products generated by the radioactive decay of the collected radioisotopes during collection and processing and from other impurities. A particularly simple and efficient separation from the implantation substrate can be achieved by thermal release from a refractory matrix. The product fraction is usually obtained in a small volume. Ligands used for the chemical separation process are eventually remaining with the product fraction and need to be eliminated before further labelling procedures. Evaporation is the most suitable way for many cases, e.g. for alpha-HIBA. The remaining product is dissolved in a small volume of solution suitable for direct labelling, e.g. 50-100 mM HCl. The obtained solution is directly used for the labelling procedure of bio-conjugates. Bio-conjugates in this context are any protein (monoclonal antibodies, their fragments, HAS=Human serum albumin, microspheres or macro-aggregates made from HAS, other protein molecules, peptides and oligonucleotides that are conjugated with chelating groups (e.g. pure or derivatives of DTPA, DOTA or any similar type) through or without linking molecules. The labelling procedure is fast (less than 10 minutes at room temperature) and quantitative. The obtained labelled bio-conjugate does not need any further purification, as it is usually needed in other protocols. The labelled bio-conjugate can be directly injected into patients for diagnostic procedures: SPECT=single photon emission computerized tomography (e.g. 157Dy), quantitative PET=positron emission tomography imaging for individual in vivo dosimetry (e.g. 83Sr, 138Nb, 142Sm, etc.) or for therapeutic protocols: radio-immuno-therapy (RIT) using beta emitting isotopes (e.g. 169Er), targeted alpha therapy (TAT) using the alpha emitting 149Tb, or Auger-therapy using the Auger electron emitters (e.g. 165Er). Nanoparticles, macro-molecules, micro-spheres, macro-aggregates, ion exchange resins or other matrices used in chromatographic systems can be labelled directly by implanting the radioactive ion beam into them. For cases where the radioisotopic purity is already sufficient or for implantation into ion exchange resins or other matrices used in chromatographic systems, this can be done directly on-line. Else, after the standard purification steps (radio-chromatographic separation of isobars) the product is again injected into an ion source, ionized, accelerated and implanted. The so obtained products are carrier-free and isotopically pure. The process can be operated with all of the technological steps in the chain as described. However, one can reduce freely the number of steps in many cases without disturbing the final quality of the labelled product for in vivo application. 1. Application: The radionuclide 153Sm (T1/2=46.3 h) is a beta-emitting isotope with great potential in endo-radionuclide therapy. It is mainly used today in EDTMP solution for palliative treatment of bone cancer. Monoclonal antibody conjugates can be labelled as well, while, on the other hand the use for peptide labelling is hampered due to the insufficient specific activity. 153Sm is produced today generally via the 152Sm(n,gamma)153Sm process using enriched 152Sm as target material. 2. Method: For the production of neutron-rich lanthanide isotopes the inventors provide the following methods, outlined at the example of 153Sm: 153Sm can be found among the products of thermal neutron induced fission of 235U in reasonable quantities (0.15% cumulative yield). High-energy fission (e.g. induced by high energy protons) increases this yield significantly and moreover removes the restriction to the fissile target nuclides (as 235U or 239Pu) and 238U or 232Th become also useful as target. The separation of the Sm-fraction from the fission product mixture provides a 153Sm preparation in non carrier added quality, with a much higher specific activity than via the classical 152Sm(n,gamma)153Sm process. Version 1: Classical Chemical Separation of 153Sm from Fission Products fission target: any fissile isotope as well as Th and U in natural composition or depleted 238U, in any possible chemical form: metallic, carbide, oxide, sulphide, etc. irradiate with thermal or fast neutrons, charged particles, electrons or photons for initiating the fission process {unit 1} conventional (wet-chemical) process for separation of the Sm fraction, the Sm-fraction will contain only 153Sm and traces of 151Sm (93 year half-life) and small fission produced quantities of stable Sm-isotopes. The 153Sm/151Sm ratio can be optimized by reducing the time between start of irradiation and Sm separation.Version 2: Method Described in Version 1, Combined with Off-Line Mass Separation insert the obtained (along version 1) Sm fraction {unit 9} into a dedicated ion source cavity (=oven) {unit 10} of amass separator evaporate the Sm {unit 10} and ionise it by using surface ionisation, laser ionisation or plasma discharge ionisation {unit 11} to obtain Sm+ ions that are extracted, accelerated {unit 12} and separated using a dedicated mass-dispersive device {unit 13} 153Sm samples produced along the 152Sm(n,gamma)153Sm process can be transformed by the same method into carrier free quality preparations as well. Thus, version 2 utilizes a combination of units 1, 9, 10, 11, 12 and 13. Version 3: Non-Target-Destructive On- or Off-Line Separation of 153Sm target material in the chemical form of carbide or carbide diluted in excess graphite create fission via one of the mentioned nuclear reactions {unit 1} heat the target during or after irradiation to temperatures above 2000° C. Sm generated in the fission process is released {units 2,3} from the target material and transported to the ion source under vacuum or inert gas flow {unit 7} Sm is ionised via surface ionisation or/and laser ionisation or plasma ionisation {unit 11} the single charged positive Sm ions are than extracted from the target ion source unit, accelerated {unit 12} and separated by passing through a mass-selective device {unit 13} carrier free 153Sm is obtained in atomic form or in a molecular sideband (oxide, halide {unit 9}) Thus, version 3 utilizes a combination of units 1, 2, 3, 7, 9, 11, 12 and 13. Variant: With the methods of versions 2 and 3 also non-carrier added 117mSn can be produced. With high-energy fission {unit 1} 117mSn is directly populated (low-energy fission populates mainly the lower-Z mass-117 isobars which decay mainly to 117gSn) and the isomeric ratio 117mSn/117gSn is strongly enhanced. Using resonant laser ionization {unit 11} the ratio 117mSn/117gSn can be enhanced further, by either using the selection rules between magnetic substrates (more transitions possible for atoms with high total spin F) or by tuning a small-bandwidth laser to selectively ionize 117mSn via its hyperfine structure differing from that of 117gSn. Particular Advantages Include: The inventors have performed a demonstration for the on-line production of mass-separated 117mSn, 153Sm, 166Ho, 169Er, etc. beams from a UCx target irradiated with 1.4 GeV protons (“proof-of-principle”). The inventors provide a completely new production process (fission), which provides intrinsically higher specific activities. Non-carrier added samples of 117mSn can be obtained where the 117mSn/117gSn ratio is improved by several orders of magnitude compared to the conventional production via 116Sn(n,gamma). The continuous, automated production without manual operation steps ideally suited for industrial production is demonstrated. Particularly strong undissociable bonds to nanoparticles can be obtained by the ion-implantation labelling. In summary, the methods provided within this embodiment comprise the following features: Any fissile isotope as well as Th and U in natural composition or depleted 238U, in any possible chemical form: metallic, carbide, oxide, sulphide, etc. can be used as target. Some of the target materials can be in form of foils, wires, powder, foam, etc. The target is irradiated with thermal or fast neutrons, charged particles, electrons or photons for initiating the fission process. After irradiation a suitable conventional (wet-chemical) process is used for separation of the Sm fraction. The Sm-fraction will contain only 153Sm and traces of 151Sm (93 year half-life) and small fission produced quantities of stable Sm-isotopes. The 153SM/151Sm ratio can be optimized by reducing the time between start of irradiation and Sm separation. The Sm-fraction produced as described above is inserted into an oven connected to an ion source. Evaporation, ionisation, off-line mass separation and collection as described above in Embodiment V for on-line lanthanide separation. The wanted isotope can be separated as atomic ion or as molecular ion in the corresponding sideband (oxide, fluoride, . . . ). With the described mass separation process 153Sm samples produced along the 152Sm(n,gamma)153Sm process can be transformed into carrier free quality preparations as well. With the same methods 1-3 also other beta-emitting radioisotopes, e.g. 141Ce, 143Pr, 147Nd, 149Pm, 161Th, 166Ho and 169Er can be produced. The latter three require fast or high-energy fission to obtain a reasonable yield. Using high-energy fission, also non-carrier added 117mSn can be produced along the same methods. Using resonant laser ionization the ratio 117mSn/117gSn can be enhanced further, by either using the selection rules between magnetic substates (more transitions possible for atoms with high total spin F) or by tuning a small-bandwidth laser to selectively ionize 117mSn via its hyperfine structure differing from that of 117gSn. The described selective ionization of isomers can also be used in the separation process of other isomers, e.g. to improve the 177gLu/177mLu ratio. 1. Preferred Aspect: Investigation of Evaporation Characteristics of Po from Liquid Pb—Bi-eutecticum In a first preferred aspect of the present invention, the invention relates to an investigation of evaporation characteristics of polonium from liquid Pb—Bi-eutecticum The evaporation behaviour of polonium and its lighter homologues selenium and tellurium dissolved in liquid Pb—Bi-eutecticum (LBE) has been studied at various temperatures in the range from 482 K up to 1330 K under Ar/H2 and Ar/H2O-atmospheres using F-ray spectroscopy. Polonium release in the temperature range of interest for technical applications is slow. Within short term (1 h) experiments measurable amounts of polonium are evaporated only at temperatures above 973 K. Long term experiments reveal that a slow evaporation of polonium occurs at temperatures around 873 K resulting in a fractional polonium loss of the melt around 1% per day. Evaporation rates of selenium and tellurium are smaller than those of polonium. The presence of H2O does not enhance the evaporation within the error limits of the inventor's experiments. The thermodynamics and possible reaction pathways involved in polonium release from LBE are discussed. a. Introduction Liquid Lead-Bismuth eutecticum (LBE) is proposed to be used as target material in spallation neutron sources [Salvatores, M., Bauer, G. S., Heusener, G.: The MEGAPIE Initiative, PSI-Report Nr. 00-05, Paul Scherrer Institut, Villigen, Switzerland, 2000] as well as in Accelerator Driven Systems (ADS) for the transmutation of long-lived nuclear waste [Gromov, B. F., Belomitlev, Yu. S., Efimov, E. I., Leonchuk, M. P., Martinov, P. N. Orlov, Yu. I., Pankratov, D. V., Pashkin, Yu. G., Toshinsky, G. I., Chenukov, V. V., Shmatko, B. A., Stepankov, V. S.: Use of Lead-Bismuth Coolant in Nuclear Reactors and Accelerator-Driven Systems. Nuclear Engineering and Design 173, 207 (1997).]. In these systems polonium is formed as a product of (p,xn) and (n,γ)-reactions according to the following processes: Within one year of operation employing a proton beam current of 1 mA around 2 g of polonium are produced in this manner [Atchison, F.: Nuclide Production in the SINQ Target, Report SINQ/816/AFN-702, Paul Scherrer Institute, Villigen, Switzerland, 1997]. Because of the high radiotoxicity of polonium its behaviour is of utmost importance with respect to the safe operation and post-irradiation handling of the target systems and materials as well as for an assessment of the potential risk of accident scenarios. While the rates of evaporation and transport are of interest for an evaluation of the risk of contamination and incorporation in case of an accident, the development of suitable techniques for the fixation of polonium requires a fundamental knowledge of the chemical mechanisms of the release process. Previous thermal evaporation studies on polonium from molten Bi and Pb—Bi-eutecticum dealt with the preparation of polonium by neutron irradiation of bismuth and subsequent separation by distillation [Gmelin's Handbook of Inorganic and Organometallic Chemistry, 8th Edition, Polonium, Supplement Vol. 1, Springer-Verlag, Berlin, 1990, p. 421. Jennings, A. S., Proctor, J. F., Fernandez, L. P.: The Large Scale Separation of Polonium-210 from Bismuth. Du Pont Rep., Large Scale Production and Applications of Radioisoptes, DP-1066, 3, Du Pont de Nemour and Co, Aiken, S C, Savannah River Lab, Vacuum 17, 584 (1967).] and hazards related to the technical use of LBE in nuclear devices [Tupper, R. B., Minushkin, B., Peters, F. E., Kardos, Z. L.: Polonium Hazards Associated with Lead Bismuth Used as a Reactor Coolant. Proc. of the Intern. Conf on Fast Reactors and Related Fuel Cycles, Oct. 28-Nov. 1, 1991, Kyoto, Japan, Vol. 4, p. 5.6-1. Pankratov, D. V., Yefimov, E. I., Burgreev, M. I.: Polonium Problem in Lead-Bismuth Flow Target. Proc. of the Intern. Workshop on the Technology and Thermal Hydraulics of Heavy Liquid Metals, Mar. 25-28, 1996, Schruns, Austria, p. 9.23. Furrer, M., Steinemann, M., Leupi, P.: Dampfdruck von Polonium-210 über einer eutektischen Blei-/Wismut-Schmelze bei 350° C. TM-43-91-08, Paul Scherrer Institut, Villigen, Switzerland, 1991]. The thermodynamics of polonium release from molten LBE at temperatures between 673 and 823 K is investigated in [Buongiomo, J., Larson, C., Czerwinski, K. R.: Speciation of polonium released from molten lead bismuth. Radiochim. Acta 91, 153 (2003).]. Additionally, calculations of the polonium release rate based on a Langmuir-type formalism are reported [Yefimov, E. I., Pankratov, D. V.: Polonium and volatile radionuclides output from liquid metal target into ion guide and gas system. Proc. of the 2. Intern. Conf. on Accelerator-Driven Transmutation Technologies and Applications, Jun. 3-7, 1996, Kalmar, Sweden, p. 1121. Levanov, V. I., Pankratov, D. V., Yefimov, E. I.: The estimation of Radiation Danger of Gaseous and Volatile Radionuclides in Accelerator Driven System with Pb—Bi Coolant. Proc. of the 3. Intern. Conf. on Accelerator-Driven Transmutation Technologies and Applications, Jun. 7-11, 1999, Prague, Czech Republic, http://www.fjfi.cvut.cz/con_adtt99/. Fischer, W. E.: Dampfdruck und Aktivierung flüchtiger Spallationsprodukte aus dem SINQ-Target, Report SINQ/821/FIN-716, Schweizerisches Institut für Nuklearforschung, Villigen, Switzerland, 1987. Li, N., Yefimov, E., Pankratov, D.: Polonium Release from an ATW Burner System with Liquid Lead-Bismuth Coolant, Report LA-UR-98-1995, Los Alamos National Laboratory, U.S.A., 1998.]. The chemical mechanism of the release of volatiles can be influenced by the composition of the vapour phase. Hydrogen will be formed by spallation reactions in the operating target. Therefore, a certain amount of H2O will be present in the system, where the vapour pressure of H2O depends on the oxide content of the liquid alloy. In case of an accident, the alloy can be exposed to air. In this embodiment the inventors study the thermal release of polonium and its lighter homologues selenium and tellurium from LBE in an inert gas/hydrogen atmosphere. Some additional experiments employing an inert gas/water atmosphere were also conducted. For a suitable experimental setup, see Example 1. b. Results and Discussion The results of the short-term evaporation experiments are shown in FIGS. 2-4. A comparison of the release behaviour of selenium, tellurium and polonium from LBE (1 h experiments) in an Ar/7%-H2 atmosphere at temperatures between 482 and 1330 K is shown in FIG. 2. Measurable amounts of the chalcogens are released at temperatures starting from 973 K. The volatility increases in the order Se<Te<Po. Accordingly, the temperatures at which 50% of the total amount of chalcogen is released decrease from 1300 K (Se) to 1270 (Te) and 1200 K (Po). In the temperature range of interest for technical applications like liquid metal spallation targets (473-723 K) no release has been observed within the experimental errors indicated as error bars in the figures. FIG. 3 shows a comparison of the release behaviour of polonium in Ar/7%-H2 and water saturated Ar atmosphere. The presence of water does not lead to a pronounced increase of the volatility of polonium between 498 and 873 K. The sample investigated at 1108 K suffered from oxidation in the water-containing atmosphere and had reacted with water and the quartz tissue within an hour to form presumably a Pb/Bi-silicate glass. However, these chemical reactions do not lead to a significant increase or decrease of the polonium evaporation rate. FIG. 4 shows a comparison of the fractional release of polonium from LBE samples of different sizes as a function of temperature. For both sample sizes a measurable release of polonium occurs only at temperatures above 973 K. However, above this temperature the release of polonium from 0.14 g samples is about twice as fast as from 0.65 g samples. From an evaluation of the surface to volume ratios and the radius ratios of the two sample sizes no clear conclusion can be drawn with respect to a desorption- or diffusion-controlled process. However, a detailed evaluation of the mechanism of the release process is beyond the scope of this work. The results of the inventor's long-term experiments are presented in FIGS. 5 and 6. FIG. 5 shows the fractional release of polonium from LBE measured in an Ar/7%-H2 atmosphere at different temperatures as a function of time for periods up to 28 days. At 646 and 721 K, which are temperatures considered for the operation of liquid metal spallation targets using LBE as the target material, no release is observed within the limits of the experimental errors. At 867 K polonium evaporates slowly with an evaporation rate of the order of 1% per day. Even at temperatures as high as 968 K it takes about 12 days to remove 85% of the present polonium. Therefore, a large concentration of polonium in the cover-gas system of a LBE spallation target due to evaporation processes seems unlikely. However, for such a system the release of polonium by other processes like sputtering or the formation of aerosols and dusts has to be taken into account as well. A comparison of tellurium and polonium release from LBE in an Ar/7%-H2 atmosphere at 968 K is presented in FIG. 6. As already indicated by the results of short-term experiments the evaporation of Te from LBE is significantly slower than Po-evaporation. In general, the results of long-term experiments show that the mechanism of the evaporation process does not change over long periods of time, i.e. no change of the reaction path is indicated. For the time dependency an approximate linear relation to the square root of release time is observed (FIG. 7) as generally known for release processes. To assign or exclude possible reaction pathways the inventors evaluated some thermochemical properties of the main species that might be involved in such an evaporation process. The main gas phase species considered are monoatomic chalcogens Q, diatomic Q2 molecules, dioxides QO2, hydrides H2Q, hydroxides Q(OH)2 and the gaseous diatomic molecules PbQ and BiQ (Q=Se, Te, Po). From these species, the dioxides can be excluded because they will be reduced in the presence of hydrogen and metals such as lead. Equilibrium constants calculated from thermodynamic data [Barin, I.: Thermochemical Data of Pure Substances, VCH, Weinheim, 1995] for reactions such asQO2+2H2⇄Q+2H2O (3)andQO2+2Pb⇄2PbO+Q (4)indicate that the equilibrium is clearly dominated by the product side. This tendency is additionally increased by a stabilizing metal-chalcogen interaction (“coupled reduction”) in solution [Neuhausen, J., Eichler, B.: Extension of Miedema's Macroscopic Atom Model to the Elements of Group 16 (O, S, Se, Te, Po), PSI-Report 03-13, Paul Scherrer Institute, Villigen, Switzerland, September 2003]. Thermodynamic data for reactions of metal chalcogenides with hydrogen and water such asPbQ+H2⇄Pb+H2Q (5)andPbQ+H2O⇄PbO+H2Q (6)show that the formation of chalcogen hydrides is not favoured. Experimental investigations indicate that polonium hydride is thermally unstable. It is possibly formed only under the presence of nascent hydrogen [Gmelin's Handbook of Inorganic and Organometallic Chemistry, 8th Edition, Polonium, Supplement Vol 1, Springer-Verlag, Berlin, 1990, p. 421]. Within this work the inventors focus on monoatomic and diatomic chalcogens and diatomic lead and bismuth chalcogenides as gas phase species. For the volatilisation process the following pathways have to be considered:1) evaporation of the chalcogen Q from LBE in form of single atoms according toQ(solv)→Q(g) (7) Approximate values for the accompanying enthalpy of evaporation can be calculated by subtracting the partial molar enthalpy of solution of the chalcogen in the liquid metal Δ HsolvQ in metal(I) from the difference of the standard enthalpy of the gaseous monoatomic chalcogen ΔH Q(g) and its enthalpy of melting ΔHmQ:Δ HvQ=(ΔHQ(g)−ΔHmQ)−Δ HsolvQ in metal(I) (8) Temperature dependency has been neglected and the enthalpy of melting at the melting point has been used as an approximation for ΔHmQ. 2) evaporation as diatomic chalcogen molecules according to2Q(solv)→Q2(g) (9) In analogy to monoatomic evaporation the enthalpy for this process can be expressed as the difference between standard enthalpy of the gaseous diatomic chalcogen minus twice the melting enthalpy of the chalcogen and the enthalpy associated with the solution of two atoms of Q in the liquid metal, hence:Δ HvQ2=(ΔHQ2(g)−2ΔHmQ)−2Δ HsolvQ in metal(I) (10)3) evaporation in form of diatomic metal chalcogenides MQ (M=Pb, Bi; Q=Se, Te, Po)Q(solv)+M(1)→MQ(g) (11) The associated enthalpy can be calculated from the enthalpy values of the monoatomic species M and Q, their enthalpies of melting, the partial molar enthalpy of solution of the chalcogen Q in the liquid metal M and the dissociation enthalpy of the diatomic molecules MQ using the following equation:Δ HMQ=(ΔHQ(g)−ΔHmQ)+(ΔHM(g)−ΔHmM)−Δ HsolvQ in metal(I)−ΔHdissMQ(g) (12) The inventors have calculated enthalpies of evaporation for these processes using available thermochemical data for ΔHQ(g), ΔHM(g), ΔHmQ, ΔHmM and ΔHQ2(g) from [Barin, I.: Thermochemical Data of Pure Substances, VCH, Weinheim, 1995] (Se, Te) and [Eichler, B.: Die Flüchtigkeitseigenschaften des Poloniums, PSI-Report 02-12, Paul Scherrer Institute, Villigen, Switzerland, June 2002] (Po). Values for Δ HsolvQ in metal(I) have been calculated using Miedema's Macroscopic Atom Model [de Boer, F. R., Boom, R., Mattens, W. C. M., Miedema, A. R., Niessen, A. K.: Cohesion in Metals, Transition Metal Alloys, North-Holland, Amsterdam 1988]. Details of the parameterisation of the model and the calculation procedure can be found in [Neuhausen, J., Eichler, B.: Extension of Miedema's Macroscopic Atom Model to the Elements of Group 16 (O, S, Se, Te, Po), PSI-Report 03-13, Paul Scherrer Institute, Villigen, Switzerland, September 2003]. The values for Δ HsolvQ in metal(I) calculated in this way are very similar for the chalcogens in liquid Pb and Bi, respectively. Furthermore, LBE does not deviate to far from ideal behaviour. Therefore, the inventors give mean values for Δ HsolvQ in metal(I) calculated for lead and bismuth below. Values for the dissociation enthalpies of diatomic molecules ΔHdissMQ(g) are estimated using a method described in [Miedema, A. R., Gingerich, K. A.: On the enthalpy of formation of diatomic intermetallic molecules. J. Phys. B: Atom. Molec. Phys. Vol. 12, 2255, (1979)]. The values for dissociation enthalpies of homonuclear diatomic molecules M2 and Q2 required for these calculations have been taken from [Barin, I.: Thermochemical Data of Pure Substances, VCH, Weinheim, 1995, Eichler, B.: Die Flüchtigkeitseigenschaften des Poloniums, PSI-Report 02-12, Paul Scherrer Institute, Villigen, Switzerland, June 2002]. The results of these calculations are compiled in Table 1. From these values it can be concluded that evaporation of chalcogens from LBE in the form of lead chalcogenide molecules seems to be the least probable reaction path from a thermochemical point of view. For a discussion on the remaining possibilities the inventors discuss the release process as three possible series of successive reactions as shown in FIG. 8. For each of these reaction sequences the rate of release and hence the observed sequence of release rates (experimentally: Se<Te<Po) will be determined by the reaction step involving the highest energy of activation. Thus, if the release process is diffusion controlled the sequence of release rates will be determined by the sequence of activation energies of diffusion. Nevertheless, the actual species released could still be either of the three possibilities Q, Q2 or MQ. No literature data are available for diffusion of chalcogens in LBE. Therefore, the inventors have to rely on estimations for evaluating the corresponding activation energies. The energy of activation for the process of self-diffusion in liquid lead is 18.6 kJ/mole [Leymonie, C.: Radioactive Tracers in Physical Metallurgy, Chapman and Hall, London 1963, p. 112]. For diffusion of lead and bismuth in LBE activation energies of 9.6 and 7.7 kJ/mole, respectively, have been estimated from molecular dynamics calculations [Celino, M. Conversano, R., Rosato, V.: Atomistic simulation of liquid lead and lead-bismuth eutectic. J. Nucl. Materials 301, 64 (2002).]. Experimental values vary in the range from 11.6 to 40 kJ/mole [Landolt-Börnstein Zahlenwerte und Funktionen aus Physik, Chemie, Astronomie, Geophysik und Technik, 6. Auflage, II. Band, 5. Teil B, Springer, Berlin 1968]. For diffusion of selenium in liquid tin an activation energy of 13.4 kJ/mole has been determined [Landolt-Börnstein Zahlenwerte und Funktionen aus Physik, Chemie, Astronomie, Geophysik und Technik, 6. Auflage, II. Band, 5. Teil B, Springer, Berlin 1968]. Assuming similar or even somewhat larger values for chalcogen diffusion in LBE it still seems unlikely that diffusion is the rate determining step since activation energies for the evaporation step are expected to be in the order of magnitude of the enthalpies of evaporation compiled in Table 1. The enthalpy values for the evaporation of monoatomic chalcogens are in agreement with the experimentally observed sequence of evaporation rates. Assuming that the corresponding activation energy values are similar, this could be interpreted as a supporting fact for the release in the form of monoatomic chalcogens. However, if there is a high enough concentration of chalcogen in the liquid alloy to facilitate the formation of Q2 molecules, the evaporation in form of Q2 species should be favoured compared to the release as monoatomic chalcogens. This has to be mainly taken into account for selenium and tellurium. Chemical analysis of the LBE used in the inventor's experiments show that the concentration of these elements are below the detection limits (<2 ppm, ICP-OES), but still these elements can be present as inactive impurities with much higher concentrations than those of the radioactive tracer determined by γ-ray spectroscopy. Polonium however is present in the inventor's samples in a carrier-free state. Therefore, the formation of Po2 is very unlikely. Considering the approximate character of the inventor's calculations the evaporation in form of BiQ molecules is possible as well. In particular, relatively small values for the enthalpies of evaporation of BiQ have been calculated for Q=Se and Po. Thus, no certain decision can be made based on the results of the inventor's calculations. Finally, it is also possible that the release process for selenium, tellurium and polonium is not identical. Definitely, the evaporation in form of BiQ molecules is much more likely than evaporation as PbQ. For further clarification of the reaction pathway, concentration dependent evaporation experiments should be performed to investigate Q/Q2-problem. For selenium and tellurium this can be achieved by the addition of inactive chalcogen as a carrier, which also reflects the operating conditions of a LBE spallation target, i.e. higher concentrations of spallation products. Furthermore, larger scale experiments in a flow system with varying gas phase composition and with the addition of suitable representatives for spallation products would be useful to establish a deeper understanding of the processes occurring in such a target. The interaction of polonium with other spallation products such as electropositive metals will most likely lead to a decrease of its evaporation rate [Neuhausen, J., Eichler, B.: Extension of Miedema's Macroscopic Atom Model to the Elements of Group 16 (O, S, Se, Te, Po), PSI-Report 03-13, Paul Scherrer Institute, Villigen, Switzerland, September 2003]. Finally, a study of segregation effects of polonium in solid LBE is of interest with respect to the storage of a spallation target after decommissioning. Given the fact that LBE melts at 398 K relatively high diffusion rates can be expected within the target material after freezing and decommissioning. Results of calculations of approximate partial molar enthalpies of segregation of polonium in lead and bismuth [Neuhausen, J., Eichler, B.: Extension of Miedema's Macroscopic Atom Model to the Elements of Group 16 (O, S, Se, Te, Po), PSI-Report 03-13, Paul Scherrer Institute, Villigen, Switzerland, September 2003] indicate that a segregation of chalcogens in solid lead and bismuth is not highly probable, but cannot be ruled out as well. Indeed, in the inventor's evaporation experiments the inventors have observed small segregation effects for the selenium samples that manifested themselves in the count rates of the lowest energy γ-lines (as a consequence, these lines were excluded from release evaluations). 2. Preferred Aspect: Volatile Elements Production Rates in a 1.4 GeV Proton-Irradiated Molten Pb—Bi Target In a second preferred aspect of the present invention, the invention relates to volatile elements production rates in a 1.4 GeV proton-irradiated molten lead-bismuth target a. Introduction Production rates of volatile elements following spallation reaction of 1.4 GeV protons on a liquid Pb/Bi target have been measured. The experiment was performed at the ISOLDE facility at CERN. These data are of interest for the developments of targets for accelerator driven systems such as MEGAPIE. Additional data have been taken on a liquid Pb target. Calculations were performed using the FLUKA and MCNPX Monte Carlo codes coupled with the evolution codes ORIHET3 and FISPACT using different options for the intra-nuclear cascades and evaporation models. Preliminary results from the data analysis show good comparison with calculations for Hg and for noble gases. For other elements such as I it is apparent that only a fraction of the produced isotopes is released. The agreement with the experimental data varies depending on the model combination used. The best results are obtained using MCNPX with the INCL4/ABLA models and with FLUKA. Discrepancies are found for some isotopes produced by fission using the MCNPX with the Bertini intranuclear cascade model coupled with the Dresner evaporation model. In the development of key experiments in the frame of the research on Accelerator Driven System (ADS) for the nuclear waste transmutation (The European Technical Working Group on ADS, A European Roadmap for Developing Accelerator Driven Systems (ADS) for Nuclear Waste Incineration, ENEA, Roma, 2001), many issues arise which require dedicated experiments. One example is the development of an ADS target, where the isotope production following the interaction of an intense proton beam with a liquid target is of fundamental importance for safety reasons. In the European roadmap for developing accelerator driven systems for nuclear waste incineration, the key experiment for the target development is MEGAPIE (G. S. Bauer, et al., Journal of Nuclear Materials 296, 17 (2001).). The aim of the MEGAPIE (MEGAWatt PIlot Experiment) project is to demonstrate the feasibility of a liquid lead bismuth eutectic (LBE) target for spallation facilities at a beam power level of about 1 MW. During the design phase of such an innovative system, many safety aspects must be considered. One of them concerns the production of volatile elements during operation. This is important for several reasons: i) some stable gases, and in particular 4He and H, are expected to be produced in relatively large quantity (in the case of MEGAPIE, about 1 liter NTP per month) and a system must be designed to handle safely the gases and avoid excessive pressure buildups. Moreover, it is important to know the production of these light elements to estimate possible damage to structural materials. ii) the production of radioactive elements is of concern for safety reason. The long-lived elements are of major concern, but short-lived elements are also of interest in case of an accident. In the last years a great research effort was devoted in basic nuclear studies of interest for ADS (accelerator driven systems) applications. This has resulted in a renovated interest in the study of isotope production following spallation reactions in heavy materials (Yu. E. Titarenko et al., Phys. Rev. C 65, 064610 (2002), R. Michel et al., Nucl Instrum. Methods B 129, 153 (1997).). Experiments performed in inverse kinematics have allowed the investigation over large mass regions of production cross sections in thin targets (T. Enqvist et al., Nucl. Phys. A 686, 481 (2001).). These experiments, in combination with the further development of Monte Carlo transport codes, have led to a deeper understanding of the spallation process and to the development of new theoretical models (A. Boudard et al., Phys. Rev. C 66, 044615 (2002).). In the case of an ADS target, where production of isotopes originates in a large volume of LBE, it is important to consider not only the production of volatile elements, but also their release rate out of the LBE volume. In the case of MEGAPIE, a cover gas system has been designed to properly handle the gas produced (W. Wagner et al., in Proceedings of the MEGAPIE Technical Review Meeting, Nantes, France 2004). A verification of the production rates estimated by the codes used during the design of the cover gas system is therefore important. The inventors chose to perform a dedicated experiment to study the production rates of stable and radioactive volatile elements in a LBE target irradiated by a proton beam of the energy of the order of the energy of the SINQ synchrotron (590 MeV). For a suitable experimental setup, see Example 2. A selection of the data is presented in this invention, with emphasis on the γ-spectroscopy data. Online Measurements The time-dependent releases of volatile elements were measured with the online measuring techniques of the tape station and the Faraday cup. Release curves of volatile elements have specific shapes typical for each element; in most of the cases the decay part can be fitted with the sum of two exponentials (U. Köster, Ausbeuten und Spektroskopie radioaktiver Isotope bei LOHENGRIN und ISOLDE, PhD thesis, Technische Universität München, and references therein (2000).). The online measurement with the tape station allows correction for partial decay of produced isotopes inside the target, before the release. In fact, since the release is dependent on the chemical properties of a given element, it is possible for instance to fit the release functions of 6He (measured with the tape station) and 4He (measured with the Faraday cup) and correct for the partial decay of the 6He. During the first measurement, with the LBE target, it was found that the short term component exhibited discontinuities probably related to splashing effects in the target which reduced for a few tens of ms the ionization efficiency of the ion source. While this affects only slightly the absolute release, which is dominated by the long component, it makes it more difficult to fit the release curve. No such effect was observed during the second measurement, with the Pb target, where the proton beam intensity was reduced to 1.5×1012 protons/pulse. In FIG. 9 the 4He current measured by a Faraday cup for 6 s after the arrival of the proton beam on the Pb target is shown. The ionization and transmission efficiency from the ion source to the Faraday cup was measured to be 0.05% for 3He. Assuming the same transmission efficiency for 4He, the production rate for 4He is 0.77 atoms/p, with a systematic uncertainty of about 20%. This value is in good agreement with calculations with MCNPX with the Bertini/Dresner models, giving 0.84 atoms/p. Offline Measurements Collection measurements were performed for a number of isotopes. The inventors investigated the release of Ne, Ar, Kr, Xe, Br, Cd, Te, I, Hg, Po, and At radioisotopes. During the first measurement run more attention was concentrated on those isotopes which are critical for the operation of an ADS target such as MEGAPIE. For a given isotope, the measured yield has two components, one from direct production from the target and one from the decay of parents. Isotopes were collected in an order chosen so that the first ones to be measured were the first reaching equilibrium, having parents with shorter half-lives. In this way most of the measured isotopes were in equilibrium with their parents, with only a few exceptions. In FIG. 10 the measured cumulative production rates for radioactive Hg isotopes are presented. Longer-lived Hg isotopes are expected to be completely released at the temperature of 600° C. The ionization efficiency was not measured for Hg, as it was only measured for noble gases. In this case only indicative results can be extracted: based on previous results from R. Kirchner (Nucl. Instrum. Methods B 126, 125 (1996)), the inventors considered an efficiency of a factor 1.5 higher than the measured Xe efficiency of 3.7(11) %. The measured values are in line with expected cumulative production rates calculated using the Monte Carlo transport codes FLUKA (A. Fassò et al., in Proceedings of the Monte Carlo 2000 conference, Lisbon, A. Kling, F. Barao, M. Nakagawa, L. Tavora, P. Vaz eds., Sprinter-Verlag Berlin, p. 159 (2001)) and MCNPX (L. S. Waters et al., MCNPX Users's Manual Version 2.4.0, LA-CP-02-408 (2002).). The two codes were coupled with the evolution codes ORIHET3 (F. Atchison and H. Schaal, Orihet 3—Version 1.12, A guide for users, March 2001) and SP-FISPACT (C. Petrovich, SP-FISPACT, A computer code for activation and decay calculations for intermediate energies. A connection of FISPACT and MCNPX, RT/ERG/2001/10, ENEA, Bologna (2001).), respectively. In the case of MCNPX, results are shown here with two different model combinations for the intranuclear cascade and evaporation models. The circles represent results from using the Bertini intranuclear cascade model with the Dresner evaporation code. The diamonds are obtained using the recently implemented INCL4/ABLA (A. Boudard et al., Phys. Rev. C 66, 044615 (2002).) model combination. The trend observed in the data as a function of the atomic mass is well reproduced by the three calculations. One should note that for 193Hg, 195Hg and 197Hg, there are isomeric states of 11.1 h, 40 h and 23.8 h half-lives, respectively. For these three isotopes, equilibrium was not achieved between formation and decay of the respective isomeric states, a process which is difficult to properly calculate with existing Monte Carlo codes. Overall these results confirm the expected production rates of Hg isotopes in a thick LBE target. Results for Xe isotopes, also measured with the LBE target at T=600° C., are shown in FIG. 11. In this case there is a clear disagreement between the values calculated with MCNPX with Bertini/Dresner, and the results from the other two calculations. The data, with an ionization efficiency of 3.7% for Xe isotopes seem to favor the other two calculation results, thus confirming recent experimental findings (T. Enqvist et al., Nucl. Phys. A 686, 481 (2001).). Similar results are obtained for the iodine isotopes. However, I is not completely released and observed production rates at 600° C. are a factor 10 lower than the calculated FLUKA and MCNPX (INCL4/ABLA) values. While production of Hg isotopes from Pb/Bi target is due to direct spallation, the Xe and I isotopes are the results from a later stage of the spallation process, the fission of highly excited spallation fragments, or as a two-step process due to neutron induced fission from high energy spallation neutrons. Thus the evaporation models, the Dresner and ABLA, are probably most responsible for the differences observed in the calculations. Among the other isotopes measured, it is of particular interest to discuss the Po and At. Production rates of 207,208,209,210At of the order of 107 atoms/μC (assuming the same ionization efficiency as for Hg) were detected, with values an order of magnitude lower for 206At. Such production rates are not of concern for an ADS. On the other hand, it is the first observation of At beam from a Pb/Bi target and this constitutes an interesting result with possible applications. Production of At comes from several possible reactions of Bi, but the most likely, given the high proton energy, is 209Bi(p, π xn)210-xAt. The At decay is responsible for the observed small quantities of Po isotopes, which contrary to At is expected to be produced in large amounts. However, as found in Ref. 15, little or no Po should be released at 600° C. Of the other isotopes measured in the first measurement, no release of Br was observed, while very little amounts of Cd isotopes were detected. For the Kr isotopes, some problems during the measurement rendered the analysis questionable and such measurement was repeated with the Pb target. b. Conclusions The first results from the measurements of production rates of volatile elements from irradiated LBE and Pb targets indicate that the results are consistent with the expectations from Monte Carlo calculations. Overall, these preliminary results confirm the expected production rates in an ADS target such as MEGAPIE, and therefore help in positively assessing such calculations, and the system designed to handle the released volatile elements. 3. Preferred Aspect: In Vivo TAT Application using 149Tb-Rituximab In a third preferred aspect of the present invention, the invention relates to targeted alpha therapy (TAT) in vivo, showing direct evidence for single cancer cell kill using 149Th-Rituximab. a. Introduction This part of the present invention demonstrates high efficiency sterilization of single cancer cells in a SCID mouse model of leukemia using Rituximab, a monoclonal antibody that targets CD20, labeled with 149-Terbium, an alpha-emitting radioisotope. Radioimmunotherapy with 5.5 MBq labeled antibody conjugate (1.11 GBq/mg) 2 days after an intravenous graft of 5·106 Daudi cells resulted in tumor free survival for >120 days in 89% of treated animals. In contrast, all control mice (no treatment or treated with 5 and 300 μg unlabeled Rituximab) developed lymphoma disease. At the end of the study period, 28.4±4% of the long-lived daughter activity remained in the body, out of which 91.1% was located in bone tissue and 6.3% in the liver. A relatively high daughter radioactivity concentration was found in the spleen (12±2%/g), suggesting that the killed cancer cells are mainly eliminated through the spleen. This promising preliminary in vivo study suggests that TAT with 149Tb is worthy of consideration as a new generation radioimmunotherapeutic approach. Single cancer cells in circulation and small cancer cell clusters can be effectively targeted with radio-immunoconjugates that specifically bind to the cells and deliver the required dose. Alpha-emitting radioisotopes may be of great advantage in this kind of therapy because of their higher linear energy transfer (LET) value and consequently, the shorter penetration track compared to β−- and γ-radiation [Hall E J. Radiobiology for the Radiologist. 4th ed. Philadelphia: Lippincott J B Comp 1994]. It has been shown that only a very few alpha-hits are sufficient to kill a cell [Maecklis R M, Lin J Y, Beresford B, Achter R W, Hines J J, Humm J L. Cellular kinetics, dosimetry, and radiobiology of alpha-particle radioimmunotherapy: inducing of apoptosis. Radiat Res. 1992; 130:220-226], and the short range of the alpha particles increases the safety profile of alpha-emitters because nonspecific irradiation of normal tissue (or plasma) around the target cells is greatly reduced or absent [McDevitt M R, Ma D, Simon J, Frank R K and Scheinberg D. Design of 225Ac-radiopharmaceuticals. Appl Rad Isot. 2002; 57:841-847]. Additionally, since single cancer cells in circulation are immediately accessible to the injected (i.v.) radioimmunoconjugate, the shorter half-lives of a few α-emitting radioisotopes may be advantageous [Allen B J, Blagojevic N. Alpha and beta emitting radiolanthanides in targeted cancer therapy: the potential role for Terbium-149. Nucl Med Commun 1996; 17:40-47]. Only few R-emitting radionuclides fulfill the requirements for this specific nuclear medical application: 255Fm, 225Ac, 224Ra, 223Ra, 213Bi, 212Bi, 211At and 149Tb. Especially the 213Bi and 211At have proven to be very promising candidates, because of the availability (225Ac/213Bi generator) and the convenient half-life of 7.2 h (211At) (for example see Zalutsky M R, Vaidyanathan G. Astatine-211 labeled radiotherapeuticals: an emerging approach to targeted alpha particle therapy. Current Pharm Design 2000; 6:1433-1455, Huber R, Seidl C, Schmid E, Seidenschwang S, Becker K-F, Schuhmacher C, Apostolidis C, Nikula T, Kremmer E, Schwaiger M, Senekowitsch-Schmidtke. Locoregional alpha-radioimmunotherapy of intraperitoneal tumor celldissemination using a tumor-specific monoclonal antibody. Clinical Cancer Research 2003; 9:3922-3928]). Today, new approaches in conjugation with chelating ligands allow the stable labeling of macromolecules (such as monoclonal antibodies) or peptides with metallic radionuclides. The first clinical proof-of-principle of targeted alpha therapy was observed using the HuM195 antibody labeled with the short-lived (46 min) 213Bi radionuclide [Jurcic J G, Larson S M, Sgouros G, McDevitt M R, Finn R D, Divgi C R, Ballangrud Å M, Hamacher K A, Ma D, Humm J L, Brechbiel M W, Molinet R, Scheinberg D A. Targeted alpha-particle immunotherapy for myeloid leukemia. BLOOD 2002; 100:1233-1239], which is a daughter product in the decay chain of 225Ac (10 d). The mother nuclide, 225Ac, is itself considered as candidate for TAT, and corresponding studies are ongoing [McDevitt M R, Ma D, Simon J, Frank R K and Scheinberg D. Design of 225Ac-radiopharmaceuticals. Appl Rad Isot. 2002; 57:841-847, McDevitt M R, Sgouros G, Finn R D, Humm J L, Jurcic J G, Larson S M, Scheinberg D A. Radioimmunotherapy with alpha-emitting radionuclides. Eur J Nucl Med 1998; 25:1341-1351]. A potential drawback with use of 225Ac is the possibility that the short-lived alpha-emitting daughter nuclides in the decay chain will escape from the place of origin, leading to uncontrolled deposition of the radiation dose throughout the body. The partial alpha-emitting nuclide 149Th (T1/2=4.118 h, Eα=3967 keV, range in tissue=28 μm), which belongs to the group of rare earth elements, has been proposed as a promising alpha-emitter for targeted alpha therapy (TAT) [Allen B J, Blagojevic N. Alpha and beta emitting radiolanthanides in targeted cancer therapy: the potential role for Terbium-149. Nucl Med Commun 1996; 17:40-47, Allen B J, Goozee G, Imam S, Sarkar S, Leigh J, Beyer G-J. Targeted cancer therapy: The potential role of terbium-149. 6th International Conference on Radiopharmaceutical Dosimetry, Gatlington, Tenn. (USA), May 7-10, 1996, CERN-PPE/96-127, 1996; Charlton D E, Utteridge T D, Allen B J. Theoretical treatment of human hemopoietic stem cell survival following irradiation by alpha particles. Int J Radiat Biol 1998; 74:111-118; Allen B J. Can alpha immunotherapy succeed where other modalities have failed? Nucl Med Commun 1999; 20:205-207]. Its chemical behaviour is very close to that of yttrium or lutetium, whose isotopes 90Y and 177Lu are currently the most predominant metallic radionuclides used in clinical radioimmunotherapy (RIT) [Wagner Jr H N, Wiseman G A. et al. Administration Guidelines for Radioimmunotherapy of Non-Hodgkin's Lymphoma with 90Y-Labeled Anti-CD20 Monoclonal Antibody. J Nucl Med 2002; 43:267-272]. Thus, existing approaches for labelling of chelated bioconjugates with these metallic radionuclides, as well as 166Ho, 153Sm, 213Bi or 225Ac, can be directly applied to 149Th. Previous in vitro studies have revealed certain advantages of 149Tb over 213Bi for treating single cells [Miederer M, Seidl C, Beyer G-J, Charlton D E, Vranje{hacek over (s)}-Durić S D, {hacek over (C)}omor J J, Huber R, Nikula T, Apostolidis C, Schuhmacher C, Becker K-F, Senekowitsch-Schmidtke R. Comparison of the radiotoxicity of two alpha emitting immunoconjugates Terbium-149 and Bismuth-213 directed against a tumor-specific, exon 9 deleted (d9) E-cadherine adhesion protein. Radiation Research 2002; 159:612-620]. These advantages, which relate to the lower energy and higher LET of α-particles emitted by 149Tb, partially compensate for its lower alpha branching (17%, FIG. 12) [Vranje{hacek over (s)} S D, Miederer M, {hacek over (C)}omor J J, Soloviev D, Beyer G-J and the ISOLDE collaboartion. Labeling of monoclonal antibodies with 149-Tb for targeted alpha therapy. J Lab Comp Radiopharm 2001; 44:718-720, Miederer M, Seidl C, Beyer G-J, Charlton D E, Vranje{hacek over (s)}-Durić S D, {hacek over (C)}omor J J, Huber R, Nikula T, Apostolidis C, Schuhmacher C, Becker K-F, Senekowitsch-Schmidtke R. Comparison of the radiotoxicity of two alpha emitting immunoconjugates Terbium-149 and Bismuth-213 directed against a tumor-specific, exon 9 deleted (d9) E-cadherine adhesion protein. Radiation Research 2002; 159:612-620]. The longer half-life of 149Tb (4.12 h) compared to the 213Bi (46 min) represents a clear advantage, both at the level of bioconjugate preparation and administration to patients for tumor cell targeting. On the other hand, the fate of long-lived daughter products that appear during the decay of 149Tb would need to be considered carefully in the dosimetry (see FIG. 12.). In this part of the present invention the inventors describe the first in vivo survival study using a 149Tb-based TAT approach in SCID (severe combined immuno-deficient) mice. SCID mice, being deficient in T and B cell immune defense, easily develop tumor masses after injection of cancer cells. Daudi cells, which are derived from a human Burkitt lymphoma, are one of several cell lines that can rapidly colonize these mice. Depending on the injection route, different tumor types can develop. As little as 100 injected (i.v.) Daudi cells are sufficient to kill SCID mice due to tumor development [Ghetie M A, Richardson J, Tucker T, Jones D, Uhr J W, Vitetta E S, Disseminated or localized growth of a human B-cell tumor (Daudi) in SCID mice. Int J Cancer 1990; 45:481-485]. Since Daudi cells express a high number of CD2O antigens Rituximab can target Daudi cells with high specificity. Thus, an early stage of this model, within three days of i.v. xenograft, before the formation of manifested tumor nodes, provides an ideal system to study the proposed advantages of 149Tb-based TAT. The primary aim of this work was to examine the efficacy of 149Tb-labeled Rituximab to specifically kill circulating single cancer cells or small cell clusters in vivo. SCID mice following i.v. xenograft with Daudi cells represent a perfect model for leukemia [McDevitt M R, Ma D, Lai L T, Simon J, Borchardt P, Frank R K, Wu K, Pellegrini V, Curcio M J, Miederer M, Bander N H, Scheinberg D A. Tumor therapy with targeted atomic nanogenerators. Science 2001; 294 (5546):1537-40]. The inventor's experimental model involves TAT intervention within three days of i.v. xenograft, and hence before the formation of manifested tumor nodes, which the inventors did not intend to target in this study. According to the inventor's experimental hypothesis, mice xenotransplanted with a lethal number of Daudi cells will survive provided that a sufficient dose of 149Th was delivered via Rituximab to all tumor cells. Secondly, the inventors aimed to obtain information about the behavior of the daughter products generally formed in the radioactive decay chain. The 17% alpha decay mode of the 149Tb generates an isobar chain with the mass number A=145 with 145Eu (T1/2=5.93 d), 145Sm (T1/2=340 d) and 145Pm (T1/2=17.7 a). The EC-process decay chain of 149Tb forms the stable 149Sm passing the 149Gd (T1/2=9.25 d) and the 149Eu (T1/2=93.1d) ([Firestone R B. Table of Isotopes. Eight Edition, New York: Wiley-Interscience, 1996], see FIG. 12). Most of these isotopes are easily detectable using high-resolution gamma spectroscopic techniques (see FIG. 12). In particular the inventors expected that differences in daughter isotope behaviour induced by the different decay modes (alpha versus EC) would be apparent. For suitable materials and methods as well as results obtained applying them, see e.g. Example 3. b. Discussion Protection of Mice Treated with Labeled Rituximab Here the inventors show that TAT with Rituximab labeled with the high purity alpha emitting radio-lanthanide 149Tb led to almost complete protection of xenografted mice over four months without detectable signs of toxicity, under conditions where all animals in the control groups had to be sacrificed during the observation period due to the development of tumor diseases. The efficacy of the radionuclide bioconjugate as opposed to the unconjugated tumor targeting antibody alone is underlined by the complete lack of protection in the control group which received 5 μg unlabeled Rituximab per animal, and the relatively poor protection afforded by the higher dose unlabeled Rituximab group (300 μg per animal). The degree of protection afforded by the 149Tb-labeled Rituximab indicates that TAT with 149Tb is, on the basis of its efficacy, worthy of further consideration as a novel radioimmunotherapeutic strategy. Biodistribution of Label and Decay Products From earlier studies the inventors have learnt that once the lanthanides are trapped in a tissue, like liver or bone, they are fixed quite stably [Beyer G-J, Offord R E, Künzi G, Jones R M L, Ravn U, Aleksandrova Y, Werlen R C, Mäcke H, Lindroos M, Jahn S, Tengblad O and the ISOLDE Collaboration. Biokinetics of monoclonal antibodies labeled with radio-lanthanides and 225-Ac in xenografted nude mice. J Label Conzpd Radiopharm 1995; 37:229-530, Beyer G-J, Münze R, Fromm W D, Franke W G, Henke E, Khalkin V A, Lebedev N A. Spallation produced 167-Tm for medical application. In: Medical Radionuclide Imaging 1980, Vienna: IAEA, 1981, Vol. 1, p. 587 (IAEA-SM-247/60) 1981]. The blood clearance for free radio-lanthanides or radiolanthanides injected in solutions containing chelate ligands (citrate, EDTMP, NTA, EDTA, DTPA and others) is fast (half-time <1 h) [Beyer G-J, Münze R, Fromm W D, Franke W G, Henke E, Khalkin V A, Lebedev N A. Spallation produced 167-Tm for medical application. In: Medical Radionuclide Imaging 1980, Vienna: IAEA, 1981, Vol. 1, p. 587 (IAEA-SM-247/60) 1981, Beyer G-J, Offord R, Künzi G, Aleksandrova Y, Ravn U, Jahn S, Backe J, Tengblad O, Lindroos M and the ISOLDE Collaboration. The influence of EDTMP-concentration on the biodistribution of radio-lanthanides and 225Ac in tumor bearing mice. Nuclear Medicine and Biology 1997; 24:367-372]. The radio-lanthanides are then present mainly in the bone matrix and the liver, with the liver uptake determined by the ionic radius of the lanthanide [Beyer G-J, Münze R, Fromm W D, Franke W G, Henke E, Khalkin V A, Lebedev N A. Spallation produced 167-Tm for medical application. In: Medical Radionuclide Imaging 1980, Vienna: IAEA, 1981, Vol. 1, p. 587 (IAEA-SM-247/60) 1981, Beyer G-J, Offord R, Künzi G, Aleksandrova Y, Ravn U, Jahn S, Backe J, Tengblad O, Lindroos M and the ISOLDE Collaboration. The influence of EDTMP-concentration on the biodistribution of radio-lanthanides and 225Ac in tumor bearing mice. Nuclear Medicine and Biology 1997; 24:367-372, Beyer G-J, Bergmann R, Schomäcker K, Rösch F, Schafer G, Kulikov E V, Novgorodov A F. Comparison of the Biodistribution of 225Ac and Radiolanthanides as Citrate Complexes. Isotopenpraxis 1990; 26:111-114]. In the case of macromolecules (like monoclonal antibodies) the blood clearance is slow (half-time ˜1 day) [Beyer G-J, Offord R E, Künzi G, Jones R M L, Ravn U. Aleksandrova Y, Werlen R C, Mäcke H, Lindroos M, Jahn S, Tengblad O and the ISOLDE Collaboration. Biokinetics of monoclonal antibodies labeled with radio-lanthanides and 225-Ac in xenografted nude mice. J Label Compd Radiopharm 1995; 37:229-530]. Thus, most of the 149Tb will decay while the labeled bioconjugate is in circulation and the free daughter nuclides formed in the radioactive decay would follow the biodistribution known for free radiolanthanides. The biodistribution found 120 days after treatment corresponds to the distribution patterns known for the free radio-lanthanide: highest daughter nuclide accumulation in bone and liver (91.1% and 6.3% of the retained activity, respectively) [Beyer G-J, Münze R, Fromm W D, Franke W G, Henke E, Khalkin V A, Lebedev N A. Spallation produced 167-Tm for medical application. In: Medical Radionuclide Imaging 1980, Vienna: IAEA, 1981, Vol. 1, p. 587 (IAEA-SM-247/60) 1981, Beyer G-J, Offord R, Künzi G, Aleksandrova Y, Ravn U, Jahn S, Backe J, Tengblad O, Lindroos M and the ISOLDE Collaboration. The influence of EDTMP-concentration on the biodistribution of radio-lanthanides and 225Ac in tumor bearing mice. Nuclear Medicine and Biology 1997; 24:367-372]. The spleen shows a radioactivity concentration almost as high as bone and significantly higher compared to liver. The inventors interpret this result as evidence that the targeted and killed cancer cells are eliminated mainly through the spleen, where the remaining radioactive daughter atoms are then trapped. The long-lived daughter products are formed along two main decay processes: the isobar chain with A=145 is generated via the alpha decay mode of the initial 149Th, while the isobar chain with A=149 is formed after the EC- or β+-process. In case of an alpha decay the recoil energy of the 145Eu daughter nuclei (110 keV) exceeds significantly the chemical binding energy. Consequently, the original molecule, the antibody-construct, is destroyed and the daughter atom is initially stabilized as free Eu3+ ion. In the case of the EC-decay mode, the bound rupture is induced due to the Auger electron emission forming free daughter species [Beyer G-J, Herrmann E and Khalkin V A. Chemical effects related to different radioctive decay processes of cerium isotopes chelated with different polyaminocarbonic acids Dubna: JINR P 12-7758, 1974. Beyer G-J, Hermann E. Chemical effects of nuclear transformations in lanthanide chelate complexes., in Proceedings of the COST Chemistry Action D18, Mid Term Evaluation Workshop on Lanthanide Chemistry for Diagnosis and Therapy, Heidelberg (Germany) Jul. 22-25, 2002, p. 26] with 100% efficiency. However, it cannot be assumed that the daughter species escapes from its place of origin; it could eventually be bound to other proteins in the immediate environment. Consequently, one may not necessarily expect identical behavior from daughter products generated in the two different pathways: alpha- or EC-process. Analysis of the γ-spectroscopic data revealed that there was no statistically significant difference in the ratio of retained 145Sm to 149Eu in the organs from that predicted by the branching ratio of 149Tb. Thus, the radioactive decay pathway does not influence the biodistribution or redistribution of the long-lived daughter lanthanides. Extrapolation to Clinical Application A preliminary dose estimation for patients injected with 5 GBq 149Tb-Rituximab was performed based on MIRDOSE 3.1 [Stabin M G. MIRDOSE: personal computer software for internal dose assessment in nuclear medicine. J Nucl Med 1996; 37:538-546]. Assuming total decay of 149Tb-Rituximab in circulation and 100% retention of daughter nuclides in the body with a bone uptake of 91%, the total equivalent dose to the bone marrow as the critical organ would be 540 mSv/5 GBq (108 mSv/GBq) (see also Table 4). 149Tb itself would contribute 66.7% of the bone marrow radiation dose (45.2% due to the alpha-radiation using an alpha-radiation weight factor of WR=10) and 21.5% due to its gamma- and beta+-radiation) while the daughter nuclides would contribute 33.3% only. The dose contribution from daughter nuclides estimated in this way represents a worst case estimation (assuming 100% retention), since only 28.4% of the long-lived daughter products were retained in mice 120 days p.i. Thus, injection of a potentially therapeutic activity, 5 GBq 149Tb-Rituximab in a 70 kg patient, would deliver a bone marrow radiation dose far below the critical level. This preliminary dose estimation is well compatible with considerations presented in the review by McDevitt et al. [McDevitt M R, Sgouros G, Finn R D, Humm J L, Jurcic J G, Larson S M, Scheinberg D A. Radioimmunotherapy with alpha-emitting radionuclides. Eur J Nucl Med 1998; 25:1341-1351]. For further reduction of the retention of the daughter nuclides one could apply single or multiple injections of chelating ligands like EDTA or DTPA during or shortly after the treatment. This approach is already practiced as a preventive action in treatments with 90Y- or 177Lu-DOTATOC [Beyer G-J, Ruth T J. The role of electromagnetic separators in the production of radiotracers for bio-medical research and nuclear medical applications. NIM B 2003; 204:694-700]. Time Constraints and Availability Spallation reaction in combination with on line mass separator technology was used for the production of 149Tb for this study. The radiochemical separation and purification of the 149Th was relatively easy to perform in about 30 minutes in this specific case, since the inventors started from non-carrier added preparations. The final 149Tb preparation was obtained in very high purity and in a small volume, the labeling of the bioconjugate was fast (10 minutes) and almost quantitative. The administration of the preparation should be carried out as rapidly as possible after purification of the 149Tb, since levels of contamination with daughter nuclides will increase with time. For example, application of a fixed dose of the 149Tb-labeled bioconjugate 4 hours after the final purification of the isotope itself (EOS) leads to an increase of the long-lived daughter content by a factor of 2. According to the preliminary dose estimation one could define a shelf-life for the labelled 149Tb-labeled bioconjugate of about 4-6 hours. For a longer delay it would be advisable to repurify the 149Tb from the accumulated daughter products, a process that could be expected to require 30 minutes. Several nuclear processes could be used to make this interesting alpha emitting isotope available on large scale: light particle (p, d, He) induced reactions on 152Gd as target material, heavy ion induced reactions on light lanthanide targets or spallation reaction on Ta as target [Beyer G-J, {hacek over (C)}omor J J, Daković M, Soloviev D, Tamburella C, Hagebø E, Allan B, Dmitriev S N, Zaitseva N G, Starodub G Y, Molokanova L G, Vranje{hacek over (s)} S D, Miederer M and the ISOLDE Collaboration. Production routes of the alpha emitting 149-Tb for medical application. Radiochim Acta 2002; 90:247-252]. Off line and on line mass separation process may support a very high isotopic purity [Beyer G-J, {hacek over (C)}omor J J, Daković M, Soloviev D, Tamburella C, Hagebø E, Allan B, Dmitriev S N, Zaitseva N G, Starodub G Y, Molokanova L G, Vranje{hacek over (s)} S D, Miederer M and the ISOLDE Collaboration. Production routes of the alpha emitting 149-Tb for medical application. Radiochim Acta 2002; 90:247-252, Beyer G-J, Ruth T J. The role of electromagnetic separators in the production of radiotracers for bio-medical research and nuclear medical applications. NIM B 2003; 204:694-700]. All the above-mentioned technologies are well-developed and available today. In summary, should 149Tb continue to show promise in further studies of TAT, then it would be technically feasible to make the isotope available in large-scale and on a routine basis. This experimental setup is e.g. suitable for the first preferred aspect of this invention. Pieces of eutectic Pb/Bi-alloy (44.8 wt. % Pb, 55.2 wt. % Bi, Impag AG, Switzerland, impurities (ppm): Ag 11.4, Fe 0.78, Ni 0.42, Sn 13.3, Cd 2.89, Al 0.3, Cu 9.8, Zn 0.2, Se<2, Te<2) of dimensions approx. 10×10×1.5 mm3 have been doped with 75Se, 121Te and 206Po by implantation of mass-separated radioactive ion beams at the on-line isotope separator ISOLDE at CERN. 206Po was prepared indirectly, by implantation of the precursors 206Rn (T1/2=2.7 min) and 210Fr (T1/2=3.2 min) respectively. The 206Rn beams were produced by 1.4 GeV proton-induced spallation of a 50 g/cm2 238UCx target (x≈4) connected via a water-cooled transfer line to a FEBIAD ion source [U. Köster for the ISOLDE Collaboration: ISOLDE target and ion source chemistry. Radiochimica Acta 89, 749 (2001).]. The condensation of non-volatile isobars in the transfer line assures beams of high isotopic purity (>>99.9%). About 38% [Audi, G., Bersillon, O., Blachot, J. Wapstra, A. H.: The NUBASE evaluation of nuclear and decay properties. Nuclear Physics A 729, 3 (2003).] of the 206Rn decays via (β+/EC)→206At→(β+/EC) to 206Po, the remaining 62% populate 202gPb and 198Pb/198Tl, which do not contribute any measurable activity after some days of decay. The beam intensity was about 2·108 206Rn+ ions per s, allowing to collect 4 kBq 206Po per minute. On another occasion a 50 g/cm2 238UCx target connected via a high temperature transfer line to a tungsten surface ionizer was used. All parts were kept above 2000° C. About 98% of the 210Fr decaus via EC/β+→210Rn→α→ or via α→206At→EC/β+→ to 206Po. Again the side branches of the decay chain do not contribute any measurable activity after some days of decay. The beam intensity of 210Fr of about 2·108 ions per s results in a production of 10 kBq 206Po per minute. Also 121Te was produced indirectly by implantation of the precursors 121g+mCs which decay by β+/EC via 121Xe and 121I to 121Te. 121Cs was produced from the same UCx target as above by 1.4 GeV proton-induced spallation-fission and then surface ionised. Despite the unfavourable target and ion source combination (neutron-deficient nuclides are much better produced by spallation of a close-by target nucleus), a 121Cs beam intensity better than 3·107 ions per s allowed to collect about 1 kBq 121Te per minute. 75Se was produced by 1.4 GeV proton-induced spallation of a 11 g/cm2 zirconia fibre target connected via an unselective, hot transfer line to a FEBIAD ion source [Köster, U., Bergmann, U. C., Carminati, D., Catherall, R., Cederkäll, J., Correia, J. G., Crepieux, B., Dietrich, M., Elder, K., Fedoseyev, V. N., Fraile, L., Franchoo, S., Fynbo, H., Georg, U., Giles, T., Joinet, A., Jonsson, O. C., Kirchner, R., Lau, Ch., Lettry, J., Maier, H. J., Mishin, V. I., Oinonen, M., Peräjärvi, K., Ravn, H. L., Rinaldi, T., Santana-Leitner, M., Wahl, U., Weissman, L.: Oxide fiber targets at ISOLDE. Nucl. Instr. Methods B 204 (2003) 303]. The cumulative ion beam intensity of 75Se+ plus precursors (75Br, 75Kr, 75Rb) was about 5.108 ions per s, allowing to collect 2 kBq of 75Se within 1 minute. The samples doped with 75Se, 121Te and 206Po were cut in pieces and afterwards melted and heated at 673 K for 1 hour together with additional LBE reduced under a hydrogen atmosphere to achieve homogeneous distribution of radionuclides as well as suitable sample sizes and activities suitable for measurement by γ-ray spectroscopy. No additional carrier was added. For the long-term release studies LBE containing 205Bi produced by neutron activation was used for diluting the samples in the same manner as described above. 205Bi was used as an internal standard for the evaluation of γ-ray spectra to correct for changes in sample shape frequently occurring on melting. For short-term experiments 206Bi produced by decay of 206Po was used as standard. The number of nuclei and concentrations of 75Se, 121Te and 206Po were determined from the peak areas of characteristic γ-rays of the respective nuclide (75Se: 400.66 keV, 121Te: 573.14 keV, 206Po: 1032.26 keV) taking into account the detector efficiency and γ-branching [http://nucleardata.nuclear.lu.se/nucleardata/toi/]. Self-absorption effects were roughly estimated based on sample thickness and mass attenuation coefficients listed in [http://physics.nist.gov/PhysRefData/XrayMassCoef/tab3.html]. Estimated experimental errors of number of nuclei and concentrations are 40% for 75Se, 25% for 121Te and 15% for 206Po resulting mainly from the crude evaluation of self-absorption effects. Typical numbers of nuclei were in the range of 4108 to 5109 for 121Te and 206Po containing LBE samples and 21010 to 51010 for 15Se containing LBE samples. Typical sample masses for short-term (1 hour) experiments were 2.5 g (75Se samples) and 0.13-0.88 g (121Te/206Po samples), whereas for long term studies on the release of 121Te and 206Po larger samples (2.5-7.5 g) have been used. The resulting mole fractions at the start of the experiment were in the range of 310−13 to 2.510−12 for 121Te/206Po containing samples and 3.2 to 7.310−12 for 75Se containing samples. Evaporation experiments (one experiment for each temperature setting) were performed using the experimental set-up illustrated in FIG. 1. Before the experiment, the samples were scratched to remove the surface oxide layer. The samples were then placed on a quartz tissue within a quartz boat, which was placed in a quartz tube. This tube was flushed with an Ar/7%-H2 mixture (purity: H2>99.993%, Ar>99.998%), which was previously run through a column containing a Pd-contact to facilitate the establishment of O2/H2/H2O equilibrium and Sicapent (with indicator, Merck, Germany) for removing moisture. A partial pressure of water of 3.7±1.7 hPa was determined using a Zr/Y2O3-solid electrolyte cell. Some additional experiments were performed in a water saturated Ar atmosphere. For this, Ar (purity >99.998%) was bubbled though a washing bottle containing water at room temperature and the drying column was removed. All experiments were performed using a continuous gas flow of 60 ml/min adjusted by a mass flow controller. The apparatus was flushed for approximately 20 min after the insertion of the sample to remove air contamination. The tube was resistance-heated to the desired temperatures. Temperatures were measured and controlled using thermocouples and a thyristor controller. Two charcoal filters were placed at the end of the tube to prevent volatile radioactive species reaching the exhaust. γ-ray spectroscopic measurements were performed using an HPGe-detector. Short-term experiments: A γ-ray spectrum of the sample was recorded before each heating experiment. The sample was then placed into the evaporation apparatus, which was flushed with the appropriate gas mixture. After approximately 20 min, the apparatus was heated to the desired temperature within 10 min and kept at this temperature for 50 min. Subsequently, the sample was cooled to room temperature within 10 min using a fan. A γ-ray spectrum was recorded after the experiment (typical measuring time: 1 hour). The fractional release of the chalcogens was calculated comparing the integrated peak areas of the following characteristic γ-rays of the respective nuclides (75Se: 264.66, 279.54, and 400.66 keV; 121Te: 507.59 and 573.14 keV; 206Po: 286.41, 311.56, 338.44, 522.47, 980.23 and 1032.26 keV [http://nucleardata.nuclear.lu.se/nucleardata/toi/]) before and after heating. The error bars given in the figures correspond to the standard errors of the mean values obtained by averaging the fractional release calculated for each characteristic γ-ray of the respective nuclide. Given the half-lives of the present nuclides (75Se: 119.8 d; 121Te: 16.8 d; 206Po: 8.8 d [http://nucleardata.nuclear.lu.se/nucleardata/toi/]) a decay correction was omitted for these short-term experiments. However, 206Bi (t1/2=6.24 d [http://nucleardata.nuclear.lu.se/nucleardata/toi/]), which is formed by decay of 206Po, has been used as an internal standard to correct for geometry and self-absorption changes that may occur between the measurements before and after heating due to the melting process. For this purpose, the peak area ratios before and after heating for characteristic γ-rays of the internal standard lying energetically close to characteristic γ-rays of the investigated volatile nuclides were determined and the signals of the volatiles were corrected accordingly. No measurable evaporation of Bi was detected at temperatures below 1280 K. For the three samples heated at temperatures higher than 1280 K a small loss of Bi was observed giving rise to a small underestimation (about 1%) of the respective release values for the chalcogens. Long-term experiments: In principle, the same experimental set-up was used as in the short-term experiments. However, the samples were kept in the apparatus for periods from 10 days up to several weeks with intermittent cooling-measuring-heating-up-cycles as described above. For the evaluation of these measurements a decay correction was applied to the integrated peak areas of the γ-ray signals of both volatile species and internal standard. 206Bi could not be used as an internal standard because it is permanently produced by decay of 206Po. Therefore, 205Bi-containing LBE was used to dilute the samples and this isotope was used as standard. This experimental setup is e.g. suitable for the second preferred aspect of the present invention. The experiment was performed at the ISOLDE facility (E. Kugler, Hyperfine Interactions 129, 23 (2000).). The spallation target consisted of a cylindrical tantalum container filled with liquid LBE. Protons pulses of 1.4 GeV and variable intensity (up to 1013 protons/pulse with a rate of one pulse every 16.8 s) impinged on the target. Following spallation reactions, the produced volatile elements exiting the liquid metal were ionized by means of a plasma ion source, then accelerated to 60 keV and sent to the magnetic mass separators and to the beam lines where the measuring stations were placed. An additional measurement was performed with a liquid Pb target. Yields were measured using three different techniques of common use at ISOLDE. Online yields of stable isotopes and of some radioactive ones were measured by a Faraday cup inserted in the beam line. A special data acquisition system was developed to trigger the current measurement by a picoamperemeter with the arrival of the proton beam on target, thus allowing the measurement of the gas release curves, characteristic of each element. For short-lived β emitting isotopes, beams were directed to a dedicated tape station and yields were measured with a plastic scintillator detector. A third measurement method was used for longer lived (T1/2≧5 min) γ emitting radioisotopes; ion beams were implanted on thin Al foils, then after irradiation an offline γ detection was performed using a calibrated HPGe detector. In order to obtain the absolute production rates from the measured yields, the efficiency of the ion source had to be measured. For this purpose, known amounts of different gas mixtures (consisting of Ar/Xe, He/Ne/Ar/Kr/Xe, and 3He/Ar/Xe mixtures) were leaked into the ion source, thus having the possibility to measure the efficiencies at any time during the experiment. For the LBE target, the measurements were performed with the target at temperatures of 400° C. and 600° C. The Pb target was at a temperature of 520° C. These temperatures are in the range of the LBE temperature in MEGAPIE during operation, which varies from 300° C. to 400° C. depending on the position inside the target. Temperature differences within these ranges are not expected to affect the releases of the noble gases and of the Hg isotopes. On the other hand, differences are expected for some isotopes such as I, Cd and Po. Having performed the experiments at higher temperatures than in MEGAPIE will allow to conclude, in case the release of a specific isotope is not observed at 600° C., that no release should be observed in MEGAPIE for the same isotope at 300-400° C., even for longer irradiation times. A selection of the data is presented in this invention (see the second preferred aspect above), with emphasis on the γ-spectroscopy data. The following Material and Methods are suitable for the third preferred aspect of the present invention. 1. Material and Methods Cell Line: Daudi cells (ATCC Nr. CCL-213) were used to simulate a leukemia model in mice. The cells were cultured in RPMI 1640 medium supplemented with 10% heat-inactivated fetal calf serum and 0.5% penicillin (10000 U/ml)/streptomycin (10 mg/ml) (Sigma-Aldrich). The cell suspension to be injected into mice was prepared by centrifuging the culture for 3 min at 1200 rpm, washing with PBS and re-suspending in PBS at 2.5-107 cells per ml. Antibody Conjugate: Rituximab antibody (Rituxan; IDEC Pharmaceuticals, San Diego, and Genentech Inc, San Francisco) is a chimeric version of anti CD-20 monoclonal antibody consisting of human IgG, constant region and murine variable region. The Rituximab antibody conjugated with SCN-CHX-A-DTPA (2-(4-isothiocyanatobenzyl)-cyclohexyl-dietylenetriamine pentaacetic acid), used in this study, was kindly provided by Dr. D. A. Scheinberg, Memorial Sloan Kettering Cancer Center, New York. Radionuclide: The 149Tb was produced using the on-line isotope separator facility ISOLDE at CERN (Geneva, Switzerland) [Kugler E. The ISOLDE Facility. Hyperfine Interactions 2000; 129:23-42, Beyer G-J, {hacek over (C)}omor J J, Daković M, Soloviev D, Tamburella C, Hagebø E, Allan B, Dmitriev S N, Zaitseva N G, Starodub G Y, Molokanova L G, Vranje{hacek over (s)} S D, Miederer M and the ISOLDE Collaboration. Production routes of the alpha emitting 149-Th for medical application. Radiochim Acta 2002; 90:247-252]. A tantalum-foil target (120 g/cm2) was irradiated with 1.0 or 1.4 GeV protons delivered from the CERN PS-Booster accelerator. The radio-lanthanides generated in the spallation process are released from the target material, which is kept at about 2200° C., ionized by surface ionization and accelerated to 60 keV. From the obtained radioactive ion beams, the A=149 isobars (149Dy, 149Tb and molecular ions 133CeO+ and 133LaO+) were implanted (60 keV) and thus collected in thin layers of KNO3 (10 mg/cm2) on aluminum backings. The 149Tb was separated from its daughters (149Gd and 145Eu) and the pseudo-isobars 133Ce and 133La by cation exchange chromatography using Aminex A5 resin and α-hydroxyisobutyric acid as eluent. A typical elution chromatogram is presented in FIG. 13. The 149Tb-fraction (150-200 μl) was evaporated to dryness and re-dissolved in 50 μl of 100 mM HCl. The final 149Tb concentration was 2 GBq/ml (54 mCi/ml) at end of chromatographic separation (EOS). Labeling Procedure: 25-40 μl of the 149Tb solution characterized above was used immediately for the labeling procedure. The pH was adjusted to 5.5 by adding 60 μl of 3 M CH3COONH4 solution, followed by the addition of 10 μl (40 mg/ml) ascorbic acid. After adding 5 μl of a stock solution of the chelated antibody in PBS (10 mg/ml) the mixture was incubated for 10 min at room temperature, before dilution to a final volume of 1.0 ml in PBS-1% human serum. The radiochemical purity of the labeled Rituximab was determined by ITLC (1.5×15 cm ITLC-SG strips, Gelman Instrument Company) using 0.1 M acetate buffer of pH 6 as a mobile phase and the linear analyzer (Berthold). The injection of the radioimmuno-conjugate into the mice was performed 1 h after EOS. The in vitro behavior of the labeled bioconjugate (immunoreactivity, cell binding, cell killing efficiency) has been described in a previous paper [Vranje{hacek over (s)} S D, Miederer M, {hacek over (C)}omor J J, Soloviev D, Beyer G-J and the ISOLDE collaboartion. Labeling of monoclonal antibodies with 149-Th for targeted alpha therapy. J Lab Comp Radiopharm 2001; 44:718-720]. With the same antibody the inventors observed up to 55% cell binding without extrapolation to infinite antigen excess. Mice Survival Studies: The in vivo studies were performed using 26 female SCID mice (C.B.-17/ICR, Iffa Credo) under the authorization Nr: GE 31.1.1049/1879/11. The mice, which were 8 weeks old at the start of the experiment and weighed 20 g on average, were kept in sterile, ventilated boxes. Before injecting cells and antibodies, mice were anesthetized by i.p. (intra peritoneal) injection of 10 ml per kg (typically 0.2 ml) of an anesthetic (2.4 ml Ketasol 50, 0.8 ml Rompun, 6.8 ml 0.9% NaCi). Each mouse received 5-106 Daudi cells by injection of 0.2 ml cell suspension in PBS into the tail vein. Two days after xenotransplantation the mice were divided into four groups: the first group received 5 μg Rituximab in 0.1 ml PBS i.v.; the second group 300 μg Rituximab in 0.1 ml PBS i.v.; the third group 5.5 MBq 149Tb-CHX-A-DTPA-Rituximab radioimmunoconjugate (5 μg labeled Rituximab in 0.2 ml, i.v.), while the fourth group was left without any treatment. A summary of the in vivo study is presented in Table 2. According to the authorized protocol the mice were surveyed for 120 days: their behavior was logged each day, their condition was supervised once a week by a veterinarian, and they were weighed three times a week. At the appearance of obvious signs of paralysis, visible tumor masses, or a weight loss of >15%, the mice were sacrificed. One mouse was sacrificed shortly after injection (2 h p.i.) and kept deep-frozen for later analysis, in order to act as a reference for later quantification of the daughter radioactivity distribution. Retention and Daughter Radioactivity Distribution: Organ samples were taken from the sacrificed mice and the radioactivity concentration of the long-lived daughter products was determined by using high-resolution gamma spectroscopy (18% HP—Ge detector in combination with the Gamma spectrometer Genie 2000, Canberra). Whole, intact mice, as well as isolated organ samples were measured. Since the radioactivity content of the samples was essentially very low, long measuring times (between 1 and 24 hours) were applied. Statistical Analysis: The survival of animals until sacrifice because of disease development or the end of the experiment (no disease development) was compared between the different groups according a Kaplan Meier analysis using the Lee-Desu evaluation of the Unistat 3.0 statistical package (Megalon, Novato Calif., USA). 2. Results Preparation of Labeled Rituximab: Mass-separated and radiochemically pure 149Tb was obtained after chromatographic separation of the collected isobars with mass number A=149 at the on-line mass separator facility at CERN (FIG. 13). The overall time needed for the radiochemical separation and the labeling procedure was 1 hour. Radiolabeling of the Rituximab with this 149Tb-preparation was almost quantitative (>99%) within 10 minutes incubation time. The obtained preparation was thus ready for injection without further purification. The radioactivity concentration of the labeled antibody solution was 27.8 MBq/ml (0.75 mCi/ml), while the specific activity was 1.11 GBq/mg (30 mCi/mg) at the moment of injection. Survival in a SCID Mouse Model of Leukemia: The inventors set out to evaluate the efficacy of 149Tb-based TAT using a SCID mouse model of leukemia [16]. The inventor's experimental model involved the i.v. xenografting of lethal number of Daudi cells followed by TAT intervention at a time point when most of the Daudi cells would be expected to remain in circulation, and before the appearance of manifested tumors, which the inventors did not intend to target in this study. Survival data over a period of 4 months for treated mice and controls are shown in FIG. 14. All mice in the untreated control group developed clear signs of Burkitt lymphoma and were consequently sacrificed within 37 days. 50% of them developed visible macroscopic tumors while the others were sacrificed when they showed clear signs of paralysis or a weight loss >15% of the initial body weight (Table 2). The injection of a single, low dose of Rituximab (5 μg/animal) did not show any therapeutic effect, and all mice in this group had to be sacrificed within 43 days. As can be seen from FIG. 14, the survival curves of this group and the control group (untreated mice) are almost identical. 83% of mice in this group expressed obvious signs of paralysis or weight loss of >3 g, while 17% of the mice developed visible macroscopic tumor masses. A different survival pattern was observed after treatment with high dose of Rituximab (300 μg per animal, corresponding to 15 mg/kg). Although a single dose of 15 mg/kg Rituximab significantly increased the life expectancy—50% of mice in this group survived 100 days—ultimately, tumors developed in all animals (an example is shown in FIG. 15a) before the end of the observation period. In contrast, the mice treated with the radioactive 149Tb-CHX-DTPA-Rituximab (5 μg Rituximab per animal) were almost completely protected over the entire observation period, with only one mouse in this group being lost after 48 days due to abdominal tumor growth. The remaining 8 mice (89%) showed normal behavior without any signs of disease for 4 months after grafting (FIG. 15b). All of these mice were sacrificed after 120 days and were found tumor free at dissection. Thus, a single injection of 5.5 MBq 149Tb-labeled Rituximab (5 μg MoAb), which corresponds to an injected dose of 280 MBq/kg body weight (7.5 mCi/kg), produced long-term survival without evidence of any disease at 120 days. The survival increase after the RIT compared to all control groups (no treatment, 5 μg and 300 μg unlabeled Rituximab) was highly significant in the statistical Lee-Desu comparisons (p<0.005). Biodistribution of Labeled Rituximab and the Daughter Radionuclides: In FIG. 16 the inventors present typical γ-spectra of retained activity in organs recorded 120 days after injecting the short-lived radioimmunoconjugate. The biodistribution of 149Th-CHX-A-DTPA-Rituximab radioimmunoconjugate shortly after injection was assessed using a single mouse sacrificed at 2 h. The retention of the long-lived daughter nuclides at 120 days after injection is presented in Table 3. After 2 h, the organs with high blood pool like spleen, heart and kidney (42, 41 and 24% ID/g), showed relatively high radioactivity concentration. High amounts of the radioimmunoconjugate was found in the liver at this time point (18±3% injected dose) confirming the results of earlier systematic studies [Beyer G-J, Offord R E, Künzi G, Jones R M L, Ravn U, Aleksandrova Y, Werlen R C, Mäcke H, Lindroos M, Jahn S, Tengblad O and the ISOLDE Collaboration. Biokinetics of monoclonal antibodies labeled with radio-lanthanides and 225-Ac in xenografted nude mice. J Label Compd Radiopharm 1995; 37:229-530]. The values in the other organs were relatively low. After 120 days, 71.6% of the primary injected radioactive atoms had been excreted from the mice. The retention of the daughter products was 28.4±4%, out of which 91.1% remained in the bone tissue and 6.3% in the liver. TABLE 1Calculated values for mean partial molar enthalpies of solution of thechalcogens Q (Q = Se, Te, Po) in liquid lead and bismuth(Δ HsolvQ in Pb/Bi(1)) and mean partial molar enthalpies of evaporationfrom liquid lead and bismuth of the chalcogens Q in the monoatomic(Δ HvQ) and diatomic state (Δ HvQ2) and as diatomic metalchalcogenides molecules (Δ HvPbQ and Δ HvBiQ).Chal-cogenΔ HsolvQ in Pb/Bi(1)Δ HvQΔ HvQ2Δ HvPbQΔ HvBiQQ[kJmol−1][kJmol−1][kJmol−1][kJmol−1][kJmol−1]Se−38.7268.2203.8219.4177.9Te−11.2205.4147.8218.0164.5Po−8.8185.4159.3231.3176.3 TABLE 2Summary of the in vivo experiments on SCID mice xenotransplanted with Daudi cellsand treated by immunotherapy or radioimmunotherapy with 149Tb-labeled Rituximab.SCID mice groupsGroup 4Group 1Group 2Group 3(control group)No. of mice per 6496groupFirst i.v.5 · 106 Daudi cellsinjectionSecond i.v.5 μg Rituximab300 μg5 μgNONEinjectionRituximab149Tb-labeled2 days afterRituximabDaudi cell(5.55 MBq)inoculationFollow-up17% developed50% developed89% no50% developed(120 days after themacroscopicmacroscopicpathologicmacroscopictherapy)tumors,tumors,changes,tumors,83% paralyzed,50% paralyzed,11% paralyzed,50% paralyzed,weight lossweight lossabdominalweight losstumor TABLE 3Biodistribution of 149Tb-labeled Rituximab in SCID mice 2 h after i.v.injection (column 2 and 3) and of the remaining daughter radioactivitydistribution 120 days after injection (column 3 and 4). Note, that bothfemurs and both kidneys were combined for the gamma spectroscopicmeasurements in order to increase the signal to background ratio.2 h p.i.120 d p.i.Organ[% i.d./Organ][%/g tissue][% i.d./Organ][%/g tissue]Bloodn.a.<0.01Liver18 ± 3 24 ± 41.8 ± 0.31.6 ± 0.2Bone113 ± 1 9.1 ± 0.726 ± 4 13 ± 2 Spleen1.9 ± 0.242 ± 40.40 ± 0.0612 ± 2 Heart4.7 ± 0.741 ± 6<0.01Lung2.4 ± 0.518 ± 4<0.01Kidney26 ± 124 ± 4 0.2 ± 0.030.50 ± 0.08Muscles<0.2<0.02Bladder30.12 ± 0.03 3.7 ± 0.9<0.01Body total10028.4 ± 4 1Bone total was calculated as 9 × both femur activity2Both kidneys were measured together*3Bladder measured with urinen.a. not done, not assessable TABLE 4Radioactivity level of long-lived daughter products retained in a patientafter injection of 1 GBq 149Tb-Rituximab antibodies, assuming 100%retention of the long-lived daughter products (worst case). The retentionhas been measured to be only 28.4% independent on the decay mode(alpha or EC) (see Table 3), thus the real activity of daughter productswould be nearly a factor 4 smaller. On the other hand, the injection of a149Tb labeled bioconjugate 4 hours after the Tb purification wouldincrease the activity of the daughter product be by a factor 2. In this waythe numbers in this table can still be seen as upper limits.149Tb149Gd149Eu145Eu145Sm145Pm4.12 h9.28 d93.1 d5.93 d340 d17.7 ytinj = 1.0 GBq0 2 d310 kBq 13 MBq0.2 MBq3.9 MBq18 kBq 5 d 11 MBq0.4 MBq2.7 MBq37 kBq 43 Bq 10 d7.3 MBq0.7 MBq1.5 MBq57 kBq 86 Bq100 d 8 kBq0.7 MBq 41 Bq70 kBq0.8 kBq 1 y0.1 MBq41 kBq2.2 kBq 10 y 050 Bq3.1 kBq
050362054	summary	The present invention concerns the field of transmissive electron microscopy, particularly under a controlled atmosphere, of samples, especially graphite insertion compounds and has as an object a device for the transfer and insitu reactions under a controlled atmosphere of samples intended for such study. This apparatus may also be used, either internally even of the microscope chamber, or in the gate of this latter, for conducting chemical reactions between the sample or samples mounted on the grid or grids and gases of controlled temperature, pressure, composition and nature, which is particularly useful for performing a kinematic study of these reactions. Insertion compounds of graphite have been the subject of intense study for a number of years, notably because, on the one hand, of their electrical conductivity parallel to the graphite layers, which is very high at ambient temperature and for certain compounds close to that of cooper, together with a strong anisotropy which may be as high as 10.sup.6, and on the other hand, of the superconductive character of certain compounds which can attain the temperature of liquid helium and, finally, of the great variety of species which may be inserted between the graphite layers, namely from alkaline metals to halogens including metallic alloys, alkaline hydrides, the alkaline-earth metals, certain lanthanides, acids, halides or oxyhalides of transition metals, etc... The class of insertion compounds of graphite comprises at present 300 to 400 different phases taking account of the stage phenomenon. Thus, a compound is said to be stage 1, 2, 3... according as two successive inserted layers are separated by 1, 2, 3... layers of graphite. In certain cases, the stage may attain the value of 12 or 13. The great variety of reagents confers to the insertion compounds various properties amply justifying the intense study that has been devoted thereto. The insertion compounds frequently have phase transitions which are revealed by variations of electrical resistivity or thermal dilation. In a great number of compounds, there exists in addition an incommensurability between the host lattice and inserted lattice and the reflections obtained are sometimes of very weak intensity when examined with x-rays. It is therefore especially interesting to be able to study these compounds by any possible methods of investigation, and more particularly by transmissive electron microscopy. Such methods of investigation are nevertheless difficult to practice, because the insertion compounds are for the most part very fragile in air, such that there arise problems in transferring the samples to the chamber of an electron microscope. Specifically, emphasis has been placed on insertion compounds of graphite for which a transfer under a controlled atmosphere proves to be indispensable. The known sample holders may certainly be used for any types of materials whose geometric characteristics, especially thickness, are compatible with the observation by transmission in an electron microscope. In addition, they allow positioning of the samples, even those not fragile in air, such that they are caused to penetrate into the reaction chamber of the sample holder after a first examination with the electron microscope, to cause them thereafter to react and to analyze them without having to use other equipment and, above all, without them being returned into contact with the air. Thus, there does not exist at present sample holders adaptable to any type of side-inlet electron microcope, which permits both a transfer of samples under a controlled atmosphere and in-situ reactions. This latter concept is very important, because it permits performing chemical reactions when the sample holder is at the very interior of the microscope chamber and the grid holder rod is in the so-called transport position. There thus arises the problem of introducing a microscope grid, first prepared in a sterile chamber under controlled atmosphere, into the examining chamber of an electron microscope, without it coming into contact with the air. To this end, a rod enclosing the sample holders, which may be of variable diameter and provided with a projection serving to maneuver its opening through the vacuum pumps, then through the door to the gate of the microscope, must penetrate into the said gate to a length of about 20 cm. Protection of the sample by an exterior sleeve and kinematic studies could be effected by equipment mounted on a clamp for connection with the gate of the microscope, leading to a modification of this latter. Such modification, aside from being inconvenient, requires a lengthening of the sample holders which is incompatible with the size of most sterile chamber gates. It would also be possible to construct a sterile chamber mounted directly on the outlet of the gate of the microscope provided this latter is suitably modified. Such an embodiment is nevertheless rather complicated. On the other hand, a transfer device is known of which the forward part is provided with an exterior gasket which is received at the interior of the microscope chamber. This device nevertheless has a number of disadvantages, namely, on the one hand, a risk of deterioration of the gasket when sliding the movable portion of the sample holder between the guide rails of the object holder, on the other hand, a lack of control of the precise positioning of the object interiorly of the chamber because the guiding of the movement in this latter is effected through the intermediary of an elastic gasket and, finally, the provision of a single grid housing on the stage, which leads to relatively long operations of loading the grid for each sample. The present invention has as an object to overcome these disadvantages. Specifically, it has as an object a device for the transfer and in-situ reactions under a controlled atmosphere of samples for transmissive electrode microscopy, essentially constituted by a cylindrical rod of slight diameter comprising, near one end, two grid holders and slidably mounted in a cover of greater diameter provided with a sealed chamber for housing the grid holders in transport position, by a traction bar integrated with the end of the rod of slight diameter opposite the grid holders, guided in the cover of greater diameter and provided at its other end with a manipulating button, the grid holders being advantageously provided in a flat of the end of the rod and the cover being advantageously formed in two parts and having a front part of lesser diameter intended to penetrate into the gate of the microscope and provided with a longitudinal opening corresponding at least to the section of the grid holders of the rod in analysis position of this latter, characterized in that it is provided in addition with a means for guiding and locking in transport and analysis positions the rod provided with the grid holders, the said rod being advantageously integrated by screwing with the corresponding end of the traction bar which is guided in the cover, of which the opening of the front part of lesser diameter is extended over a portion of the length of this front part, on both sides, by longitudinal grooves, the said front part being connected to the rest of the cover by screwing of a shouldered portion, a gasket effecting the sealing at the level of the screwed assembly.
051270307	summary	DESCRIPTION 1. Technical Field The invention relates to tomographic imaging using penetrating radiant energy and more particularly to production of tomographic images which do not require use of image reconstruction techniques such as used in conventional computed tomography. 2. Cross-Reference to Related Application The present invention is an improvement of the invention described in Annis application Ser. No. 888,019 for Tomographic Imaging, the disclosure of which is incorporated herein by this reference. 3. Background Art An improved form of tomographic imaging is described in the cross-referenced application. As used herein, the term "tomography" or "tomographic imaging" represents imaging a selected slice of an object where the slice may or may not be planar; the term is analogous to laminography or planigraphy. The apparatus described in the- copending application includes a source of penetrating radiant energy to illuminate the object to be imaged, a sweep arrangement to form a pencil beam and to sweep the pencil beam over a line in space to form a sweep plane, some apparatus to support the object to be imaged so that the sweep plane intersects the selected slice of the object to be imaged, a radiation detector to detect energy scattered by the object, and a collimator. In the cross-referenced application the collimator is described as a line collimator because it is constructed to focus on or in the immediate vicinity of a line in space. By arranging the sweep plane to intersect the selected slice in or along the focal line of the collimator, an essential characteristic of tomographic imaging is satisfied. That essential characteristic is some way to localize or focus in on energy scattered by a single elementary volume of the object being imaged in preference to radiation scattered by other portions of the object. Since the illumination travels a linear path which is arranged to intersect the object, there are many elementary volumes of the object along that linear path which may scatter energy. If all or a significant portion of that scattered energy were detected, there would be no way to determine which of the energy was scattered by one elementary volume in preference to others. The line collimator performs a part of this localizing process by focusing on a line which can be considered the locus of a plurality of elementary volumes of the object. The focal line of the collimator is arranged so that it lies within the selected slice. In this fashion, energy scattered by elementary volumes of the object which do not lie within the selected slice is filtered. Because the pencil beam exists only at one path within the sweep plane, at any instant of time, the completion of the localizing function is effected. More particularly, while the collimator can accept energy scattered by any elementary volume lying along the preferential line, the pencil beam illuminates only a single elementary volume lying along the preferential line at any instant in time. An image of all the elementary volumes lying along the preferential line is formed as the pencil beam sweeps along the preferential line. At different instants in time, different ones of the elementary volumes along this preferential line are illuminated and energy scattered by the different elementary volumes lying along the preferential line are detected one after the other. The scattered energy which passes the collimator is detected and processed, sampled, digitized and stored. Thus as the pencil beam sweeps the line in space a line image is created in the digital memory storing the processed signals. A tomographic image of the entire slice is created by providing relative movement between the object and the imaging apparatus so the preferential line coincides with different linear segments of the selected slice as the relative motion displaces the selected slice relative to the imaging apparatus. In the copending application the sweep plane coincides with a plane of symmetry of the collimator, and the collimator is located between the source of illuminating radiation and the object being imaged. This arrangement works well (i.e. it is sensitive) for features in the selected slice which lie generally parallel to the sweep plane. However, the arrangement has a number of disadvantages.- As described in the copending application, it is important to collect as much of the scattered energy as possible. However, the parameters of the collimator determine the slice thickness and this requires that the channels of the collimator (defined between adjacent pairs of vanes) near the plane of symmetry be relatively narrow. The thin channels reduce the scattered flux which is accepted. This reduction can be understood by considering the need for thin channels and the practical limits on the thickness of the plates between which the channels are defined. The need for thin channels and the practical limits on plate thickness results in a minimum ratio of open area of the collimator's face to the blocked area of the face. Furthermore, the need to split the collimator so as to allow the sweep plane to pass through the collimator also reduces the effective solid angle subtended by the collimator at the selected slice. The present invention arranges the sweep plane so it does not coincide with a plane of symmetry of the collimator. In accordance with the present invention the slice thickness is defined not by the dimensions of the collimator, but by the illumination or pencil beam itself. This eliminates the constraint on the spacing of the vanes, for example it allows the spacing of the vanes which define the collimator to be made equal. Furthermore, depending on the orientation of the sweep plane relative to the plane of symmetry of the collimator, the solid angle subtended by the collimator can be significantly increased even as compared to the apparatus described in the copending application. For example in one embodiment of the invention the sweep plane is arranged to be substantially perpendicular to a plane of symmetry of the collimator. In another embodiment of the invention the sweep, plane makes an angle other than 90.degree. with a plane of symmetry of the collimator. The present invention has an additional advantage over the arrangement shown in the copending application in that x-ray photons which are "doubly scattered" (scattered twice) will not be detected with as high an efficiency as they would be in the arrangement of the copending application for a collimator with the same number of sheets or veins. Double scattered x-ray photons may pass through the collimator and enter the detector even though they do not suffer their first or second scatter along the focal line (the preferential line). The only requirements are that the second scatter take place within the acceptance wedge of the collimator and that the final angle of the second scatter be parallel to the extended line of the collimator plate (vein or sheet) which intersects the point of the second scatter. Since the x-ray beam in accordance with the present invention does not intersect the acceptance wedge of the collimator except at the focal line or preferential line (see FIG. 2 or 3), the first scatter, in order to generate a detected photon, must position the photon into the acceptance wedge in order for the photon to pass through the collimator after the second scatter. More particularly, the imaging apparatus in accordance with the invention is arranged to produce a tomographic image of a selected slice and includes a source of penetrating radiation and a sweep means for forming energy from the source into a pencil beam and for repeatedly sweeping the pencil beam over a line in space. The motion of the pencil beam defines a sweep plane which defines, at least in part, the size and location of the selected slice. There is means for supporting an object to be examined so the pencil beam and sweep plane intersect the object. Further means are provided for preferentially detecting radiation scattered, at any instant, by one of a group of selected volume elements in the slice, the second means subtending a large solid angle relative to the selected volume element, the second means including: radiation detector means developing at any instant in time a single signal reflecting radiation impinging on the radiation detector means, and a collimator located between the object and the radiation detector means, the collimator including: a plurality of radiation transmitting channels collectively establishing a field of view which intersects the sweep plane in a linear segment which is a locus of said selected volume elements so that the collimator passes radiation scattered by different elementary volume elements of the object lying along the linear segment as the sweep illuminates the different volume elements and where the collimator has an axis of symmetry which does not coincide with the sweep plane. Generally there is the desire to maximize the scattered energy which is detected without compromising the localizing function. This is accomplished by maximizing (or at least increasing) the solid angle subtended by the detector or collimator. Preferably the solid angle is within range of 0.05.pi. to 2.pi. steradian, or larger; a typical angle is .pi./2 steradian. Depending on the application, there may be one or more collimators, each associated with an element of the radiation detector. More particularly, it is a particular advantage of the invention that, at any instant, any radiation which is scattered, and which passes a suitable collimator (one which has a field of view intercepting the sweep plane in a bounded line along which the pencil beam is swept) was generated by the same elementary volume element of the object being imaged. Accordingly, wherever such scattered energy is detected, it can be used in forming the pixel representing that particular elementary volume. In some embodiments of the invention, the sweep plane is perpendicular to a plane of symmetry of the collimator, and the collimator is located only on one side of the object being illuminated. This limitation necessarily limits the solid angle subtended by the collimator to no more than 2.pi. steradians. However, by placing a second collimator (and associated radiation detector) on the other side of the object, this limitation on the solid angle is removed and the solid angle subtended by the collimators together can approach 4.pi.. The same is true for cylindrical bodies where the selected slice lies in the cylinder wall. With cylindrical bodies, although the "second" collimator could be placed on the other side of the body, that would mean that energy scattered by the selected slice might have to travel through another cylindrical wall section before reaching the collimator/detector. It would be preferable, given a sufficient radius of curvature of the cylindrical body, to locate the "second" collimator and detector, "inside" the cylindrical object. The use of multiple collimator/detectors is not limited to embodiments in which the sweep plane lies perpendicular to a plane of symmetry of the collimator(s).
043483534	summary	BACKGROUND OF THE INVENTION The present invention relates generally to nuclear reactor fuel assemblies and more particularly to a reusable system for removably attaching a nuclear reactor fuel assembly duct tube to a nuclear reactor fuel assembly inlet nozzle. To produce power from the nuclear reactor, it is necessary to assemble a concentration of fissionable uranium, thorium, and/or plutonium in a quantity and in a physical configuration capable of sustaining a continuous sequence of fission reactions. This concentration is frequently referred to as the reactor core. The heat that the fission reactions generate is transferred to a fluid such as, for example, liquid sodium. The sodium, in turn, transfers its heat to a secondary coolant, or ultimate coolant, which can be used to drive turbines which power electrical generation equipment. Because the radiation, pressure, temperature, flow velocity, and other environmental conditions within the reactor core are quite hostile, the reactor core must be of sturdy construction. These conditions produce a number of phenomena with which it is very difficult to cope. Thus, for example, thermal, pressure, and irradiation effects tend to produce considerable creep in materials which can expand dimensions, cause bow and other effects which shorten fuel assembly life. Also, due to varying material requirements throughout a reactor core, it is often difficult to achieve ideal fuel behavior due to fabrication constraints. These effects have been known to produce a type of deterioration in that dimensional constraints of a fuel assembly in a reactor core can no longer be met over a period of time. These conflicting requirements have been reconciled to a great extent by loading pellets of uranium or plutonium dioxide into long slender tubes called fuel rods. With the tubes loaded with pellets and the ends of each tube sealed, these fuel rods are arranged longitudinally parallel with each other and are arranged in generally hexagonal arrays of about 200 fuel rods. Each array is called a fuel assembly. These fuel assemblies all are mounted side-by-side in a larger, generally right circular cylindrical configuration that characterizes the reactor core. Each fuel assembly often includes an upper handling socket and a lower nozzle which are both attached to a duct tube containing the fuel rods. The fuel rods receive radial support from the duct tube and longitudinal support from the lower nozzle. Coolant usually flows in through the lower or inlet nozzle of the fuel assembly, up through the spaces between fuel rods and out through the upper handling socket of the fuel assembly. Often, duct material incompatibility with the environment can degrade the operating life expectancy of the fuel assembly. This gives a need, with a significant cost incentive, for being able to remove duct tubes from their inlet nozzles for various reasons such as inspection, testing, maintenance or replacement, without the destruction of the duct tube and/or inlet nozzle. The prior art attaches the duct tube to the inlet nozzle by welding, and uses destructive techniques to separate them. SUMMARY OF THE INVENTION It is an object of the invention to quickly and inexpensively attach a nuclear reactor fuel assembly duct tube to a nuclear reactor fuel assembly inlet nozzle. It is another object of the invention to quickly, inexpensively and nondestructively remove an attached nuclear reactor fuel assembly duct tube from a nuclear reactor fuel assembly inlet nozzle. It is a further object of the invention to allow repeated attachment and removal of a nuclear reactor fuel assembly duct tube and a nuclear reactor fuel assembly inlet nozzle. Additional objects, advantages and novel features of the invention will be set forth in part in the description which follows, and in part will become apparent to those skilled in the art upon examination of the following or may be learned by practice of the invention. The objects and advantages of the invention may be realized and attained by means of the instrumentalities and combinations particularly pointed out in the appended claims. To achieve the foregoing and other objects and in accordance with the purpose of the present invention, as embodied and broadly described herein, the nuclear reactor fuel assembly duct tube and inlet nozzle attachment and removal system may comprise a nuclear reactor fuel assembly inlet nozzle, a nuclear reactor fuel assembly duct tube, a retaining collar and a locking nut. The inlet nozzle's upper end has a top section with the shape of a generally equilateral polygon and a bottom section with the general shape of a cylinder. The two sections are coaxially joined. Each top section's side contains an outside recess while the bottom section contains outside threads. The duct tube's lower end has a similar shape to that of the previously mentioned polygon, and its sides have outwardly-extending protrusions followed by terminal deflectable locking tabs which end in inwardly-extending flanges. The locking tabs can be made to slide over the inlet nozzle's top section, and the flanges can be made to engage the top section's recesses. The retaining collar's top segment has a shape similar to that of the previously mentioned polygon and its bottom segment has a generally cylindrical shape. The two segments are coaxially joined. The top segment, during attachment, surrounds and restrains the flanges and engages the protrusions. The bottom segment, during attachment, is positioned above only part of the outside threaded portion. The locking nut's inside threaded portion connects with the outside threaded portion of the inlet nozzle to lock the retaining collar against the protrusions during attachment. Several benefits and advantages are derived from the invention. The invention's rapid attachment and removal feature allows easy removal of a nuclear reactor fuel assembly duct tube from its inlet nozzle for inspection, testing, maintenance and the like. The invention's reusability feature allows the removed duct tube and/or inlet nozzle to be reconnected or replaced with a different one. This is to be contrasted with the prior art which provides for a destructive, one-time removal system. The invention's reusability feature and rapid attachment and removal feature provide an economic benefit for test and/or commercial reactors. For example, various duct tube material configurations can be tested without degrading fuel assembly life and the limited life of assemblies could be extended.
044302908	claims	1. A plasma confining device comprising: a plurality of cusp magnetic field means each for generating a cusp magnetic field confining a charged particle plasma, said cusp magnetic field having a line cusp portion, and some of said charged particles escaping through said line cusp portion forming a sheet-shaped charged particle flow substantially within a two-dimensional plane along said line cusp portion, and a deflection magnetic field means for generating a deflection magnetic field having a magnetic flux perpendicular to said plane, an edge of the deflection magnetic field being perpendicular to the direction of advancement of the escaping charged particles so that each of the escaping particles enters the deflection magnetic field, is deflected, leaves the magnetic field, and returns substantially within said plane, toward and through the line cusp portion into the inside of the plasma, wherein said deflection magnetic field means further includes neutralizing coils which generate a neutralizing field which will weaken said cusp magnetic field at said deflection magnetic field means, thereby reducing the radius of said deflection field means.
051911574	claims	1. A method of safely disposing of hazardous waste, including the steps of: forming a borehole in the earth's crust extending from the surface to the interior of a geopressure cell by penetrating the geopressure barrier seal; placing the hazardous waste material in the geopressure cell by moving the hazardous waste through the borehole; and reestablishing the geopressure seal of the geopressure cell to prevent migration of the hazardous waste from the geopressure seal. 2. The method as set forth in claim 1, wherein the step of placing further includes: lowering sealed containers filled with hazardous material down the borehole. 3. The method as set forth in claim 2, wherein the steps of placing further includes: stacking the lowered sealed containers in the borehole. 4. The method as set forth in claim 3, wherein the step of placing further includes: centering the stacked sealed container in the borehole to enable surrounding encasement of the containers within the geopressure cell. 5. The method as set forth in claim 1, wherein the step of placing further includes: pumping the hazardous material down the borehole and into the geopressure cell. 6. The method as set forth in claim 1, wherein the step of forming a borehole further includes: penetrating a transition zone seal forming the geopressure cell. penetrating a seal forming the geopressure cell. substantially filling the borehole from the geopressure cell to the surface with cement. substantially filling the borehole with a suitable material. 7. The method as set forth in claim 1, wherein the steps of forming a borehole further includes: 8. The method as set forth in claim 1, wherein the step of reestablishing the seal of the geopressure cell, includes: 9. The method as set forth in claim 1, wherein the step of reestablishing the seal of the geopressure cell, includes:
	claims	1. A probe system for detecting a coolant level and flow velocity in a nuclear reactor, in combination, comprises a conductivity probe; and a time-domain reflectometer probe, wherein the probe system is at least partially arranged in a downcomer of the nuclear reactor. 2. A boiling water reactor, comprising:a reactor pressure vessel;a core shroud arranged concentrically inside the reactor pressure vessel to provide an annular downcomer forming a coolant flow path between a wall of the reactor pressure vessel and the core shroud; andsystem as recited in claim 1 for determining one of a coolant level and a flow velocity within the reactor. 3. The reactor according to claim 2, wherein the conductivity probe comprises one of an electrical conductivity (EC) probe and a thermal conductivity (TC) probe. 4. The reactor according to claim 2, wherein the probe system is at least partially arranged within the downcomer of the reactor for determining the coolant level and the flow velocity within the downcomer of the reactor. 5. The reactor according to claim 2, wherein the boiling water reactor is a natural circulation boiling water reactor. 6. The probe system according to claim 1, wherein the conductivity probe comprises one of an electrical conductivity (EC) probe and a thermal conductivity (TC) probe. 7. The probe system according to claim 1, wherein the nuclear reactor comprises a boiling water nuclear reactor. 8. The probe system according to claim 7, wherein the boiling water nuclear reactor comprises a natural circulation boiling water reactor.
	claims	1. An arrangement for generating extreme ultraviolet radiation from a plasma generated by energy beam with high conversion efficiency comprising:a pulsed energy beam;a plasma generation chamber, said pulsed energy beam being directed to a location in said chamber where it interacts with a target;a target feed device containing a mixing chamber for generating a mixture of particles of an emission-efficient target material with at least one carrier gas and containing an injection unit for dispensing individually defined target volumes into the plasma generation chamber in a metered manner in order to supply only as much emission-efficient target material to the interaction location as can be converted into radiation by an energy pulse;said target feed device having a gas liquefaction chamber;said target material being supplied to the injection unit as a mixture of solid metal particles in liquefied carrier gas;said injection unit having a droplet generator with a nozzle chamber and a target nozzle for generating a defined droplet size and series of droplets; andmeans, which are controllable in a frequency-dependent manner and which are triggered by the pulse frequency of the energy beam, being connected to the injection unit for generating a time-controlled series of droplets. 2. The arrangement according to claim 1, wherein the liquefaction chamber is arranged downstream of the mixing chamber so that the solid particles are supplied to the liquefaction chamber so as to be mixed with the carrier gas, and the liquefaction chamber is designed for liquefying the mixture. 3. The arrangement according to claim 1, wherein the liquefaction chamber is arranged upstream of the mixing chamber so that the liquefaction chamber is designed for liquefying the clean carrier gas, and the mixing chamber is designed for mixing the solid particles with the liquefied carrier gas. 4. The arrangement according to claim 1, wherein the solid emission-efficient particles comprise tin or a tin compound. 5. The arrangement according to claim 1, wherein the solid emission-efficient particles comprise lithium, or a lithium compound. 6. The arrangement according to claim 1, wherein the solid emission-efficient particles have a size of less than 10 μm. 7. The arrangement according to claim 1, wherein the carrier gas is a noble gas, preferably argon. 8. The arrangement according to claim 7, wherein the noble gas is argon. 9. The arrangement according to claim 1, wherein the carrier gas is nitrogen. 10. The arrangement according to claim 1, wherein light noble gases are mixed in with a carrier gas that is selected as the main component in order to limit more narrowly the spectral band width of the EUV emission at 13.5 nm. 11. The arrangement according to claim 1, wherein individual droplets ejected from the injection unit have a diameter between 0.01 mm and 0.5 mm. 12. The arrangement according to claim 1, wherein means for removing individual targets are arranged downstream of the target nozzle of the injection unit so that the frequency of the individual targets arriving in the interaction location exactly corresponds to the pulse frequency of the energy beam. 13. The arrangement according to claim 12, wherein electric deflecting means are arranged downstream of the target nozzle of the injection unit for lateral deflection of unnecessary individual targets from the series of droplets dispensed by the target nozzle. 14. The arrangement according to claim 12, wherein a mechanical closure device is arranged downstream of the target nozzle of the injection unit for defined elimination and passage of individual targets from the series of droplets dispensed by the target nozzle. 15. The arrangement according to claim 12, wherein the target generator of the injection unit has a pressure modulator at the nozzle chamber in order to increase the chamber pressure temporarily for ejecting an individual droplet when needed, and a nozzle antechamber is arranged downstream of the target nozzle, wherein a pressure which is higher than that in the plasma generation chamber and which is adapted to the gas pressure of the gas feed to the mixing chamber is adjusted in the nozzle antechamber to prevent unwanted dripping of target material from the target nozzle as long as no pressure pulse is generated by the pressure modulator. 16. The arrangement according to claim 15, wherein the pressure of the gas feed to the mixing chamber is adjusted so as to be slightly higher than that in the nozzle antechamber in order to adapt the pressure in the nozzle antechamber. 17. The arrangement according to claim 1, wherein a sufficient quantity of particles is provided in a reservoir and supplied to a plurality of mixing chambers which are arranged in parallel and connected to the injection unit so as to be switchable in series for continuous injection into the plasma generation chamber. 18. The arrangement according to claim 1, wherein the particles are provided so as to be mixed with the carrier gas in a mixing chamber and a line connection point with a feed line from another carrier gas feed is arranged downstream of the mixing chamber, wherein at least one of the feed lines to the connection point has a throughflow regulator which is controllable by a measuring device which is arranged downstream of the connection point and which determines the proportion of particles in the gas flow in order to adjust a desired mixture ratio of mixed carrier gas and clean carrier gas. 19. The arrangement according to claim 18, wherein the measuring device for controlling the mixture ratio is an optical scatter light measuring unit. 20. The arrangement according to claim 1, wherein the pulsed energy beam is at least one laser beam. 21. The arrangement according to claim 1, wherein the pulsed energy beam is an electron beam. 22. The arrangement according to claim 1, wherein the pulsed energy beam is an ion beam.
	summary
	description	This application is a continuation-in-part (CIP) of pending U.S. patent application Ser. No. 11/848,414, filed on Aug. 31, 2007, the entirety of which is incorporated by reference. The present invention relates to a method for shielding from debris, a nuclear fuel assembly including a fuel bundle and a upper tie plate for supporting the fuel assembly and, particularly, relates to a debris shield that is releasable, for the purpose of cleaning or replacement, from an Upper Tie Plate (UTP) of the fuel assembly. In a nuclear fuel assembly, liquid coolant/moderator flows into the assembly thru the bottom and exits as a water and steam mixture from the top during normal operation. The core of a nuclear fuel reactor includes a plurality of fuel assemblies arranged in vertical side-by-side relation, each containing a bundle of fuel rods. The fuel assemblies include a fuel bundle and a housing formed by a hollow metal channel. The fuel bundle includes an array of parallel fuel rods, water rods and one or more tie plates, spacers, and finger springs, that support the rods in the bundle. Generally, a fuel bundle includes an upper tie plate near the top of the bundle, which is removable and a lower tie plate at the bottom of the bundle, which is conventionally a non-removable unit. Debris may fall through a conventional upper tie-plate and become lodged within the fuel bundle where the debris may cause fuel rod fretting during normal reactor operating conditions. Fretting is potentially damaging to the fuel rods, resulting in what is typically known as a “leaker”. Conventional efforts to address debris falling down into a fuel assembly typically focus on the prevention of debris from entering within the coolant itself and coolant flow passages, prior to coolant flow entering the fuel assemblies within the core. Conventional efforts typically involve administrative controls regarding the treatment of coolant flow passages and handling of fuel assemblies such that debris does not enter the passages of the fuel assemblies. These controls are designed to alleviate the sources of debris such that debris does not fall down into fuel assemblies. Nevertheless, there is a risk that debris will fall down into a fuel assembly, especially while the coolant flow stops and the reactor core is open or when service work is preformed on the core. There is a long felt need for procedures and special devices to ensure that debris does not fall into fuel assemblies from above, especially during refuel operations, fuel inspections and when the coolant is in a reverse coolant flow pattern. Further there is a long felt need for efficient and non-intrusive methods to insert, remove and clean devices associated with the UTP that capture debris that would otherwise flow down into a fuel bundle assembly from above. A nuclear reactor fuel assembly comprising: a fuel bundle including an array of fuel rods mounted in an upper tie plate and housed within a hollow metal channel, and a debris shield in the upper tie plate and above the fuel rods, wherein the debris shield has a surface at least co-extensive with an at least partially open area of the upper tie plate. A nuclear reactor fuel assembly has been developed including: an upper tie plate having apertures to receive and support fuel rods of a fuel bundle; the fuel bundle including an array of the fuel rods mounted and housed in walls of a hollow metal channel, and a porous debris shield which can be removable from the upper tie plate, attached as a permanent integrated structure in the upper tie plate, or mounted such that the debris shield remains over the fuel bundle while the upper tie plate is removed, wherein the shield is porous. A nuclear reactor fuel assembly has been developed comprising: a fuel bundle including an array of fuel rods mounted in an upper tie plate and housed within a hollow metal channel, and a debris shield which can be removed from the upper tie plate, fixed to the upper tie plate or remain above the fuel rods while the upper tie plate is removed, wherein the debris shield has a surface at least co-extensive with an open area of the fuel bundle inside the hollow metal channel. A method has been developed to prevent debris falling into a nuclear reactor fuel assembly including a bundle of fuel rods mounted within or below an upper tie plate and housed in a hollow metal channel, the method comprising: inserting a debris shield which can be a removable unit, attached as a permanent integrated structure within the upper tie plate or remain attached to the fuel bundle while the upper tie plate is removed; maintaining the shield within the upper tie plate or above the fuel rods, while the fuel assembly is in an operating nuclear reactor core; flowing coolant through the bundle and the debris shield during operation of the nuclear reactor core, wherein the debris shield has the ability for capturing, screening, filtering, deflecting, and removing falling debris in the fuel assembly. A nuclear reactor fuel assembly is disclosed comprising: a fuel bundle including an array of fuel rods housed within a hollow metal channel; an upper tie plate including a horizontal, rectangular frame having a rib to receive an upper end of a tie rod of the fuel bundle, wherein an outer surface of the frame is adjacent the hollow metal channel, and a debris shield in the upper tie plate and above the fuel rods, wherein the debris shield has a surface at least co-extensive with an at least partially open area of the upper tie plate. A method is disclosed to capture and remove debris falling into a nuclear reactor fuel bundle assembly including a bundle of fuel rods mounted below an upper tie plate and housed in a hollow metal channel, the method comprising: inserting a debris shield in the upper tie plate; maintaining the shield in the upper tie plate and above the fuel rods, while the fuel bundle assembly is in an operating nuclear reactor core; flowing coolant through the fuel rods and the debris shield during operation of the nuclear reactor core; capturing debris falling in the fuel assembly on the debris shield; after capturing the debris, removing the fuel bundle assembly with the inserted debris shield from the nuclear reactor core to a maintenance/fuel inspection pool and thereafter may require removing the debris shield from the upper tie plate, performing at least one of (a) removing the captured debris from the removed debris shield and thereafter reinserting the debris shield back into the upper tie plate and (b) inserting another debris shield into the upper tie plate, and after the reinsertion of the debris shield or the insertion of the another debris shield, moving the fuel bundle assembly from the maintenance/fuel inspection pool to the nuclear reactor core once again. A method is disclosed to capture and remove debris falling into a nuclear reactor fuel bundle assembly including a bundle of fuel rods mounted below an upper tie plate and housed in a hollow metal channel, the method comprising: inserting a debris shield in the upper tie plate; maintaining the shield in the upper tie plate and above the fuel rods, while the fuel bundle assembly is in an operating nuclear reactor core; flowing coolant through the fuel rods and the debris shield during operation of the nuclear reactor core; capturing debris falling in the fuel assembly on the debris shield; after capturing the debris, removing the fuel bundle assembly with the inserted debris shield from the nuclear reactor core to a maintenance or fuel inspection pool and thereafter may require removing the debris shield from the upper tie plate, performing at least one of (a) removing the captured debris from the removed debris shield and thereafter reinserting the debris shield back into the upper tie plate and (b) inserting another debris shield into the upper tie plate, and after the reinsertion of the debris shield or the insertion of the another debris shield, moving the fuel bundle assembly from the maintenance/fuel inspection pool to the nuclear reactor core once again. A method is disclosed to seat an upper tie plate in a fuel bundle assembly for a nuclear reactor, wherein the fuel bundle assembly includes a bundle of fuel rods mounted below the upper tie plate and housed in a hollow metal channel, the method comprising: attaching a debris shield to upper portions of the water rods, and after attaching the debris shield, seating the upper tie plate over the debris shield and securing the upper tie plate to the tie rods of the fuel bundle assembly. A method is disclosed to remove a fuel rod from a fuel bundle for a nuclear reactor, wherein the fuel bundle includes a bundle of fuel rods mounted below the upper tie plate, the method comprising: removing the upper tie plate from the fuel bundle, while the debris shield remains attached to an upper portion of at least one of the fuel rods or of a water rod, and after removing the upper tie plate, removing the fuel rod by lifting the rod up through the debris shield, while the debris shield remains attached to an upper portion of at least one of the fuel rods or of a water rod. A fuel bundle is disclosed comprising: a plurality of fuel rods; an upper tie plate having a cavity to receive an upper tie plate debris shield, wherein the upper tie plate is not directly connected to the fuel rods; the upper tie plate debris shield is seated in the cavity and has at least one opening through which extends an upper end of one of the fuel rods. Debris shields are disclosed herein that mitigate the entry of foreign material into the top of a nuclear fuel bundle assembly. The debris shields deflect, capture, screen and retain or aid in the removal of foreign particles introduced into the top of the fuel assembly. By preventing the entry of foreign particles into the nuclear fuel bundle assembly, the possibility of a fuel rod fretting failure is substantially reduced, if not eliminated. The debris shield may be a removable unit from the upper tie plate or a permanently integrated structure within or below the upper tie plate, but above the fuel and water rods. Preventing debris falling into the nuclear fuel bundle assembly is expected to assure the expected operational life of the nuclear fuel bundle assembly by reducing the risk of fuel rod failure and the premature discharge from the core of the reactor. FIG. 1 is a side view showing in cross-section of a conventional nuclear fuel bundle assembly 10 shaped generally as a vertical column with a square cross-section. The reactor core 21 includes many fuel assemblies arranged side by side in a predefined array. The nuclear fuel bundle assembly typically includes, for example, an array of full-length fuel rods 11 and part-length fuel rods 12 arranged in parallel. The fuel rods are supported by an upper tie plate 13, a lower tie plate 14, lower grid 29 of the tie plate, which supports the rods within the nuclear fuel bundle. One or more spacers 15 arranged at locations along the length of the fuel rods. One or more water rods 23, 230 typically extend through the center of the array of fuel rods and are attached to the upper tie plate 13 by an upper end plug 30. The typical nuclear fuel bundle assembly 10 comprises the fuel rods 11, 12, water rods 23, 230, an upper and lower tie plates 13, 14, and tie rods 24, spacers 15, finger springs 18, expansion springs 16, and a hollow metal channel 20. Fuel rod expansion springs 16 extend from the upper end plugs of 19, 30 and 28 on each of the water rods 23 and 230 and the full-length fuel rods 11 and the tie rods 24, to the under side of the upper tie plate 13. Hex nuts 17 are used to secure the threaded upper end plugs 28 of the tie rods 24, to the top of the upper tie plate 13, a lock tab washer 170 (FIG. 2) is used to keep the hex nut 17 secured to the tie rods 24. The tie rods 24 extend through the upper tie plate 13 of the tie plate. The opposite end of each tie rod is secured to the lower tie plate's grid 29, of a conventional nuclear fuel bundle assembly 10. The lower tie plate 14 includes finger springs 18 on the outer sidewalls of the lower tie plate 14 that engage a hollow metal channel 20 that provides a hollow housing for the nuclear fuel bundle assembly 10. The hollow metal channel 20 is typically an elongated hollow metal tube, rectangular in cross-section and having a length that extending from the upper tie plate 13 to the lower tie plate 14. The hollow metal channel 20 covers the fuel rods and water rods in the nuclear fuel bundle assembly 10. Channel posts 25 and channel post 26, located on opposite corners of the upper tie plate 13, provide guides to align the channel onto the fuel bundle 32. A channel fastener clip (not shown) secures the hollow metal channel 20 to the threaded channel posts 26, on one side of an upper tie plate 13. The threaded channel post 26 is used to insure correct ordination and alignment of the hollow metal channel 20 to the fuel bundle 32 during the channeling operation. Otherwise, the hollow metal channel 20 cannot be securely fastened to the upper tie plate 13 by the channel fastener (not shown) as required. Generally, a U-shaped lifting handle 22 is attached as part of the upper tie plate 13. The handle 22 may be used to raise and lower the nuclear fuel bundle assembly 10 into a reactor core 21 or to otherwise move the assembly around the facility as needed. Debris may enter the top of a conventional nuclear fuel bundle assembly 10, especially during non-operating or operating conditions such as, refuel, new fuel receipt, transport to and from the core, when the coolant flow stops flowing upward through the core, and when moderator flow may be stagnate or reversed. Debris falling into the top of the fuel bundle assembly may become lodged in the grid of the tie plate, spacer bracket, between the fuel rods or between a hollow metal channel wall and a fuel rod. The crevices in the fuel bundle can trap the debris within the fuel bundle. The debris may fall through an upper tie plate 13 and become lodged within a location inside the nuclear fuel bundle assembly 10 where it could cause fuel rod fretting during normal reactor operating conditions. FIGS. 2, 3 and 4, show an upper tie plate assembly 100 having an internal debris shield 102 (FIG. 2) that may be a removable unit or a permanent integrated structure within the upper tie plate. The shield 102 may be mounted to or integral within a horizontal support frame 104 of the assembly. The debris shield 102 fills an area enclosed by the horizontal support frame 104 of the upper tie plate. The debris shield may be inserted in the bottom of the upper tie plate or in a horizontal slot within the upper tie plate. The debris shield may or may not be integral with the frame structure of the upper tie plate, and may or may not be removable from the frame structure. The debris shield 102 is porous and screens or filters debris from the fluid flow through the nuclear fuel bundle assembly 10. Coolant flows through the debris shield preferably without substantial pressure loss through the debris shield 102. An array of cylindrical pin supports 106 are provided by and are seated within openings in the debris shield 102 to allow the passage of the upper end plugs 28, of the fuel rods 24 and end plugs 30 of fuel rods 11, along with the upper end plugs 19 of the water rod. These cylindrical pin supports 106 provide structural support for the fuel rods and the water rods, in that the debris shield supports the threaded tie rods 24 and the threaded upper end plug on each of the water rods 23 and 230, and the non-threaded fuel rods 11, and the associated expansion springs 16. In one embodiment, the cylindrical pin supports 106 have a diameter slightly greater than a diameter of a fuel rod with an associate expansion spring. The cylindrical pin, along with the fuel rod and expansion spring, may be removed from the debris shield by sliding the cylindrical pin, fuel rod and expansion spring up through the debris shield. Sliding the cylindrical pin, fuel rod and expansion spring up through an opening in the debris shield allows the fuel rod to be removed after the upper tie plate is removed and while the debris shield remains on the other fuel rods and water rods. The upper end plugs 30, for the fuel rods 11 and the upper end plugs 19, for the threaded water rods 23, 230 and the threaded upper ends 28 of tie rods 24, extend through the pin supports 106 in the debris shield 102. A hex nut 17 is used to secure the threaded ends of the upper end plugs, such as of the water rods. Lock tab washers may be used on the hex nut. The hex nut 17 are general seated on an upper surface of the upper tie plate 13. The expansion springs 16 slide over the upper end plugs of the fuel and water rods and apply an upward bias force on a bottom surface of the debris shield 102. The extension of the upper end plug, on the fuel rods 24, passes through the pin support 106 in the debris shield 102. The debris shield 102 is secured to the top of the threaded upper end plugs on each of the water rods 23 and 230 by hex nut 17. A recess cavity under the upper tie plate assembly 100 allows the hex nut 17 to be seated, in a locked position, on top of the debris shield and in the cavity of the upper tie plate. The cavity may be in the structural rib of the upper tie plate that extends diagonally across the top of the debris shield. The fuel rods 11 and water rods 23, and 230 may not be directly secured to the top of the upper tie plate 13. They are secured indirectly to the upper tie plate through the debris shield. The upper tie plate may be removed while the debris shield remains attached to the upper ends of the water rods and, optionally, to upper ends of the fuel rods. The upper tie plate assembly 100 may be removed from the nuclear fuel bundle assembly, while the internal debris shield 102 remains attached directly to the water rods 23, 230, and while the debris shield supports the fuel rods 11 and 24. If the debris shield 102 seats in a bottom cavity 109 of the upper tie plate, then the upper tie plate may be lifted to separated the debris shield 102 from the upper tie plate, as is shown in FIG. 4. To remove the upper tie plate, the hex nuts 17 and lock tab washers 170 are removed from the tie rods 24 that extend through the frame of the upper tie plate. Thereafter the upper tie plate may be lifted off the fuel bundle assembly. The debris shield 102 remains attached to the water rods 23, 230, while fuel rods 11 and 24 are supported and held securely in their original positions by extending through their respective cylindrical pin supports in the debris shield. The fuel rods are held in place without the aid of an upper tie plate, which has been removed. The upper tie plate may be replaced or repaired, because it can be removed from the nuclear fuel bundle separately from the debris shield and fuel and water rods. Flow passages 108 through the debris shield 102 have an axis or axes that are preferably in a direction other than the vertical axis of the nuclear fuel bundle assembly 10. These non-straight flow passages stop, trap, and catch debris materials falling from above. By way of example, the passages through the debris shield 102 slant in a first direction in the upper half of the debris shield and slant in a second direction in the lower half of the debris shield, with a corner 116 formed between the two halves. The slanting passages of the debris shield 102 blocks light passing vertically through the debris shield. Due to the slanting of the flow passages 108 in the debris shield 102, it is not possible to look directly through the debris shield 102. The view is blocked because the flow passages 108 are not straight and have corners 116. Just as the view is blocked, debris is blocked and trapped in the corners 116 and between the two halves of the debris shield, by the non-straight lines within the flow passages 108. In the example shown in FIGS. 2 to 4, the debris shield 102 is an arrangement of angled metal strips arranged side-by-side to form a chevron pattern 112 in cross-section that is generally perpendicular to the fuel bundle assembly 10. The passages 108 are formed between the side-by-side strips. Alternatively, the debris shield 102 may be formed of a porous material, such as a wire or fabric mesh, sponge, grid, array of crossing bars or slats, or other matrix material. The debris shield 102 is generally porous flat plates having perimeter edges 110 that abut the interior surfaces of the frame 104 of the upper tie plate assembly 100, from below. The debris shield 102 preferably remains in the nuclear fuel bundle assembly 10 during operation of the nuclear reactor core 21. The debris shield 102 may have a porosity, open mesh or matrix structure that allows coolant, especially emergency coolant, to flow down through the debris shield without substantial creating flow resistance. The porous, chevron, mesh or matrix structure of the debris shield 102 blocks the passage of debris without substantial impending the general flow. The debris shield serves as a filter, or a screening device that allows the passage of fluids, such as coolant, and blocks the passage of foreign particulates. Preferably the debris shield 102 should have a pore size that minimizes the passage of most any foreign debris, without imposing a significant fluid pressure drop across the debris shield 102. The debris shield 102 filters and captures debris in the coolant flow, especially debris flowing downward from above the fuel bundle assembly 10 and to the upper tie plate assembly 100. The passages 108 of the debris shield 102 are too narrow to allow larger debris particles to enter the passages or to pass all the way through. Larger debris particles are captured on an upper surface 114 of the debris shield 102. Smaller debris particles may enter the flow passages 108 of the debris shield 102 and become lodged within the corners and crevices 116 of the debris shield, such that they do not flow down below the upper tie plate assembly 100 and into the nuclear fuel bundle assembly 10. The horizontal support frame 104 is preferably a rigid structural frame. The frame 104 may be porous, e.g., as have small vertical openings 31 to allow fluid to pass through the frame. These openings 31 have a small diameter to help block the passage of debris from entering the fuel bundle. The multiple openings 31 in the structure of the horizontal support frame 104 increase the effective flow area of passages through the upper tie plate assembly 100 to compensate or offset any flow restrictions due to the debris shield 102. Datum points 117, e.g., vertical ribs on the outer walls of the frame, attached or are integral within the horizontal support frame 104. The datum points aid in centering the hollow metal channel 20 about the upper tie plate 13. The datum points may also be used during fabrication of the nuclear fuel bundle assembly 10, to square the upper tie plate assembly 100 to the full-length fuel rods 11, the part length fuel rods 12, the tie rods 24, and the water rods 23 and 230. FIG. 3A is an enlarged sectioned view showing the connection between a water rod 23 or 230 11 and the debris shield 102 which is seated in a cavity of the upper tie plate 102. The water rod has a threaded upper end which receives a hex nut 17. A cylindrical pin support 106 is seated in a vertical opening 101 of the debris shield. The cylindrical support pin is hollow to receive the upper end of the water rod. The cylindrical support pin 106 may includes an upper flange 107, e.g., having a shape of a washer, that provides an abutment for an upper surface of the debris shield and, on the other side of the flange, an abutment for the hex nut 17. The upper tie plate has a recess 105 to receive the hex nut 17 and the upper tip of the water rod. The recess 105 may be shaped such that the hex nut cannot turn while seated in the recess. Further, the bottom edge of the cylindrical pin support 106 may be attached to an upper end of an expansion spring 16. FIG. 3B is an enlarged sectioned view showing the connection between an unthreaded upper end plug 30 of a fuel rod 11 and a debris shield 102 which is seated in a lower cavity of an upper tie plate 100. The expansion spring 16a extends through the cylindrical pin support 106a and abuts against or is connected to a upper annular ledge 111 of a recess for the fuel rod 11 in the upper tie plate 100. The expansion springs on the unthreaded fuel rods may apply an upward basis force to the bottom surface of the upper tie plate. The cylindrical pin support 106a is hollow and forms an aperture having a large diameter to receive the expansion spring. Accordingly, the cylindrical pin support 106a and the corresponding opening 101 in the debris shield may have larger diameters than do the cylindrical pin support 106 (FIG. 3A) for the water rod and the corresponding opening 101 for the pin support 106. When the upper tie plate 100 is removed, the expansion spring 116 no longer abuts or is disconnected from the ledge of the upper tie plate. The fuel rod and expansion spring may be lifted up through the debris shield to be removed from the fuel assembly. FIGS. 5 and 6 are perspective views of the side and top of an alternative upper tie plate assembly 130 having a horizontal support frame 132 with a horizontal slot 134 to receive a debris shield 136 that can be either a removable unit or a permanent integrated structure within the upper tie plate. The debris shield shown in FIG. 5 is a removable debris shield and is shown partially removed from the horizontal support frame 132 in that figure. During operation the debris shield 136 is fully inserted into the slot 134 and enclosed by the horizontal support frame 132, as is shown in FIG. 6. The upper tie plate assembly 130 includes an upper tie plate handle 22 that is attached or integral with the horizontal support frame. The horizontal support frame 132 may include a rectangular outer support wall structure having hollow ribs 138, which may receive tie rod end couplings, e.g., threaded pins of tie rods 24 and provides backing for the expansion springs 16 of tie rods. The horizontal support frame may be porous, e.g., have small vertical openings 131 to allow fluid to pass through the horizontal support frame and block the passage of foreign debris. The openings in the horizontal support frame increase the effective flow area of passages through the upper tie plate 130 and thereby compensate or offset any flow restrictions due to the debris shield 136. Datum points 117 attached to the horizontal support frame 132 that center the hollow metal channel 20 with the upper tie plate and are used during fabrication to square the upper tie plate 130 to the full-length fuel rods 11, the part length fuel rods 12, the tie rods 24, and the water rods 23 and 230. FIG. 7 is a side view of the upper tie plate assembly 130 having a horizontal support slot 134 and the debris shield plate 136 in the slot. The debris shield 136 may be a perforated flat plate having a wavy cross-sectional shape that results in openings in the debris shield that slant 143 with respect to the vertical axis of the nuclear fuel bundle assembly 10. The perforations may be machined, e.g., electrical discharge machining, punched, drilled or formed by casting. The perforations 143 are small to prevent the passage of foreign debris. The perforations 143 are slanted due to the wavy cross-sectional shape of the debris shield 136. The slanted perforations 143 aid in the capture, the blocking and deflect captures debris, and particularly debris flowing axially with respect to the fuel bundle assembly 10. The wavy shape of the debris shield 136 also assists in securing the debris shield in the slot 134 by causing the upper ridges and lower grooves 133 of the debris shield 136 to be biased against the upper and lower surfaces of the slot 134, when the debris shield is position within the cavity of the upper tie plate. Further, the upper ridges and lower grooves 133 may be solid and have no perforations at their upper and lower apexes. The ridges and grooves 133 may be devoid of coolant flow openings so as to avoid vertically aligned perforations that may tend to pass foreign debris particles that would be blocked by the slanted openings 143 on the angled sides 137 of the upper ridges and lower grooves 133. Further, the solid groove forms a V-shaped channel 139 to capture and hold foreign debris. The debris may remain in the V-shaped channel until the debris shield 136 is removed from the upper tie plate 130. Foreign particulates may become lodged or trapped within the debris shield 136, because of the upper end plugs of fuel rods and water rods 19, 24 and 30, sticking through the solid portions of the V-shaped grooves and ridges 133 on the debris shield 136. A lower pressure gradient is created by coolant flowing through the opening from below and through the slanted openings 143 on the angled sides 137 of the debris shield 136. Upper end plugs first pass through the lower matrices 158 of the upper tie plate, then through the debris shield 136, where the solid V-shaped grooves and ridges are located within the channel 139. Next the upper end plugs are allowed to pass through the upper matrices 156 of the upper tie plate, where they may be are terminated. When the debris shield is removed, the debris captured within the V-shaped channel may be washed off, vacuumed, and/or vibrating from the debris shield, the debris shield 136 is thereafter reinserted once again back within the upper tie plate. Alternatively, the debris shield 136 with the captured debris may be discarded and a new debris shield re-inserted into the upper tie plate. FIGS. 8 and 9 is a top-down view of the upper tie plate assembly 130. The metal matrices 140 of pin supports 141 and ribs 142 interconnecting the pin supports and ribs 144 connecting the horizontal support frame 132 to the matrices are clearly shown in FIG. 8. The matrices 140 may include an upper planar matrix 156 of pin supports and interconnecting ribs, and a lower planar matrix 158 of pin supports and interconnecting ribs. The upper and lower planer matrix may have identical patterns of pin supports and ribs. The upper and lower planar matrices 156, 158 define an upper and lower surface of the slot 134 that receives the debris shield 136. Alternatively, the matrices 140 may also be a single planar matrix that is arranged immediately above or below the debris shield 136. The metal matrices 140 are mounted to the horizontal support frame 132 and extend over an open area inside of the horizontal support frame. The metal matrices 140 may be formed by the machining of a metal casting, casting, or the complete machining of the upper tie plate assembly. The matrix 140 may be attached to the horizontal support frame and to each other by spokes, rods or a matrix structure (collectively matrix structure). The pin supports 141 in the metal matrices 140 include cylindrical supports 141 to receive fuel rod end couplings, e.g., threaded pins of tie rods 24 and expansion springs 16 of fuel rods 11 and the upper end plugs of the water rods 23 and 230. The matrices 140 may include solid metal braces 145 especially near the corners and below the supports for the handle 22. FIG. 9 is a top-down view of the debris shield plate 136. The pin supports 135 for the threaded ends 28 of the tie rods and the upper end plugs for fuel rods 11, and the water rods 23 and 230, are arranged to align with the pin support cylinders 141 in the upper matrices 156 and lower matrices 158 of the upper tie plate assembly 130. The pin supports are substantially larger than the flow passages through the debris shield 136. The pin supports are filled with the upper end plugs 19, 30 and 28 that are attached to the ends of the water rods, 23, 230 and the fuel rods 11 and 24, that by themselves do little to limit fluid passage around the pin supports 141. However, once the debris shield 136 has been installed within the upper tie plate 130 cavity, fluid leakage through each of the matrices are restricted around each of the pin support, for fluid leakage. The angled sides 137 of the upper ridges and lower grooves 133 contain the pin supports 135 in the debris shield 136 for receiving the threaded ends 28 of tie rods 24, fuel rods 11, water rods 23, 230 and upper end plugs for each type of pin design. These pin supports 135 will be filled by the rod's upper end plugs, which block foreign debris and fluids that might otherwise may flow through the pin supports. FIG. 10 shows a third embodiment of an upper tie plate assembly 150 as a honey comb design, having a removable debris shield 152 that slides in a horizontal slot 154 between an upper planar section 156 and a lower planar section 158 of the upper tie plate assembly 150. The debris shield 152 may be a removable unit or a permanent integrated structure in the upper tie plate. In FIG. 10, the debris shield 152 is shown being inserted into the slot 154 of the upper tie plate. FIG. 11 shows the debris shield 152 fully inserted and secured within the upper tie plate assembly 150. FIG. 12 is the side view of the debris filter 152 fully inserted into the upper tie plate cavity 154. FIGS. 13 and 14 show top and bottom views, respectively, of the upper tie plate assembly 150. A three-sided frame 160 of the upper tie plate assembly 150 holds together the upper and lower planar sections 156, 158. The frame has sidewalls on three of its four sides. The fourth side is open and forms the slot or cavity 154 for the debris shield 152. The frame may be porous, e.g., have small vertical openings 131 to allow fluid to pass through the frame and block the passage of foreign debris. The openings in the frame increase the effective flow area of passages through the upper tie plate assembly 150 and thereby compensate or offset any flow restrictions due to the debris shield 152 once the debris shield is placed within the slot or cavity 154 of the upper tie plate. Datum points 117, e.g., vertical outer ribs, attached to the three sided horizontal support frame 160 centers the hollow metal channel 20 with the upper tie plate and are used during fabrication to square the upper tie plate 150 assembly to the fuel rods and the water rods. The horizontal support frame 160 of the upper tie plate, supports a lifting handle 22 and the channel posts 25 and 26. A channel fastener clips (not shown) are used to secure the hollow metal channel 20 to the threaded channel post 26. The horizontal support frame 160 includes upper and lower planar sections 156, 158 that are load bearing structures of the upper tie plate 150, and provide structural support for the debris shield 152, hollow metal channel 20, tie rod 24 and other components of the nuclear fuel bundle assembly 10. Optionally, a structurally strong debris shield 152 may serve as a load bearing structure and replace one or more of the upper and lower planar sections 156, 158 and part or all of the horizontal support frame 160. The debris shield 152 includes pin support apertures 161 to receive the fuel rod 30 and 24, and the upper end plugs 19 of the water rod, for example, and that are aligned with the pin supports 141 of the upper and lower planar sections 156, 158 of the upper tie plate assembly 150. The debris shield 152 may be generally planar and have edges 162 (FIG. 10) that abut the interior walls of the horizontal support frame 160, and upper debris shield surfaces 164 and lower surface 166 that are adjacent interior surfaces of the upper and lower planer sections 156 and 158 of the upper tie plate. The debris shield may slide horizontally into the slot 154. Because of the thickness of the debris shield 152 is nearly the same approximately size the same as the width of the slot 154, a slight force may be needed to place the debris shield fully within the cavity 154 of the upper tie plate assembly 150. The debris shield is slid into the slot 154 and the pin supports 161 of the debris shield 152 are aligned with the pin supports 161 in the upper and lower planar sections 156, 158 of the horizontal support frame 160. After the debris shield 152 is fully inserted within the slot 154, the tie rods 24, the water rods 23 and 230 and the full-length fuel rods 11 upper end plugs, are inserted into the aligned pin supports of the frame and the debris shield 152. Alternatively, the debris shield 152 may be seated in a bottom cavity of the upper tie plate 150, if the horizontal support frame 160 of upper tie plate lacks a lower planar section 158. With the alternative upper tie plate, the tie rods and end plugs may be inserted in the pin supports 161 of the debris shield 152 before the upper tie plate is placed on the fuel bundle 32. Further, the upper tie plate may be removed from the fuel bundle 32, while the fuel rod 11, and the water rods 23 and 230, and the tie rods 24 are left attached, or secured to the debris shield 152. The debris shield 152 may be a honey-combed metallic structure, a wire or fabric mesh, sponge, grid, array of crossing bars or slats, or other matrix that is porous. The material forming the shield should withstand service in a nuclear reactor core. Preferably, the passages through the debris shield are not entirely straight and include at least one bend or curve. Bends and curves in passages of the debris shield tend to trap debris, especially strands of chips, wires and rods. The passages through the debris shield 152 may be numerous to minimize any fluid pressure drop across the debris shield, while maintaining the debris screening, or filtering function of the debris shield. Fluid flows through the disjointed passages, but debris is filtered out of the fluid by the debris shield. A characteristic of the debris shield may be that light does not shine through the debris shield because of the bends and curves in passages. The bends and curves (see FIG. 12) in the flow passages 159 of the debris shield 152 may be formed by laminating two or more layers 168 of debris shield material, e.g., a honey-combed metallic layer, such that the passages in each layer are not aligned. By way of example, the passages in each layer may be angled, e.g., 5 degrees to 45 degrees, with respect to the axis of the fuel bundle assembly 10. The direction or slope of the passage angles in each layer 164, 166 may be different so that the layers form disjointed passages 159 through the laminated debris shield 152. Alternatively, the debris passages through each layer of the debris shield may be offset with respect to the passages of adjacent layers and gaps between the layers to allow fluid to pass through the debris shield with relatively small fluid resistance. The debris shield 102, 136 and 152 may remain in the fuel bundle assembly 10 during operation of the nuclear reactor core. The debris shield preferably has a porosity that allows coolant, especially emergency coolant, to flow downward through the debris shield without substantial flow resistance. The porosity and disjointed fluid passages of the debris shield blocks the passage of debris. The debris shield serves as a screening device or a filter that allows passage of fluids, such as cooling fluid, and blocks the passage of particulates. Preferably the debris shield should block the passage of particles of debris material having a pore size that minimizes the size of the debris while maintaining the optimal flow of coolant. The debris shields 102, 136 and 152 shown herein are exemplary shields. The debris shield 102 may be configured as a plate, having a chevron porous structure in cross-section, and integral to the upper tie plate. The debris shield 102 is a load bearing structure that includes apertures to receive and support tie rods, fuel rods and the upper end plugs of a fuel bundle assembly 10. Because of the load bearing debris shield 102, upper and lower planer support structures are not needed. The debris shield 136 may be formed of one or more layers of porous metallic layers. The layers may have a wavy cross-sectional shape that imparts a slant to the passages in the shield and thereby may improve the filtering function of the shield. The debris shield 136, as shown, is not load bearing and is inserted in a slot of a load bearing frame of the upper tie plate. The debris shield 152 may be a laminated plate having layers 168 and flow passages 159 that are disjointed and have bends and curves to trap debris. The debris shield 152 may be a non-load bearing and has apertures through which pass tie rods, fuel rods and upper end plugs that attached to openings in upper and lower planer sections of the upper tie plate. Alternatively, the debris shield 152 may be load bearing, having pin support apertures and seats in a bottom cavity of a frame 160 of an upper tie plate having an upper (but not lower) planer matrix of pin supports and ribs, such as shown in FIGS. 2, 3 and 4. The debris shields 102, 136 and 152 block downwardly flowing debris, have relatively little resistive area to emergency cooling flow and allow recirculation of fluid flowing through and around the shield to the top of the bundle during application of the emergency core cooling system. Debris shields having other shapes, compositions and arrangements in the top of a fuel bundle assembly 10 may be fashioned to serve the function of preventing debris falling into a bundle, in substantially the same way of blocking passage of debris falling downward into the bundle while passing coolant, to achieve the result of substantially no debris being introduced in the bundle due to debris falling down past the upper tie plate. Each of the three embodiments of the debris shield 102, 136 and 152 are suitable for blocking screening, and filtering debris from coolant flow passing through the upper tie plate. The debris shield 102 (first embodiment) may be held in place by, for example, one or more of the following methods: (i) threaded upper end plugs 28 of the tie rods 24, (ii) the water rods 23, 230 upper end plugs 19, (iii) the full length fuel rods 11 upper end plugs 30 within the fuel bundle assembly 10, and (iv) a binding force exerted between the cavity opening in the upper tie plate for the debris shield and the debris shield itself. Further, the debris shield 102 may be a removable unit or a permanent integrated structure mounted within the upper tie plate. In addition, the debris shield 102 may be a removable unit or a permanent integrated structure in the upper tie plate and above the fuel rods, wherein the debris shield has a surface at least co-extensive with an open area of the fuel bundle inside the hollow metal channel 20. The debris shields 102, 136 and 152 may be used in a method to prevent debris falling into a nuclear reactor fuel bundle assembly 10, including a bundle of fuel rods mounted below an upper tie plate and housed in a hollow metal channel 20, the method comprising: inserting a debris shield which can be either a removable unit, or attached as a permanent integrated structure within the upper tie plate. Similarly, a method has been developed of maintaining the debris shield in the upper tie plate and above the fuel rods, while the fuel bundle assembly 10 is in an operating nuclear reactor core; flowing coolant through the bundle and the debris shield during operation of the nuclear reactor core, capturing screening or deflecting and the removal of debris falling in the fuel bundle assembly with the debris shield. The debris shields 102, 136 and 152 deflect, catch or remove foreign materials potentially introduced into the top of the fuel assembly. The debris shields 102, 136 and 152 may themselves provide structural support for the bundle and thereby render unnecessary a separate array of pin supports in the upper tie plate. Accordingly, the upper tie plate may comprise a rigid debris shield to which is attached the fuel rods, water rods and possible tie rods, wherein the upper tie plate does not include a conventional pin support array. Coolant flows through the debris shield 102, 136 and 152 is preferably without substantial pressure loss across the debris shield. To reduce the flow resistance of the debris shield, the flow passages may be relatively wide but slanted. The slanting of the passages in the debris shield enhances the ability of the shield to trap, screen or filter debris. Due to slanting, small debris particles cannot flow directly through the shield and will tend to become trapped or lodged within the shield. The view through the debris shield is blocked because the non line-of-sight passages in the debris shield prevent a top-down line-of-sight view through the upper tie plate assembly. The debris shield 102, 136 and 152 serves as a screen or as a filter that allows passage of fluids, such as coolant, and blocks the passage of particulates. The debris shield preferably has a porosity that allows coolant, especially emergency coolant, to flow downward through the debris shield without substantial flow resistance. Preferably the debris shield should have a pore size that minimizes the size of the debris without imposing a significant fluid pressure drop across the debris shield. The debris shield 102, 136 and 152 may be formed of a porous material, such as a wire or fabric mesh, sponge, grid, array of crossing bars or slats, or other matrix material. The passages of the debris shield are preferably too narrow to allow larger debris particles to enter the passages. Further, the frame of the upper tie plate may be porous, e.g., have small vertical openings to allow fluid to pass through the frame and block passage of debris. The openings in the frame increase the effective flow area of passages through the upper tie plate and thereby compensate or offset any flow restrictions due to the debris shield. The debris shield 136 may have a wavy cross-sectional shape that results in openings in the shield that slant with respect to the vertical axis of the fuel bundle assembly 10. The slant in the opening is advantageous in blocking and trapping debris. The wavy shape of the debris shield 136 assists in securing the debris shield in the slot by causing the upper ridges and lower grooves of the debris shield to be biased against the upper and lower surfaces of the slot. The upper ridges and lower grooves 133 of the debris shield contain the apertures for receiving the threaded ends of tie rods, fuel rods, water rods and upper end plugs for each type. The debris shield 136 preferably has porosity in the sides between the upper and lower most portions of the grooves and ridges that allows coolant, especially emergency coolant, to flow downward through the fuel bundle assembly, while retaining the captured debris that's fallen from above the fuel bundle assembly 10. A structurally strong debris shield 152 may serve as a load bearing structure and replace one or more of the upper and lower sections of the frame. The debris shield 152 may be a honey-combed metallic structure, a wire or fabric mesh, sponge, grid, array of crossing bars or slats, or other matrix that is porous. The passages through the debris shield may be numerous to minimize any fluid pressure drop across the debris shield, while maintaining the debris screening and filtering functions of the debris shield. FIGS. 15a, 15b and 15c are schematic diagrams showing a maintenance or fuel inspection pool 180 to receive fuel bundle assemblies 10 that have been removed from the reactor core 21 for maintenance and service. Coolant/moderator 182 generally covers the fuel bundle assemblies in the reactor core and maintenance or fuel inspection pool. As shown in FIG. 15a, a crane 184 grasps the handle of a fuel bundle handle and lifts the fuel bundle assembly 10 from the reactor core. The crane 184 moves the fuel bundle assembly, such as along an overhead beam 186, to the maintenance or fuel inspection pool 180. The crane lowers the fuel bundle assembly in the maintenance or fuel inspection pool. As shown in FIG. 15b, the hollow metal channel 20 has been removed and the fuel bundle 32 is ready for inspection and service in the maintenance or fuel inspection pool 180. The debris shield 152 may be removed from the upper tie plate by sliding the debris shield horizontally to remove it from a slot or cavity 154 in the upper tie plate. Once removed, the debris shield may be cleaned and inspected and reinserted into the slot of the upper tie plate. Alternatively, the debris shield may be replaced with a new debris shield which is inserted into the slot of the upper tie plate. As shown in FIG. 15C, the upper tie plate may be removed from the fuel bundle assembly while the debris shield 102 remains attached to the upper portion of the fuel bundle 32 and the channel is removed. The upper tie plate may be replaced or serviced and then reinstalled over the debris shield and onto the fuel bundle 32. After the upper tie plate has been removed, the debris shield may then be taken off the fuel bundle 32 by removing the hex nuts that secure the water rods 23 and 230 upper threaded end plugs to the debris shield. Once removed from the fuel bundle, the debris shield can be cleaned to remove debris trapped in the shield. Alternatively, the debris shields may be replaced with a new debris shield. Further, after the upper tie plate has been removed and while the debris shield 102 remains attached to the water rods 23 and 230, of the fuel bundle 32, one or more fuel rods 11, 12 may be removed through openings in the debris shield. Specifically, the upper end of the fuel rod is inserted in a cylindrical pin support 106 (FIG. 3) that is seated in an opening of the debris shield. The opening in the debris shield is sufficiently wide to pass a fuel rod and the expansion spring. The cylindrical pin supports 106 may be securely seated in the debris shield by an upper grid matrix of pin supports in the upper tie plate or by lock washers, hex nuts 17 or other securing mechanisms that releasably secure the cylindrical pin supports to an upper surface of the debris shield. The cylindrical pin supports allow an upper end of the fuel rod to extend through the support and may provide an upper end stop for the expansion springs on the fuel rods. To remove the fuel rod 11, 12, the upper end of the fuel rod is grasped and pulled upward such that fuel rod, cylindrical pin support and expansion spring moves as an assembly up through the opening in the debris shield 102. A single fuel rod may be removed from the fuel bundle while debris shield is covering and supporting the fuel bundle, once the upper tie plate has been removed. Similarly, an assembly of a fuel rod, expansion spring and cylindrical pin support may be inserted in the fuel bundle by lowering the assembly through the opening in the debris shield such that the cylindrical pin support seats in the opening the debris shield. The debris collected by the debris shield may be inspected to detect potential problems in the nuclear reactor core 21. The debris captured by the debris shield potentially may be broken pieces of components from the nuclear reactor core, lose components, metal shavings produced by rubbing of components and other fragments. Inspection by operates of the debris from the debris shield may indicate a problem in the nuclear reactor core. Based on the inspection, the operator may determine or at least suspect that a component of the nuclear reactor core may require repair, replacement or at least further inspection. Accordingly, inspection of the debris from the debris shield may be performed while or after the debris shield is removed from the upper tie plate and fuel bundle 32. The cleaned or new debris shield is applied to the fuel bundle in much the same manner that it was removed. The hollow metal channel 20 is reattached to the fuel bundle 32. The completed fuel bundle assembly is lifted by the crane from the maintenance or fuel inspection pool 180 and placed in its proper location in the array of fuel bundle assemblies in the reactor core. The fuel bundle assembly may not be returned to its original location. Rather, the fuel bundle assembly 10 may be returned to a different location in the reactor core, where the location is determined based on reactor performance. Removal of a single fuel rod, as is describe above, provides the ability to remove one or more fuel rods from a fuel bundle without removing the debris shield or other components of the fuel bundle, except for the upper tie plate. The upper tie plate and individual fuel rods can be repaired or replaced without completely disassembling the fuel bundle. Accordingly, repairing and replacing upper tie plates and individual fuel rods may be preformed while the fuel bundle is in the maintenance/inspection pool. Performing these repairs and replacements in the maintenance/inspection pool and without disassembling the entire fuel bundle saves time and money by reducing the amount of work required to service the upper tie plate or replace individual fuel rods. While the invention has been described in connection with what is presently considered to be the most practical and preferred embodiment, it is to be understood that the invention is not to be limited to the disclosed embodiment, but on the contrary, is intended to cover various modifications and equivalent arrangements included within the spirit and scope of the appended claims.
062158368	claims	1. A method of ultrasonically examining a remotely located weld in a nuclear steam supply system, said method comprising the steps of: positioning an ultrasonic transducer in a collapsible shoe to produce longitudinal ultrasonic waves between approximately 45.degree. and 60.degree.; moving said collapsible shoe through a pipe segment, said collapsible shoe maintaining ultrasonic coupling between said ultrasonic transducer and said pipe segment; returning an ultrasonic signal from said ultrasonic transducer having a sound path indicative of the geometry of said remotely located weld, said ultrasonic signal including maximum amplitude signals; evaluating only said maximum amplitude signals having sound paths of between approximately 0.25" and 0.95". positioning an ultrasonic transducer in a collapsible shoe to produce ultrasonic shear waves at 45.degree.; moving said collapsible shoe through a pipe segment, said collapsible shoe maintaining ultrasonic coupling between said ultrasonic transducer and said pipe segment; returning an ultrasonic signal from said ultrasonic transducer having a sound path indicative of the geometry of said remotely located weld, said ultrasonic signal including maximum amplitude signals; evaluating only said maximum amplitude signals having sound paths between approximately 0.25" and 0.65". 2. A method of ultrasonically examining a remotely located weld in a nuclear steam supply system, said method comprising the steps of:
	summary
052992519	summary	BACKGROUND OF THE INVENTION 1. Field of the Invention This invention relates to an exposure apparatus for producing an Ultra Large Scale Integrated Circuit (ULSI), a liquid crystal display panel and so on. In particular, the invention relates to an alignment system to adjust positions of masks and targets to be exposed. 2. Description of the Related Art Light exposure apparatus have been applied to produce large scale integrated circuits. However, as the patterns of the circuits get finer and finer, it becomes harder to expose such circuit patterns by the light exposure apparatus because of resolution and productivity requirements. Recently X-ray exposure apparatus have been developed instead of light exposure apparatus because the X-ray exposure apparatus can transfer finer circuit patterns than can light exposure apparatus. In the case of such exposure by the X-ray apparatus, it is very difficult to transfer reduced patterns from a mask to a target such as a semiconductor wafer. Therefore, the patterns have to be transferred isometrically in an exposure process. In the case of isometric exposure, the gap between the mask and the wafer should have a minimum space of 10-50 .mu.m. Such exposure is called a proximity exposure. A general X-ray exposure apparatus used to expose a semiconductor wafer is shown in FIG. 1. FIG. 1 is a schematic vertical sectional view. In FIG. 1 X-ray 51 from syncrotron orbital radiation (SOR) is reflected on an X-ray reflect mirror (not shown) and passes through a port 52 kept in high vacuum pressure. After that, the X-ray passes through a window 53 and a bellow 54, then into a chamber 55. In the chamber 55 there re alignment systems 58 which detect the positions and distance between the X-ray mask 56 and the semiconductor wafer 60 by LASER beam 61 as a detecting beam. Behind the chamber 55 there is a mask stage 57 which holds and moves a X-ray mask 56 and a wafer stage 59 which holds and moves a semiconductor wafer 60. In this structure, the X-ray 51 passing through the window 53 into the chamber 55 is exposed on the X-ray mask 56 and transfers patterns on the X-ray mask to the semiconductor wafer 60 which is spaced apart from the X-ray mask by dozens of .mu.m. The exposure process as described above is repeated several times. Therefore, the positions and the distance between the X-ray mask 56 and the semiconductor wafer 60 must be detected in each exposure process and be adjusted to the right position precisely. The alignment system 58 is applied to the detection described above. A conventional structure of such alignment systems are shown in FIG. 2 and FIG. 3. Alignment marks are written on both an X-ray mask and a semiconductor wafer. The alignment system takes the picture data of these alignment marks and gets the information about the relative distance between the mask and the wafer by the processing of the picture data. An explanation of the alignment marks written on the X-ray mask is described hereunder. Three alignment marks (X mark 6a, Y mark 6b, .theta. mark 6c shown in FIG. 2 and FIG. 3) are written on an exposed area in the X-ray mask 6. Three optical alignment systems are installed in the alignment systems, the optical alignment system 11a for detecting the X mark position, the optical alignment system 11b for detecting the Y mark position and the optical alignment system 11c for detecting the .theta. mark position. As shown in FIG. 2 and FIG. 3, these optical alignment systems 11a, 11b, 11c, comprise objective lenses (12a, 12b, 12c), three prisms (13a, 13b, 13c), and half mirrors and reflective mirrors (not shown in FIG. 2 and FIG. 3). A detection beam passes through the half mirror and the objective lens and goes to the prism. The path of the beam is bent 90 degrees by the prism. After reflection by the prism the beam reaches the alignment mark. Then, the beam is reflected by the alignment mark and passes back the same way, i.e., through the prism and the objective lens. The beam passed through the objective lens reflects on the half mirror and the reflective mirror. The beam reflected on &he reflective mirror is processed as picture data expressing the position of the alignment mark. The positions of the alignment marks 6a, 6b, 6c may change in each exposure processes and also may change depending on the chip sizes. Therefore, the optical alignment systems need to be moved according to the changes of the positions of the alignment marks. The optical alignment systems 11a, 11b and 11c are arranged on moving bases 16a, 16b and 16c respectively. The bases can move in X-Y directions individually. The X-ray exposure apparatus is used in the proximity exposure. All of the alignment marks 6a, 6b, 6c for detecting should be in the exposure area, which is generally sized 10 mm to 30 mm in the shape of a square. The optical alignment systems 11a, 11b, 11c should be arranged over the alignment marks 6a, 6b, 6c, respectively, and they should not interfere with each other in such a narrow space. Therefore, the moving bases 16b and 16c, opposite each other, are arranged on the same plane. However, the moving base 16a is arranged above the moving base 16b and 16c as described in FIG. 2 and FIG. 3. Thus the position of the optical alignment system 11a installed on the moving base 16a is above the other optical alignment systems 11b, 11c. As shown in FIG. 2 and FIG. 3, the distance h.sub.2 from the optical axis of the optical alignment system 11b to the alignment mark 6b is equal to the distance h.sub.3 from the optical axis of the optical alignment system 11c to the alignment mark 6c (h.sub.2 =h.sub.3). However, the distance h.sub.1 from the optical axis of the optical alignment system 11a to the alignment mark 6a is bigger than the distance h.sub.2 or the distance h.sub.3 respectively (h.sub.1 >h.sub.2, h.sub.1 >h.sub.3). When the optical alignment system 11a is not arranged in the same plane in which the optical alignment system 11b and 11c are arranged as described above, the focal length f.sub.1 of the object lens 12a attached in the optical alignment system 11a is different from the focal lengths f.sub.2 and f.sub.3 of the object lenses 12b and 12c attached in the optical alignment system 11b and 11c respectively. The focal length f.sub.2 is equal to the sum of the length h.sub.2 and the length h.sub.2 ' which is the distance from the center of the lens 12b to the alignment mark 6b. The case of the focal length f.sub.3 is as same as f.sub.2. On the other hand, the focal length f.sub.1 is equal to the sum of the length h.sub.1 and the length h1' which is the distance from the center of the lens 12a to the alignment mark 6a. The value of f.sub.1 is not equal to the value of f.sub.2 or f.sub.3 respectively. Therefore, the optical power of lenses are different from each other in these optical alignment systems and many different parts for producing the systems are needed. Also the production cost of these different systems is very high and production takes many processes. The width of X mark 6a, Y mark 6b and .theta. mark 6c are shown in FIG. 2 as c1, c2, and c3, respectively. The value of c1 is not equal to the value of c2 or the value of c3. The value c2 is equal to the value of c3 (c1.noteq.c2=c3). Because the optical power of the lenses are different from others it is required to detect the marks as a same scaled picture with a data processing system. SUMMARY OF THE INVENTION Accordingly, it is an object of the present invention to provide a exposure apparatus which can be composed of common optical parts and protected from the interference of the optical alignment systems with each other in the same plane. To accomplish the above described object, an exposure apparatus is provided which comprises: a light source for exposing; PA1 a mask holding means for alternately holding and releasing a mask; a target holding means for alternately holding and releasing a target; a moving means for moving the mask holding means and the target holding means individually; and, at least two alignment detecting means located in the same plane, the alignment detecting means including a moving base a narrow front portion.
	claims	1. An electron accelerator based method for producing molybdenum-99 (Mo-99) comprising:i) providing an electron accelerator producing a high energy electron beam;ii) providing one molybdenum converter/target unit (Mo-unit) comprising molybdenum-100 (Mo-100) wherein said Mo-unit simultaneously serves both as a braking radiation (bremsstrahlung) converter and a radioisotope production target; andiii) directing said electron beam onto said Mo-unit, thereby producing braking radiation (bremsstrahlung) which subsequently reacts in the same Mo-unit with said Mo-100 via the (γ,n) reaction to produce Mo-99 in said Mo-unit, in which the Mo-99 product accumulates;wherein said Mo-unit further comprises molybdenum-98 (Mo-98), the method further comprising slowing down the neutrons produced in step iii) with a low atomic number liquid and reacting them with said Mo-98 via the (n,γ) reaction to produce additional Mo-99 in said Mo-unit, thereby maximizing the efficiency in the production of Mo-99. 2. The electron accelerator based method of claim 1 for producing Mo-99 and other radioisotopes, further comprising the step of placing one or more external target materials outside the Mo-unit and adjacent to it so that said bremsstrahlung photons and neutrons around said Mo-unit generate further radioactive isotopes via (γ,n) and (n,γ) reactions on said one or more external target materials. 3. The method of claim 2, wherein said external target materials comprise Xe-124, and wherein said further radioactive isotopes are I-123 and I-125. 4. The method of claim 2, wherein said external target materials are selected from F-19, O-16, N-14 and C-12, and wherein said further radioactive isotopes are F-18, O-15, N-13 and C-11, respectively. 5. The method of claim 2, wherein said external target materials comprise low enriched uranium (LEU) which is used in a photo-fission (γ,f) reaction. 6. An electron accelerator based apparatus for producing molybdenum-99 (Mo-99), comprising:a) an electron accelerator producing a high energy electron beam;b) one converter/target unit made from molybdenum (Mo-unit) comprising molybdenum-100 (Mo-100) wherein said Mo-unit serves both as a braking radiation (bremsstrahlung) source and as a radioisotope production target; andc) means for directing said electron beam onto said Mo-unit to produce braking radiation (bremsstrahlung) which subsequently reacts with said Mo-100 via the (γ,n) reaction to produce and accumulate Mo-99 in said Mo-unit;wherein said Mo-unit further comprises molybdenum-98 (Mo-98), the apparatus further comprising a low atomic number liquid which slows down said neutrons produced in the (γ,n) reaction, the neutrons subsequently reacting with said Mo-98 via the (n,γ) reaction to maximize the efficiency in the production of Mo-99. 7. The apparatus of claim 6, wherein the low atomic number liquid is distilled water. 8. The method of claim 1, wherein said Mo-unit comprises natural molybdenum. 9. The method of claim 1, wherein said low atomic number liquid is water. 10. The apparatus of claim 6, wherein said Mo-unit comprises natural molybdenum. 11. The apparatus of claim 6, wherein said low atomic number liquid is water, which serves both for cooling the Mo-100 unit and for slowing down the neutrons.
	description	1. Field of the Invention The present invention relates generally to the fabrication of fuel assembly support grids for nuclear reactors, and more particularly to an improved method of fabricating a fuel assembly grid using a flared sleeve. 2. Description of the Related Art In most pressurized water nuclear reactors, the reactor core is comprised of a large number of elongated fuel assemblies. These fuel assemblies typically include a plurality of fuel rods held in an organized array by a plurality of grids that are spaced axially along the fuel assembly length and are attached to a plurality of elongated thimble tubes of the fuel assembly. The thimble tubes typically receive control rods, plugging devices, or instrumentation therein. Top and bottom nozzles on opposite ends of the fuel assembly are secured to the ends of the thimble tubes that extend slightly above and below the ends of the fuel rods. The grids, as is known in the relevant art, are used to precisely maintain the spacing between the fuel rods in the reactor core, resist rod vibration, provide lateral support for the fuel rods and, to some extent, vertically restrain the rods against longitudinal movement. The grids are typically made of materials such as stainless steel, Inconel, and alloys of Zirconium. One type of conventional grid design includes a plurality of interleaved straps that together form an egg-crate configuration having a plurality of cells which individually accept the fuel rods therein. Depending upon the configuration of the thimble tubes, the thimble tubes can either be received in cells that are sized the same as those that receive fuel rods therein, or can be received in relatively larger thimble cells defined in the interleaved straps. The straps are configured such that the cells each include a plurality of relatively compliant springs and a plurality of relatively rigid dimples, with the springs and dimples being formed into the metal of the interleaved straps and protruding outwardly therefrom. The springs and dimples of each cell engage the respective fuel rod extending through the cell. Outer straps of the grid are attached together and peripherally enclose the inner straps of the grid to impart strength and rigidity to the grid. Because the straps that make up a grid are relatively thin, i.e., on the order of 0.0105 inches in some cases and 0.008 inches in others, they tend to be flexible, making it difficult to securely attach (such as by welding) the lengthy thimble tubes directly to the straps. Instead, it has become common practice to first attach a relatively shorter cylindrical sleeve made of, for example, stainless steel or alloys of Zirconium, to each of the cells of the straps that are to receive a thimble tube by welding the sleeve and straps together. Next, each of the thimble tubes is inserted through the corresponding sleeves and is attached to the sleeves, and thus the grid, by bulging the thimble tubes out at a location either above or below each grid. FIG. 1 is a partial isometric view of a prior art sleeve 5 attached to the straps 10 of a grid. In prior art grid fabrication methods, sleeve 5 is placed on top of and aligned with cell 15 formed by straps 10. Cell 15 is provided with a generally circular shape for mating with sleeve 5 by providing a radius on straps 10 at that point. Sleeve 5 is then butt welded to straps 10 at butt weld joint 20 shown in FIG. 1. The butt welding of sleeve 5 to straps 10 has proven to be problematic primarily due to the fact that there is normally a gap between sleeve 5 and straps 10 that form cell 15, a condition known as poor sleeve to strap fit-up. This condition is the result of the difficulty in providing cell 15 with a perfectly circular, appropriately sized shape for mating with sleeve 5. This gap makes it difficult to effectively join sleeve 5 and straps 10. In addition, as is known, the straps, such as straps 10, used to construct fuel assembly grids are elongated sheets of metal. Each sheet is formed with a plurality of parallel slots that extend transversely down approximately one half of the strap. For hexagonally shaped grids made with three sets of interconnected straps, the first set of aligned straps has all of its slots extending downwardly from the top edge of the strap, the second set of straps has all of its slots extending upwardly from the bottom edge of the strap, and the third set of straps has half of its slots extending downwardly from the top edge of the strap and half of its straps extending upwardly from the bottom edge of the strap in an alternating fashion. As is known, this grid is constructed by inserting the slots of certain of the straps into the slots of certain other of the straps to form an interconnecting lattice pattern. The result is a number of what are known as loose straps. Loose straps are the portions of the straps adjacent to each slot which, because of the slot, remain free to move after the straps are interconnected with one another. In order to prevent any potential strap or spring to rod fretting resulting from the movement of the loose straps, the loose straps are fixed in place by attaching them either to the solid portion of an adjacent strap or to a sleeve in the case of loose straps that are positioned adjacent to a sleeve. In the prior art, loose straps are so attached by laser seam welding a weld tab, commonly referred to as a sail, provided on the straps to the adjacent strap or sleeve, as the case may be. This is shown in FIG. 1 at weld seam 25. Again, problems have arisen due to the fact that gaps between each sleeve and the adjacent loose straps make it difficult to effectively join the two materials. Another problem that exists with prior art sleeves, such as sleeve 5 shown in FIG. 1, and prior art grid fabrication methods is that, due to poor sleeve to strap fit-up, the straps forming a cell for receiving a thimble tube will often encroach on the path of the thimble tube (the straps do not form a perfect circle such that some portions may extend inside the circumference of the sleeve), thus causing difficulty during thimble tube loading. In addition, in the prior art, thimble tube loading has also been adversely impacted as a result of the weld heat affected zone entering the inside diameter of the straps and sleeve, causing the internal diameter thereof to become smaller after grid welding due to the shrinkage of the material in that area. Thus, there is a need for a grid fabrication method that alleviates sleeve to grid attachment problems and thimble tube loading problems caused by poor sleeve to strap fit-up. The present invention relates to a method of fabricating a grid of a fuel assembly of a nuclear reactor. The method includes providing a plurality of interconnected straps that form a lattice pattern. The lattice pattern of the straps defines a plurality of cells. The method further includes providing a sleeve that has a cylindrical portion and a flared portion and inserting the sleeve into at least one of the cells. When inserted in the cell, at least a portion of the cylindrical portion of the sleeve will reside inside the cell and the flared portion will extend above the top end of the cell and overhang a perimeter of the cell. The flared portion is melted and the melted material flows over and fuses to the straps that define the cell. The straps that define the cell may include weld tabs over which the melted material flows and to which it fuses. The melting of the flared portion is done by directing a heat source, preferably a laser beam, at the flared portion. The melting and fusing steps also preferably cause any loose straps surrounding the cell to become attached to the sleeve. FIGS. 2 and 3 show isometric and side elevational views, respectively, of sleeve 30 utilized in a grid fabrication process according to the present invention. Sleeve 30 is preferably made of stainless steel or an alloy of Zirconium and includes cylindrical portion 35 at a first end thereof and flared portion 40 at a second end thereof. Flared portion 40 consists of a portion of sleeve 30 that expands outwardly in shape as compared to cylindrical portion 35 such that it has a cross-sectional diameter at any point thereof that is greater than the cross-sectional diameter of cylindrical portion 35. Flared portion 40 may be a portion that gradually increases in diameter from the bottom to the top thereof in a curved fashion as shown in FIGS. 2 and 3. Alternatively, the change in diameter may be abrupt such that flared portion 40 is a larger diameter plate or cylindrical piece that sits directly on top of cylindrical portion 35. According to the present invention, sleeve 30 is attached to a grid by first inserting sleeve 30 into cell 45 formed by straps 50 as shown in FIG. 4. As seen in FIG. 4, each of straps 50 preferably includes at least one weld tab 55 such that weld tabs 55 are located at various locations around the perimeter of cell 45. When sleeve 30 is inserted into cell 45, at least a portion of cylindrical portion 35 resides inside cell 45 and flared portion 40 extends above the top end of cell 45 and overhangs the outer perimeter of cell 45, and, preferably, weld tabs 55 such that flared portion 40 is located above weld tabs 55. Next, a heat source, such as a laser beam used in laser seam welding, is directed at flared portion 40, causing the material of flared portion 40 to melt and flow over weld tabs 55 (or alternatively some other portion of straps 50 if weld tabs 55 are omitted) and become fused thereto, thereby connecting sleeve 30 to straps 50 and at the same time attaching the loose straps 50′ surrounding cell 45 to sleeve 30 such that they are no longer free to move. Thus, according to the present invention, a single welding step replaces two separate welding steps (the butt weld and the loose strap to sail weld) of the prior art. In addition, due to the flared nature of flared portion 40 and the fact that it overhangs the outer perimeter of cell 45 and, preferably, weld tabs 55, the attachment step of the present invention compensates for and is not adversely effected by any gaps that are the result of poor sleeve to strap fit-up. Further, because sleeve 30 is inserted into, and preferably entirely through, cell 45, a thimble tube will never come into contact with straps 50 during thimble tube loading (sleeve 30 provides a good interface for the thimble tubes). Finally, because of the flared nature of flared potion 40, the heat affected zone from the welding does not reach the internal diameter of strap 50 or sleeve 30, preventing the shrinkage problem presented by the prior art. 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 claims appended and any and all equivalents thereof.
054224926	summary	BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates, in general, to the monitoring of the criticality of nuclear fuel in fuel manufacturing or handling facilities, and in particular, to a new and useful device for calibrating a criticality detector. 2. Description of Related Art In nuclear fuel manufacturing plants, criticality monitors and detectors are used to monitor the criticality of the nuclear fuel. To ensure correct operation, these systems have to be calibrated periodically in order to maintain proper adjustments to the electronics of the system. Presently, the calibration procedures for these criticality monitors are conducted by hand. This procedure includes having an operator attach a radiation source to a hand-held rod near the criticality detector and measure the distance from the detector to the radiation source with a tape measure. In performing this calibration procedure by hand, the operator is usually required to stand on a ladder in order to remove the criticality detector or hold the radiation source for measuring. Not only is this procedure not very accurate for calibration purposes, but it is also hazardous to the operator due to the exposure to the radiation source, and his position on the ladder. Presently, there is no known device for calibrating a criticality detector which is accurate, safe to a user and cost efficient. SUMMARY OF THE INVENTION The present invention pertains to a calibrating device for a criticality monitoring system for a nuclear fuel manufacturing plant which allows for a more accurate, safe and cost-effective method of calibrating a criticality detector over known methods. The present invention comprises a radiation protective holder for containing a radiation source which aligns the radiation source with the criticality detector. A measuring rod having a plurality of graduations or calibrations is aligned parallel with the axis of the holder. A nylon guide is connected to the holder and movably connected to the measuring rod for moving the holder and the radiation source along the axis for detection by the criticality detector. The nylon guide is aligned in the same horizontal plane as the radiation source for indicating a calibration measurement. It is an object of the present invention to provide a calibration device for a criticality monitoring system for a nuclear fuel manufacturing plant which is accurate, safe to a user and cost-effective. The various features of novelty which characterize the invention are pointed out with particularity in the claims annexed to and forming a part of this disclosure. For a better understanding of the invention, its operating advantages and specific objects attained by its uses, reference is made to the accompanying drawings and descriptive matter in which a preferred embodiment of the invention is illustrated.
046541888	claims	1. A method for shielding persons working around a nuclear reactor having a reactor head and a shroud extending upward from the reactor head, comprising: (a) mounting a plurality of swingout arms around the shroud, each swingout arm being pivotable about a respective axis that is substantially vertical and that is fixed with respect to the shroud; (b) positioning a shielding member adjacent a swingout arm with a hoist; (c) pivoting the swingout arm horizontally away from the shroud and toward the hoist; (d) transferring the shielding member from the hoist to the swingout arm so that the swingout arm supports the shielding member; (e) pivoting the swingout arm horizontally back toward the shroud; and (f) repeating steps (b) through (c) until the shroud is substantially surrounded by shielding members. (a) mounting around the shroud a plurality of means for supporting said shielding members, each means for supporting being pivotable about a respective axis that is substantially vertical and that is fixed with respect to the shroud; (b) positioning a shielding member adjacent a means for supporting; (c) mounting the shielding member on the means for supporting so that the shielding member has a first edge positioned further from the shroud than a second edge thereof; (d) rotating the shielding member horizontally toward the shroud by moving the first edge toward the shroud; and (e) repeating steps (b) through (c) until the shroud is substantially surrounded by shielding members. a plurality of swingout arms each having a first end and a second end; means disposed around the shroud for mounting said swingout arms at the first ends thereof, each swingout arm being pivotable about a respective axis that is substantially vertical and that is fixed with respect to the shroud, the second ends of said swingout arms being horizontally movable between first positions that are spaced apart from the shroud and second positions adjacent the shroud; a plurality of shielding members; and means for hanging at least one shielding member to each swingout arm. 2. The method of claim 1, wherein step (f) is conducted by surrounding the shroud with shielding members having edges which overlap. 3. The method of claim 1, wherein step (f) is conducted by surrounding the shroud with shielding members in the form of lead panels with edges which overlap. 4. The method of claim 3, wherein said hoist is a stud tensioner hoist having a hook, wherein each swingout arm has at least one hanger attached thereto, wherein step (b) comprises lifting the panel with the stud tensioner hoist hook via a suspension member attached to the panel, said suspension member including a first element and a second element which are affixed to one another and which are rotatable with respect to said panel, the stud tensioner hoist hook engaging the first element as step (b) is conducted, and wherein step (d) comprises hanging the panel substantially vertically from the first element while the second element projects toward the hanger, inserting the second element in the hanger, removing the stud tensioning hoist hook from the first element, and supporting the panel substantially vertically from the second element. 5. The method of claim 1, wherein step (a) comprises mounting a shield support around the shroud, said shield support including a permanent rail encircling the shroud, said swingout arm being pivotably attached to said permanent rail. 6. The method of claim 5, wherein the shroud includes a ring of vertically disposed beams, and wherein the step of mounting a shield support comprises dividing the permanent rail into a plurality of sections, affixing brackets to the sections, and affixing the brackets to the beams. 7. A method according to claim 6, wherein the brackets have flanges with bores therein, and wherein the step of affixing the brackets to the beams comprises aligning the brackets with the beams, drilling the beams through the bores, and bolting the flanges to the beams. 8. A method according to claim 6, wherein the shroud has doors, and wherein the step of mounting a shield support is conducted by positioning the shield support so that the swingout arms allow access to the doors when the shielding members are hung. 9. A method of claim 5, wherein the reactor has vertically disposed lifting rods affixed thereto, and wherein the step of mounting a shield support comprises drilling bores through the permanent rail at positions corresponding to the lifting rods, threading the lifting rods through the bores, and lowering the shield support. 10. The method of claim 9, wherein the step of mounting a shield support further comprises affixing sleeves around the lower portions of the rods, said sleeves having ends, and lowering the shield support until the permanent rail is supported at the ends of the sleeves. 11. The method of claim 10, wherein the step of affixing sleeves comprises positioning a first elongated sleeve having a C-shaped cross section adjacent each lifting rod, positioning a second elongated sleeve having a C-shaped cross section adjacent the first sleeve and opposite thereto, in clamping the first and second sleeves together. 12. The method of claim 11, wherein the step of clamping the sleeves together comprises placing a U-shaped element around both sleeves, and bolting a retaining member to the U-shaped element so that the U-shaped element and the retaining member together encircle the sleeves. 13. A method for installing shielding members around a reactor shroud, said shielding members having edges, comprising: 14. The method of claim 13, wherein step (e) is conducted by surrounding the shroud with shielding members so that their edges overlap, the second edge of each panel being positioned between the shroud and an overlapping first edge of an adjacent shielding member. 15. The method of claim 14, wherein step (a) is conducted by mounting a shield support around the shroud, said shield support including a permanent rail encircling the shroud and a plurality of swingout arms pivotably attached to the permanent rail, wherein step (b) is conducted with a hoist, wherein step (c) comprises pivoting the swingout arm away from the shroud and toward the hoist and transferring the shielding member from the hoist to the swingout arm, and wherein step (d) is conducted by pivoting the swingout arm back toward the shroud. 16. In a nuclear reactor system having a reactor head and a shroud extending upward from the reactor head, an apparatus for shielding persons working around the reactor head, said apparatus comprising: 17. The apparatus of claim 16, wherein said shielding members are panels, and wherein said means for hanging comprises at least one hanger affixed to each swingout arm and a suspension member affixed to each panel, each said suspension member including a mounting bracket affixed to the panel and a member having first and second extending portions that are disposed at an angle with respect to each other, said member being rotatably mounted on said mounting bracket. 18. The apparatus of claim 17, wherein said means disposed around the shroud comprises a generally hoop-shaped permanent rail, brackets affixing the permanent rail to the shroud, and means for pivotably mounting the first ends of the swingout arms to the permanent rail. 19. The apparatus of claim 17, wherein said reactor head has a plurality of lifting rods attached thereto, said lifting rods being substantially parallel to one another, and wherein said means disposed around the shroud comprises a generally hoop-shaped permanent rail having bores therein, said lifting rods being threaded through said bores, and means for pivotably mounting the first ends of the swingout arms to the permanent rail. 20. The apparatus of claim 19, further comprising sleeves disposed around each lifting rod and between the reactor head and said permanent rail, and means for clamping the sleeves together.
	summary
052971759	description	DESCRIPTION OF THE PREFERRED EMBODIMENT The assembling machine according to an embodiment of the present invention will be described in conjunction with FIG. 1. Incidentally, in the description of the present embodiment, the members which are commonly used in the aforementioned fuel assembly will be designated by the same numerals, hence, detailed description thereof will be omitted. The assembling machine according to the present embodiment is roughly constructed by a base 20, having a rectangular-plate shape, which is provided in an inclined manner to be inclined from one edge side to the other edge side in a longitudinal direction; nine supporting posts 30, respectively disposed on an upper surface of the base 20 in its inclined direction, each of which supports the supporting grid 4 such that the grid cells 5 formed in the supporting grid 4 are inclined along with the inclined direction; a fuel magazine 40, positioned in a relatively upper portion in the inclined direction of the base 20, which holds a plenty of fuel rods 6 in parallel along with the inclined direction in a free-engaged condition (i.e., a free-fall condition); a stopper 50, having a rectangular-plate shape, which covers a lower-edge portion of the fuel magazine 40 in the inclined direction so as to regulate the falling movement of the fuel rod 6; a cylinder (releasing mechanism) 55 which drives the stopper 50 in a direction perpendicularly crossing an extending direction of the fuel magazine 40 so as to release the regulating operation applied to the fuel rod 6; a rotating mechanism 60 which rotates the base 20; another rotating mechanism 70, arranged below the fuel magazine 40, which rotates the fuel magazine 40; a vertical lifting mechanism 80, arranged below the rotating mechanism 70, which vertically lifts up or down the fuel magazine 40 and rotating mechanism 70. On the upper surface of the foregoing base 20, at a position facing with the lower surface of the supporting post 30 which is located at the lowest position among the supporting posts, there is established a regulating plate 21 which avoids the falling movement of the fuel rod 6 inserted in the supporting grid 4. Below the lower surface of the upper edge portion of the foregoing base 20, there are provided an angle-type fitting member 22 which fits with a lower surface of a supporting plate 41 supporting the fuel magazine 40 so as to regulate the excessive rotation of this supporting plate 41; and a cylinder 23 which moves this fitting member 22 to be close to or apart from the supporting plate 41. The foregoing supporting posts 30 are provided to support each of the supporting grids 4 in a state where the opening portions of the grid cells 5 of each of the supporting grids 4 are arranged linearly along with a longitudinal direction of the base 20. Thus, under a state (as illustrated in FIG. 1) where the fuel magazine 40 is fixed on the supporting plate 41, the fuel rod 6 which is fallen down from the fuel magazine 40 can be inserted through the inside of the grid cell 5 of the supporting grid 4. The rotating mechanism 60 which rotates the base 20 is constructed by a rotation shaft 61 which is fixed at one edge side of the base 20 (i.e., left-edge side of FIG. 1); a supporting base 62 which supports this rotation shaft 61 in a free-rotation manner; and a hydraulic cylinder 63, arranged in the vicinity of the other edge side of the base 20, of which tip edge portion is connected by a pin with respect to the base 20 in a free-rotation manner. The tip edge portion of this hydraulic cylinder 63 and the base 20 are interconnected with each other such that they can be slightly and relatively moved in a horizontal direction, and consequently, the cylinder 63 will not prevent the rotation of the base 20. The rotating mechanism 70 which rotates the fuel magazine 40 is mounted horizontally on the vertical lifting mechanism 80, and it is constructed by a supporting plate 71 which is lifted up or down by the vertical lifting mechanism 80; a rotation shaft 72 which is fixed to one edge side of the supporting plate 41 supporting the fuel magazine 40; a supporting base 73, planted on the supporting plate 71, which supports the rotation shaft 72 in a free-rotation manner; and a hydraulic cylinder 74, arranged in the vicinity of the other edge side of the supporting plate 41, of which tip edge portion is fixed by a pin to the supporting plate 41 in a free-rotation manner. As similar to the foregoing hydraulic cylinder 63, the tip edge portion of this hydraulic cylinder 74 and the supporting plate 41 are interconnected with each other such that they can be slightly and relatively moved in a horizontal direction, and consequently, the cylinder 74 will not prevent the rotation of the supporting plate 41. The vertical lifting mechanism 80 is fixed to the supporting plate 71 at the upper edge portion thereof, and it is constructed by hydraulic cylinders 81, 82 which horizontally supports the supporting plate 71 such that this supporting plate can be freely lifted up or down; a base plate 83 which supports these hydraulic cylinders 81, 82; and a rail 84 which supports the base plate 83 such that this base plate can be moved in a direction (horizontal direction in FIG. 1) along with an extending direction of the fuel magazine 40. The above-mentioned base plate 83 and rail 84 construct a mobile mechanism which moves the fuel magazine 40 in a backward direction. Next, the operation of the assembling machine according to the present embodiment will be described. First, in an initial state, each of the hydraulic cylinders 63, 74, 81, 82 is contracted so that the base 20 and fuel magazine 40 are located in a horizontal state. At this time, in order to avoid the interference to be occurred between the right edge portion of the base 20 and the left edge portion of the fuel magazine 40, the base plate 83 is moved slightly in a right direction by the rail 84. In addition, the supporting grid 4 is located at each of the supporting posts 30 such that the grid cells 5 are directed in an extending direction (i.e., horizontal direction in FIG. 1) of the base 20. Further, the fuel magazine 40 is mounted on and then fixed to the upper portion of the supporting plate 41 such that the extending direction thereof matches with the extending direction of the base 20. Thereafter, the cylinder 55 is driven to fall down the stopper 50 so that this stopper 50 will cover the front edge surface of the fuel magazine 40, by which the fuel rod 6 is prevented from being pulled out from this front edge surface. Then, the hydraulic cylinder 63 is stretched, while the base 20 is rotated about the rotation shaft 61, so that the inclining movement of the base 20 is stopped at the predetermined inclined angle. Then, the base plate 83 is slightly moved close to the base 20 by the rail 84, while the hydraulic cylinders 81, 82 are stretched, so that the supporting plate 71 is lifted up. Such lift-up movement is stopped when the height of the supporting plate 41 is raised approximately identical to the height of the upper edge portion of the base 20. Next, the cylinder 23 attached to the base 20 is stretched so that the fitting member 22 is slightly projected in an upward direction of the base 20, by which the fitting member 22 is brought into contact with the lower surface of the left edge portion of the supporting plate 41. Then, the hydraulic cylinder 74 is stretched so that the supporting plate 41 is rotated about the rotation shaft 72, by which both of the supporting plate 41 and fuel magazine 40 are located approximately on the extending line of the base 20. In short, in this state, the fuel magazine 40 is located at the position which is the upper side of the inclined direction of the base 20 by the same inclined angle. At this time, the lower surface of the left edge portion of the supporting plate 41 comes in contact with the fitting member 22, by which the inclined angle of the supporting plate 41 can be adjusted. In other words, under operation of this fitting member 22, the inclined angle of the fuel magazine 40 can be set approximately identical to the inclined angle of the base 20. Next, as similar to the prior art, the spring 10 provided in the supporting grid 4 attached to each of the supporting posts 30 is supported in an escaping state, which is the state where the fuel rod 6 can be inserted through the inside of the grid cell 5 without any resistance. Then, the cylinder 55 is driven so as to gradually lift up the stopper 50. Thereafter, when the lower edge position of the stopper 50 is moved up to the position which is higher than the position of the fuel rod 6 stored in the lowest portion of the fuel magazine 40, the fall-down regulation applied to this fuel rod 6 is released, so that this fuel rod 6 is fallen down toward the supporting grid 4 which is supported by the supporting post 30. Thus, this fuel rod 6 can be inserted through the grid cell 5 of the supporting grid 4. The tip edge portion of the fuel rod 6 to be inserted through the supporting grid 4 comes in contact with the regulating plate 21, so that the fall-down movement is stopped. Next, as similar to the aforementioned operation, by gradually raising up the stopper 50, the fuel rods 6 stored in the inside of the fuel magazine 40 can be sequentially inserted through the supporting grid 4. According to the machine of the present embodiment, by raising up the stopper 50 as described above, the fuel rod 6 can be inserted through the supporting grid 4. Therefore, as compared to the prior-art machine which uses the pull-in rod and the like, it is possible to remarkably simplify the mechanical construction of the machine. Further, since the insert of the fuel rod 6 can be made by merely raising up the stopper 50, the holding operation of the fuel rod 6 by the pull-in rod and the like are not required, by which the inserting operation of the fuel rod 6 can be simplified. As described above, after inserting the fuel rod 6 through the supporting grid 4, as similar to the prior art, the pull-out operation of the key and the inserting operation of the control-rod guide pipe etc. are performed so that the fuel assembly can be produced. Next time when the assembling operation of the fuel assembly is performed, the hydraulic cylinders 63, 74, 81, 82 are contracted again so as to set them in an initial state. Thereafter, the other fuel magazine 40 is mounted on the upper portion of the supporting plate 41, and then the foregoing operations are performed. Incidentally, the shape of the plate-shape base 20 is merely limited to the shape by which the upper surface of the base can be inclined along with one direction, thus, it is not limited to the rectangular shape. In addition, it is obvious that the thickness of the base is not subjected to any restriction. The present invention is not restricted by the aforementioned embodiment, and consequently, it can contain several kinds of modified examples.
	claims	1. An image reading method for producing image data by irradiating an image carrier including two-dimensionally distributed spots of a labeling substance with a stimulating ray to excite the labeling substance and photoelectrically detecting light released from the labeling substance, the image reading method further comprising a stimulation and detection step of irradiating the image carrier with a line beam of the stimulating ray to excite the labeling substance and photoelectrically detecting light released from the labeling substance after the completion of irradiation with the stimulating ray. 2. An image reading method in accordance with claim 1 wherein the image carrier is intermittently moved relative to the line beam of the stimulating ray in a direction perpendicular to a longitudinal direction of the line beam and the stimulation and detection step is performed each time the image carrier is moved, thereby scanning the whole surface of the image carrier with the line beam of the stimulating ray and image data are produced by photoelectrically detecting light released from the labeling substance contained in the spots two-dimensionally distributed in the image carrier. claim 1 3. An image reading method in accordance with claim 1 , wherein the stimulation and detection step is repeated two or more times. claim 1 4. An image reading method in accordance with claim 1 , wherein the line beam of the stimulating ray is emitted from a laser diode array or a laser diode array constituted by one or more laser diodes. claim 1 5. An image reading method in accordance with claim 1 , wherein a laser beam emitted from a laser stimulating ray source is shaped using a lens to produce the line beam of the stimulating ray. claim 1 6. An image reading method in accordance with claim 1 , wherein the line beam of the stimulating ray is emitted from an LED array constituted by one or more LEDs. claim 1 7. An image reading method in accordance with claim 1 , wherein a stimulating ray emitted from an LED stimulating ray source is shaped using a lens to produce the line beam of the stimulating ray. claim 1 8. An image reading method in accordance with claim 1 , wherein a stimulating ray emitted from an LED stimulating ray source is shaped by a slit to produce the line beam of the stimulating ray. claim 1 9. An image reading method in accordance with claim 1 , wherein light released from the labeling substance is photoelectrically detected using a solid state imaging device. claim 1 10. An image reading method in accordance with claim 9 , wherein light released from the labeling substance is photoelectrically detected using a CCD line sensor. claim 9 11. An image reading method in accordance with claim 10 , wherein light released from the labeling substance is photoelectrically detected using a cooled CCD line sensor. claim 10 12. An image reading method in accordance with claim 9 , wherein light released from the labeling substance is photoelectrically detected using a photodiode array. claim 9 13. An image reading method in accordance with claim 12 , wherein light released from the labeling substance is photoelectrically detected using a cooled photodiode array. claim 12 14. An image reading method in accordance with claim 9 , wherein light released from the labeling substance is photoelectrically detected using a MOS type imaging device. claim 9 15. An image reading method in accordance with claim 14 , wherein light released from the labeling substance is photoelectrically detected using a cooled MOS type imaging device. claim 14 16. An image reading method in accordance with claim 1 , wherein the labeling substance is formed of a fluorescent substance. claim 1 17. An image reading method in accordance with claim 16 , wherein the image carrier is constituted as a membrane filter including the fluorescent substance contained in two-dimensionally distributed spots. claim 16 18. An image reading method in accordance with claim 16 , wherein the image carrier is constituted as a gel support including the fluorescent substance contained in two-dimensionally distributed spots. claim 16 19. An image reading method in accordance with claim 16 , wherein the image carrier is constituted as a micro-array including the fluorescent substance contained in two-dimensionally distributed spots. claim 16 20. An image reading method in accordance with claim 1 , wherein the image carrier is constituted as a stimulable phosphor sheet formed with a stimulable phosphor layer including a radioactive labeling substance contained in two-dimensionally distributed spots. claim 1 21. An image reading apparatus adapted for irradiating an image carrier including a labeling substance contained in two-dimensionally distributed spots with a stimulating ray and photoelectrically detecting light released from the labeling substance, thereby producing image data, the image reading apparatus comprising at least one stimulating ray source for emitting a stimulating ray, a stimulating ray shaping means for shaping the stimulating ray emitted from the at least one stimulating ray source into a line beam, a sensor for photoelectrically detecting light released from the labeling substance, and a control means for performing a stimulation and detection step of irradiating the image carrier including the labeling substance contained in the two-dimensionally distributed spots with the line beam of the stimulating ray to stimulate the labeling substance, stopping irradiation with the line beam of the stimulating ray and causing the sensor to photoelectrically detect light released from the labeling substance after the completion of irradiation with the line beam of the stimulating ray. 22. An image reading apparatus in accordance with claim 21 , which further comprises a scanning means for intermittently moving the image carrier relative to the line beam of the stimulating ray in a direction perpendicular to a longitudinal direction of the line beam and wherein the control means is constituted so as to perform the stimulation and detection step each time the image carrier is intermittently moved by the scanning means, thereby scanning a whole surface of the image carrier with the line beam of the stimulating ray and the sensor is constituted so as to photoelectrically detect light released from the labeling substance contained in the spots two-dimensionally distributed in the image carrier to produce image data. claim 21 23. An image reading apparatus in accordance with claim 21 , wherein the control means is constituted so as to repeat the stimulation and detection step two or more times. claim 21 24. An image reading apparatus in accordance with claim 21 , wherein the at least one stimulating ray source and the stimulating ray shaping means are constituted as a laser diode array or a laser diode array provided with two or more laser diodes. claim 21 25. An image reading apparatus in accordance with claim 21 , wherein the at least one stimulating ray source is constituted as a laser stimulating ray source and the stimulating ray shaping means is constituted as a lens. claim 21 26. An image reading apparatus in accordance with claim 21 , wherein the at least one stimulating ray source and the stimulating ray shaping means are constituted as an LED array provided with one or more LEDs. claim 21 27. An image reading apparatus in accordance with claim 21 , wherein the at least one stimulating ray source is constituted as an LED stimulating ray source and the stimulating ray shaping means is constituted as a lens. claim 21 28. An image reading apparatus in accordance with claim 21 , wherein the stimulating ray shaping means is constituted as a slit. claim 21 29. An image reading apparatus in accordance with claim 21 wherein the sensor is constituted as a solid state imaging device. claim 21 30. An image reading apparatus in accordance with claim 29 , wherein the sensor is constituted as a CCD line sensor. claim 29 31. An image reading apparatus in accordance with claim 29 , wherein the sensor is constituted as a cooled CCD line sensor. claim 29 32. An image reading apparatus in accordance with claim 29 , wherein the sensor is constituted as a photodiode array. claim 29 33. An image reading apparatus in accordance with claim 32 , wherein the sensor is constituted as a cooled photodiode array. claim 32 34. An image reading apparatus in accordance with claim 29 , wherein the sensor is constituted as a MOS type imaging device. claim 29 35. An image reading apparatus in accordance with claim 29 , wherein the sensor is constituted as a cooled MOS type imaging device. claim 29 36. An image reading apparatus in accordance with claim 21 , which further comprises a stimulating ray cut filter disposed in a path of light released from the labeling substance for cutting at least a light component having a wavelength of the stimulating ray. claim 21 37. An image reading apparatus in accordance with claim 21 , wherein the labeling substance is formed of a fluorescent substance. claim 21 38. An image reading apparatus in accordance with claim 37 , wherein the image carrier is constituted as a membrane filter including the fluorescent substance contained in two-dimensionally distributed spots. claim 37 39. An image reading apparatus in accordance with claim 37 , wherein the image carrier is constituted as a gel support including the fluorescent substance contained in two-dimensionally distributed spots. claim 37 40. An image reading apparatus in accordance with claim 37 , wherein the image carrier is constituted as a micro-array including the fluorescent substance contained in two-dimensionally distributed spots. claim 37 41. An image reading apparatus in accordance with claim 21 , wherein the image carrier is constituted as a stimulable phosphor sheet formed with a stimulable phosphor layer including a radioactive labeling substance contained in two-dimensionally distributed spots. claim 21
	claims	1. An apparatus for decreasing the birefringence of at least one starting halogenated optical material comprising:a chamber for providing an exposure atmosphere,a support member located within said chamber for supporting at least one starting halogenated optical material under tensile stress;a source having a cathode and an anode for providing a large area electron beam within the chamber, said large area electron beam irradiating said at least one starting halogenated optical material under tensile stress, andcontrol means to control the large area electron beam to decrease the birefringence of said at least one starting halogenated optical material under tensile stress. 2. The apparatus for decreasing the birefringence of at least one starting halogenated optical material of claim 1 further comprising:heating means to raise the temperature of said at least one starting halogenated optical material under tensile stress during said large area electron beam irradiating said at least one starting halogenated optical material under tensile stress. 3. The apparatus for decreasing the birefringence of at least one starting halogenated optical material of claim 2 wherein said heating means is located within said chamber. 4. The apparatus for decreasing the birefringence of at least one starting halogenated optical material of claim 2 wherein said heating means can raise said temperature of said at least one starting halogenated optical material under tensile stress to between 10 degrees Celsius and 1000 degrees Celsius during said large area electron beam irradiating said at least one starting halogenated optical material under tensile stress. 5. The apparatus for decreasing the birefringence of at least one starting halogenated optical material of claim 1 further comprising:an aperture mask for limiting said large area electron beam irradiating said at least one starting halogenated optical material under tensile stress to a selected area on said at least one starting optical material under tensile stress. 6. The apparatus for decreasing the birefringence of at least one starting halogenated optical material of claim 5 wherein said aperture mask forms a pattern to the birefringence in said starting halogenated optical material under tensile stress. 7. The apparatus for decreasing the birefringence of at least one starting halogenated optical material of claim 1 further comprising:an embossing structure for limiting said large area electron beam irradiating said at least one starting halogenated optical material under tensile stress to a selected area on said at least one starting optical material under tensile stress. 8. The apparatus for decreasing the birefringence of at least one starting halogenated optical material of claim 7 wherein said embossing structure forms a pattern to the birefringence in said starting halogenated optical material under tensile stress. 9. The apparatus for decreasing the birefringence of at least one starting halogenated optical material of claim 1 wherein the atmosphere in said chamber is between 1 milliTorr and 760 milliTorr. 10. The apparatus for decreasing the birefringence of at least one starting halogenated optical material of claim 1 wherein said at least one starting halogenated optical material is a mixture of at least two different halogenated optical materials. 11. The apparatus for decreasing the birefringence of at least one starting halogenated optical material of claim 1 further comprising:said large area electron beam irradiating a first sub-layer of said at least one starting halogenated optical material under tensile stress, andcontrol means to control the large area electron beam to decrease the birefringence of said first sub-layer of said at least one starting halogenated optical material under tensile stress with the remaining second sub-layer of said at least one starting halogenated optical material retaining its original birefringence. 12. The apparatus for decreasing the birefringence of at least one starting halogenated optical material of claim 11 wherein said large area electron beam partially penetrates said at least one starting halogenated optical material to form said first sub-layer. 13. A method for decreasing the birefringence of at least one starting halogenated optical material comprising:placing said at least one starting halogenated optical material under tensile stressirradiating said at least one starting halogenated optical material under tensile stress with a large area electron beam source, andcontrolling the energy of the electron beam source to decrease the birefringence of said at least one starting halogenated optical material under tensile stress. 14. The method for decreasing the birefringence of at least one starting halogenated optical material of claim 13 further comprising the step of:heating said at least one starting halogenated optical material under tensile stress during said irradiating said at least one starting halogenated optical material under tensile stress. 15. The method for decreasing the birefringence of at least one starting halogenated optical material of claim 14 wherein said heating can raise said temperature of said at least one starting halogenated optical material under tensile stress to between 10 degrees Celsius and 1000 degrees Celsius during said irradiating said at least one starting halogenated optical material under tensile stress. 16. The method for decreasing the birefringence of at least one starting halogenated optical material of claim 13 further comprising the step of:masking said at least one starting halogenated optical material under tensile stress to limit said irradiating said at least one starting halogenated optical material under tensile stress to a selected area on said at least one starting halogenated optical material under tensile stress. 17. The method for decreasing the birefringence of at least one starting halogenated optical material of claim 16 wherein said masking is done by an aperture mask. 18. The method for decreasing the birefringence of at least one starting halogenated optical material of claim 17 wherein said masking is done by an embossing structure. 19. The method for decreasing the birefringence of at least one starting halogenated optical material of claim 13 wherein the atmosphere during said irradiating said at least one starting halogenated optical material is between 1 milliTorr and 760 milliTorr. 20. The method for decreasing the birefringence of at least one starting halogenated optical material of claim 13 further comprising the step of:forming in said at least one starting halogenated optical material a decreased birefringence by creating additional bond structure in said at least one starting halogenated optical material under tensile stress. 21. The method for decreasing the birefringence of at least one starting halogenated optical material of claim 13 wherein said at least one starting halogenated optical material is a mixture of at least two different halogenated optical materials. 22. The method for decreasing the birefringence of at least one starting halogenated optical material of claim 13 whereinirradiating a first sub-layer of said at least one starting halogenated optical material under tensile stress with a large area electron beam source, andcontrolling the energy of the electron beam source to decrease the birefringence of said first sub-layer of said at least one starting halogenated optical material under tensile stress with the remaining second sub-layer of said at least one starting halogenated optical material retaining its original birefringence. 23. The method for decreasing the birefringence of at least one starting halogenated optical material of claim 22 wherein said large area electron beam partially penetrates said at least one starting halogenated optical material to form said first sub-layer.
	description	A remotely installable jet pump diffuser weld repair device structurally replaces cracked adapter or lower ring to tail pipe welds and the tail pipe to shell welds. The device is remotely installable in the limited space in the shroud to vessel annulus of boiling water reactor (BWR) power plants. The jet pump assembly is part of the reactor recirculation system. Each assembly includes a riser assembly, a riser brace, two inlet-mixer assemblies, and two diffuser assemblies. Each assembly is installed in the annulus between the reactor pressure vessel (RPV) and the shroud. There are twenty jet pumps (ten jet pump assemblies) installed in a typical General Electric BWR. The riser assembly is welded to the reactor pressure vessel (RPV) at the riser brace location, and the recirculation inlet nozzle at the penetration. The two diffuser assemblies (DA) are welded to the shroud support plate (see FIG. 1). The two inlet-mixer assemblies are removable components. The entrance end of each inlet-mixer assembly seats into the top of the riser transition piece. The exit end fits into a slip joint with the top of the diffuser assembly. Lateral support for the inlet-mixer is provided by the restrainer bracket, which is welded to the riser pipe. Two adjusting screws (set screws), each threaded into the restrainer bracket, and the inlet-mixer wedge provide three points of lateral support for the inlet-mixer. Lateral support for the riser assembly is provided by the riser brace. Existing jet pump assembly components are fabricated from Type 304 stainless steel, with the exception of the diffuser lower ring, which is fabricated from Ni—Cr—Fe Alloy 600. The welds are designated DF-3 or DF-2 as shown in FIG. 1. Cracking attributed to Intergranular Stress Corrosion Cracking (IGSCC) has been observed at the Alloy 600 to stainless steel transition weld (DF-3) between the lower ring and tail pipe. The postulated crack in a DF-3 or a DF-2 weld would detach the diffuser from the jet pump assembly, resulting in the total loss of the jet pump assembly. This could in turn result in failure of jet pump functionality in providing recirculation flow path to reactor core and loss of providing the 2/3-core height coverage in the event of LOCA accident. The failure of a jet pump assembly occurring during operation will cause a change in the monitored jet pump flow, which would be detected, allowing the plant to be brought to a safe shut down condition in accordance with plant technical specifications. Jet pump diffuser weld cracks that have required repair have been a relatively recent occurrence. There has only been one other type of jet pump diffuser repair developed using a tie rod assembly. The scope of that repair was limited to hardware design and analysis and installation tooling concept design. Full scale mockup testing was not done. Repair clamps for welds in other tubular portions of the jet pump have been developed specifically for the thermal sleeve to elbow weld in the riser portion of the jet pump. U.S. Pat. Nos. 6,053,652, 6,108,391 and 6,086,120 relate to jet pump riser thermal sleeve to elbow weld repair. In an exemplary embodiment of the invention, a jet pump diffuser weld repair device includes a lower ring section and an upper ring section respectively sized to fit around a circumference of the diffuser on opposite sides of the weld to be repaired. The lower and upper ring sections are provided with a plurality of aligned gripper slots. A corresponding plurality of grippers are fit into the gripper slots, where at least one of the gripper slots and the grippers defines cam surfaces shaped to drive the grippers radially inward as the lower and upper ring sections are drawn toward each other. A plurality of connector bolts are secured between the lower ring section and the upper ring section. Tightening the connector bolts draws the lower and upper ring sections toward each other. In another exemplary embodiment of the invention, a weld repair device for a thin wall welded pipe includes a lower ring section and an upper ring section respectively sized to fit around a circumference of the pipe on opposite sides of the weld to be repaired. A plurality of grippers are fit into aligned gripper slots in the lower and upper ring sections. The grippers include a double-tapered outer surface defining cam surfaces shaped to drive the grippers radially inward as the lower and upper ring sections are drawn toward each other by a plurality of connector bolts secured between the lower ring section and the upper ring section. In yet another exemplary embodiment of the invention, a method of repairing a weld connection in a jet pump diffuser includes the steps of forming pockets in an exterior surface of the diffuser on opposite sides of the weld to be repaired; fitting a plurality of grippers into a corresponding plurality of aligned gripper slots in the lower ring section and the upper ring section, wherein at least one of the gripper slots and the grippers defines cam surfaces shaped to drive the grippers radially inward as the lower and upper ring sections are drawn toward each other; placing the lower ring section and the upper ring section around a circumference of the diffuser on opposite sides of the weld, respectively, with lugs on the grippers engaging the pockets; and tightening connector bolts secured between the lower ring section and the upper ring section to draw the lower and upper ring sections toward each other. The jet pump diffuser repair described herein structurally replaces any one of the DF-3 or DF-2 welds. This will provide for the required vertical and lateral support of the diffuser assembly, even if complete failure of one of these welds occurs. The design of the repair assumes that other welds in the diffuser assembly remain intact (welds connecting collar/shell/tail pipe) and other components of the jet pump assembly as well. Thus, it is preferably preferable that the installation of only one jet pump diffuser repair is permitted on any jet pump pair. The estimated design life of the diffuser repair clamp is 40 years. With reference to FIG. 2, a lower ring section includes a clamp half lower female 3 and a clamp half lower male 4 bolted together using two connector bolts 6. The upper ring section includes a clamp half upper female 1 and a clamp half upper male 2 bolted together using two connector bolts 6. The female clamp half connections contain T-slots 14 or other suitable structure into which the corresponding T's 16 or the like on the male clamp halves are inserted (see FIG. 7). This connection takes the hoop loads in the assembly clamp rings so that the connector bolts 6 are only loaded axially, which allows the size of the connector bolts 6 to be minimized since they only experience the axial load due to connector bolt torquing. No shear loads are transmitted through the connector bolts 6. This feature also aligns the bolt holes on the male clamp halves with threaded holes in the female clamp halves to help facilitate the remote installation of the connector bolts. A plurality of grippers 5 fit into aligned slots 10 in the clamp halves 1-4. The grippers 5 have a 10° taper 5a, 5b on the surface that contacts the clamp halves 1-4. The clamp half slots 10 have a matching 10° angle on the gripper contact surfaces. The upper and lower clamp rings 1-4 are joined using guide bolts 7 and nuts 8. The guide bolts 7 and nuts 8 have spherical surfaces where they contact the clamp halves 1-4, which facilitate assembly when the upper and lower clamp rings are not perfectly parallel to each other. Torquing the guide bolts 7 draws the upper and lower clamp rings 1-4 toward each other, which forces the grippers 5 radially inward. This presses engaging lugs 5c of the grippers 5 into pockets 12 machined into the jet pump diffuser (see FIGS. 5 and 6). The grippers 5 bear against the jet pump diffuser pockets 12 providing a positive connection across the weld being repaired. FIG. 3 illustrates the repair device installed on the DF-3 weld. FIG. 4 illustrates the repair device installed on the DF-2 weld. The shallow 100 tapers between the grippers 5 and the clamp half slots 10 and the grippers 5 and the jet pump diffuser pockets 12 provide a mechanical advantage in the clamp tightening mode and prevent the applied loads across the welds from being transmitted radially outward into the clamp halves 1-4 during the plant operating transients where the applied loads exceed the clamping load. This is important because the applied loads on the jet pump diffuser during a Loss-of-Coolant-Accident are quite large, and if the repair clamp rings had to be able to accommodate a radial component resulting from those loads, the size of the clamp rings would have to be significantly increased. This could potentially make the repair impossible to install due to the access and space limitations in the annulus surrounding the jet pump diffusers. With continued reference to FIG. 5, the DF-3 weld repair spans the transition from Alloy 600 to Type 304 material. The lower ring below the weld is Alloy 600 and tail pipe above the weld is Type 304. Type XM-19 material is used for the gripper for strength reasons and to address the different thermal expansion rates of Alloy 600 and Type 304 material. Since the value of the coefficient of thermal expansion for Type XM-19 lies between the values for Alloy 600 and Type 304, it is possible to select pocket locations above and below the weld that result in no relative differential expansion between the clamp and the jet pump. This prevents both excessive loosening and tightening. As long as the length of Alloy 600 material is 0.388 times the length of the Type 304 material (which was derived from a differential expansion evaluation performed for the design), there is no net differential expansion between the repair and the jet pump diffuser. As shown in FIG. 6, the DF-2 weld repair spans the transition from the tail pipe to the shell. While there is no material change at this location, there is a change in wall thickness and angle. The jet pump contour changes from a vertical cylinder to a slightly angled conical shape. The wall thickness changes from around 0.38″ to 0.25″. The 0.25″ thickness is too thin to support a single pocket of the necessary depth. Thus, the diffuser pockets 12 are machined as multiple shallow grooves for the DF-2 weld. The repair location is near the bottom of the shroud to vessel annulus. The available space in this region is very limited. Also, the jet pump sensing lines and jet pump riser restrict access to this region. The space and access restrictions at the location of the repair require that the repair be installed in pieces and assembled in place on the jet pump diffuser. When the guide bolts 7 are tightened, the upper and lower clamp rings 1-4 are drawn toward each other, and the tapered slots 10 on the clamp rings push the double-tapered outside surface of the grippers 5 inward against the jet pump diffuser. The grippers 5 engage pockets 12 machined in the tail pipe and the lower ring preferably by remote Electrical Discharged Machining (EDM). This engagement provides a positive connection that controls the separation of the diffuser at the crack location. All bolting is retained by ratchet type locking springs that prevent counterclockwise rotation. Future removal of the repair is possible by reversing the installation sequence. The locking springs will be unlocked remotely using a tool that displaces the ratchet spring and disengages the ratchet teeth from the interfacing teeth on the clamp bolt. Also, all bolting is oriented vertically in order to improve the installation access. If the alloy 600 diffuser surfaces that are machined by the EDM process are not polished, the repair life may be limited, such as limited to two fuel cycles. The installation of the jet pump diffuser repair will not affect the operation or performance of the jet pump assembly. One advantage of this repair over the tie rod style repair is that it does not require machining holes through the core support plate. This is beneficial since any holes in the shroud support plate would have to be plugged after the tie rod repair is removed. These plugs would require periodic inspection throughout the remaining life of the plant, which would add expense and outage time. Detailed structural analysis has been performed on this repair design. Detailed finite element analysis modes have been created and used to determine the stresses for all applicable plant operating conditions. The repair has been shown to meet all of the applicable stress limit criteria. The affect of the repair on the stiffness of the jet pump has been evaluated. It has been determined that in some instances, the use of a jet pump slip joint clamp is required to meet jet pump slip joint flow-induced vibration criteria. While the invention has been described in connection with what is presently considered to be the most practical and preferred embodiments, it is to be understood that the invention is not to be limited to the disclosed embodiments, but on the contrary, is intended to cover various modifications and equivalent arrangements included within the spirit and scope of the appended claims.
	summary
054935997	abstract	A toroidal x-ray tube (I) is supported and selectively positioned by a gantry (II). The x-ray tube includes a toroidal housing (A) in which a rotor (20) is rotatably mounted. One or more cathodes (C) are mounted on the rotor for generating an electron beam which strikes an anode (B) to generate a beam of x-rays which pass through a patient aperture (62) to strike a detector (60). The x-ray tube includes pre-collimators (70, 74) having slots (72, 76) for passage of the x-ray subsequent to generation thereof and prior to being collimated by the collimator (90). A ring collimator (90) collimates an x-rays formed into a fan shaped beam. The collimator (90) includes a first ring (92) and a second ring (94) which are concentric. The distance between the first and second rings (92, 94) is adjustable to adjust the slice thickness of the final image. The x-ray tube provides improved final images in that reduction of off-focal radiation occurs due to the utilization of pre-collimators and the collimation of x-rays is flexible due to adjustability of slice thickness.
	abstract	A fuel exchange apparatus, comprising: a traveling carriage moving horizontally in one direction; a traversing carriage moving horizontally on the traveling carriage in a direction orthogonal to the one direction in which the traveling carriage moves; and a fuel holding unit attached to the traversing carriage, and including an telescopic tube enabling to extend and contract, a holding tool for holding a fuel assembly and releasing the held fuel assembly, and a lifter for raising and lowering the holding tool by winding and running out linear members for suspending and supporting the holding tool from the traversing carriage, wherein the holding tool is selectively placed in a constrained state in which the holding tool is subject to a constraint by the telescopic tube and in a freely suspended state in which the holding tool is released from the constraint by the telescopic tube and freely suspended by the linear members.
	description	The present application claims priority from Japanese application JP 2005-004660 filed on Jan. 12, 2005, the content of which is hereby incorporated by reference into this application. The present invention relates to a scanning transmission electron microscope (: STEM). More concretely, it relates to a control device and method for controlling an electron beam which has passed through a specimen. The STEM is a device for visualizing specimen structure with a sub-nanometer high space resolution. A raster scanning of an electron beam which is converged down to a nanometer order is performed on a specimen whose film is thinned down to a few hundreds of nm. Moreover, a signal generated from an electron-beam irradiation area is detected, then being synchronized with the raster scanning. This synchronization allows the two-dimensional image to be formed. Examples of the signal generated from the electron-beam irradiation area are a transmitted electron beam, a secondary electron beam, and characteristic X-rays. FIG. 2 is a basic configuration diagram of the general STEM. The direction parallel to an optical axis 20 of the lens-barvel is defined as the Z direction, and the plane perpendicular to the optical axis is defined as the XY plane. A primary electron beam emitted from an electron gun 1 is accelerated up to a few hundreds of kV, then being formed in shape by a first condenser lens 2 and a second condenser lens 3. Moreover, the primary electron beam passes through a condenser aperture 4 for restricting an aperture angle of the primary electron beam, then being focused on a specimen 21 by an upper objective lens 7-1. Although an objective lens is, physically, a single lens, the specimen 21 is set up in a gap of the objective lens. For this reason, the upper objective lens 7-1 and a lower objective lens 7-2 are assumed as the ray diagram. Here, the lens 7-1 allows focal point of the primary electron beam to be achieved on the specimen, and the lens 7-2 has a role of projecting the transmitted electron beam which has passed through the specimen 21. An irradiation lens system including these lenses permits the primary electron beam to be converged down into a sub-nm diameter on the specimen 21. The transmitted electron beam which has passed through the specimen 21 is projected onto an electron-beam detection system by the lower objective lens 7-2 and a projection lens 9. A raster scanning of the electron beam is performed within the XY plane by a scanning coil 5. Furthermore, a control signal for the scanning coil 5 and an output signal from an electron detector 14 are synchronized with each other, thereby forming a STEM image within a computer, and displaying the STEM image on a monitor. The characteristic of the STEM is that changing the detection signal permits various specimen information to be imaged in an easy and convenient manner. For example, if high-angle scattered electrons are detected using an electron annular detector 12, a high-angle annular dark field (: HAADF) image can be acquired. If low-angle scattered electrons in proximity to the optical axis 20 are selected using an angle selection aperture 13, and if the low-angle scattered electrons selected are detected using the electron detector 14, a bright field (: BF) image can be acquired. JP-A-2001-93459 has disclosed the following technology: Namely, according to this technology, in the device configuration for acquiring an electron energy loss spectrum (: EELS) image, a change in incident position of the transmitted electron beam relative to the electron detector is cancelled using a de-scanning coil. Here, this change will occur in accompaniment with a change in incident position of the primary electron beam relative to the specimen. If the incident position of the primary electron beam relative to the specimen 21 is changed, the incident position of the transmitted electron beam relative to the electron detector changes. Accordingly, when selecting electrons under a certain condition out of the transmitted electrons by an aperture or a slit, and detecting the selected electrons, the transmitted electron beam displaces relative to the electron detector. On account of this displacement, the condition of the transmitted electrons to be detected at the electron detector changes depending on the incident position of the primary electron beam. This condition change causes an artifact to occur in the STEM image. The use of the de-scanning coil makes it possible to suppress this artifact. In JP-A-2001-93459, however, no disclosure has been made concerning a concrete control method for the de-scanning coil. It is an object of the present invention to provide an electron microscope or its application device where a higher image correction is implemented. This implementation is accomplished by performing the de-scanning of the transmitted electron beam with an accuracy which is higher than the one in the conventional technology. In order to implement the high-accuracy image correction, the present invention actually measures the displacement quantity of the transmitted electron beam, then controlling the de-scanning of the transmitted electron beam on the basis of an actually-measured result. For this purpose, in the electron microscope or an image correction system according to the present invention, there are provided an actually-measuring unit for actually measuring a displacement quantity of the transmitted electron beam, a de-scanning unit for de-scanning the transmitted electron beam, and a determination unit for determining a de-scanning quantity of the transmitted electron beam on the basis of the actually-measured displacement quantity. moreover, the electron microscope or the image correction system feedbacks the determined de-scanning quantity to the de-scanning unit, thereby executing the image correction. As the actually-measuring unit for actually measuring the displacement quantity, the electron detector, e.g., a scintillator-equipped CCD camera, is usable. As the de-scanning unit for de-scanning the transmitted electron beam, the use of, e.g., the de-scanning coil, is preferable. If, however, some other appropriate unit exists, using that unit is also preferable. The determination unit for determining the de-scanning quantity is implemented by, e.g., a method where a computation unit for determining the de-scanning quantity on the basis of the actually-measured value is caused to execute a predetermined algorithm. In a system where the de-scanning coil makes it possible to cancel the displacement of the transmitted electrons caused by the primary-electron-beam scanning, the most important device performance is the correction accuracy. In the conventional control system, however, implementation of an enhancement in the correction accuracy has required a complicated and troublesome device adjustment. Accordingly, we have devised the following system by digitizing the control over the scanning coil 5: Namely, while being synchronized with a digital control signal resulting from this digitization, values in a de-scanning table registered in a FM (2) are outputted to the de-scanning coil 11. Here, the de-scanning table is created as follows: Positions of the transmitted electrons before and after activating the scanning coil 5 and the de-scanning coil 11 are photographed using a camera. Then, based on a result acquired by analyzing a resultant displacement quantity of the transmitted electrons, the de-scanning table is created. Moreover, even if a set state of the device has been changed, it is possible to deal with this change by updating the de-scanning table. This makes it possible to execute a high-accuracy de-scanning at any time. Furthermore, updating the de-scanning table is automatically executed in accordance with an analysis flowchart registered in a computer 16. This makes the complicated and troublesome adjustment task unnecessary. If an end user has recognized that the transmitted electrons are being oscillated by the incident-electron-beam scanning, all that the end user has to do is to send an instruction of updating the de-scanning table. The enhancement in the correction accuracy makes it possible to enlarge the field-of-view of the STEM image. This tremendously improves the usability of the device. Also, it becomes possible to make smaller the hole diameter of the angle selection aperture 13. This also allows an enhancement in the selection accuracy of the transmitted electron beam. Other objects, features and advantages of the invention will become apparent from the following description of the embodiments of the invention taken in conjunction with the accompanying drawings. In the electron microscope or the image correction system according to the present invention, there are provided an actually-measuring unit for actually measuring a displacement quantity of the transmitted electron beam, a de-scanning unit for de-scanning the transmitted electron beam, and a determination unit for determining a de-scanning quantity of the transmitted electron beam on the basis of the actually-measured displacement quantity. Moreover, the electron microscope or the image correction system feedbacks the determined de-scanning quantity to the de-scanning unit, thereby executing the image correction. As the actually-measuring unit for actually measuring the displacement quantity, the electron detector, e.g., a scintillator-equipped CCD camera, is usable. If the transmitted electrons enter the scintillator, photons are generated from the entrance positions, then being detected by the CCD camera. The electron image photographed by the CCD camera is sent to an image processing unit. The image processing unit analyzes the displacement quantity of the transmitted electron beam caused by a displacement of the primary electron beam, then recording the analysis result into a memory. As the de-scanning unit for de-scanning the transmitted electron beam, the use of the de-scanning coil is preferable. If, however, some other appropriate unit exists, using that unit is also preferable. For example, in the STEM of a configuration as illustrated in FIG. 3, the transmitted electrons to be detected by the electron detector 14 may be corrected by causing the position of the angle selection aperture 13 to follow the displacement quantity of the transmitted electron beam. Also, the transmitted electrons which are to form the STEM image may be corrected as follows: The electron detector 14 is changed to an electron-image detector, thereby being formed into a system for forming the STEM image by a detection signal in a specified pixel area. Then, the specified pixel area is caused to follow the displacement quantity of the transmitted electron beam, thereby correcting the transmitted electrons. The determination unit for determining the de-scanning quantity is implemented by, e.g., the following method: Namely, measurements are made concerning the correspondence between the control quantity of the scanning coil of the primary electron beam and the displacement quantity of the transmitted electron beam, and the correspondence between the control quantity of the de-scanning unit and the displacement quantity of the transmitted electron beam. Then, from these actually-measured values, a computation unit calculates the control quantity of the de-scanning unit necessary for canceling the displacement of the transmitted electron beam caused by the displacement of the primary electron beam. The control quantity of the de-scanning unit is stored into a record unit such as the memory together with an electron-optics condition at the time of the measurements. At the time of executing the de-scanning, the de-scanning control quantity which is necessary in correspondence with an electron-optics condition at the time of the execution is read out, then being sent to the de-scanning unit. Incidentally, if the de-scanning control quantity under the electron-optics condition at the time of the execution has been not recorded in the memory, a de-scanning control quantity already recorded in the memory is sent to the de-scanning unit after a necessary computation has been applied thereto. For example, if the magnification differs between the measurement time and the execution time, a value acquired by making the magnification correction to the de-scanning control quantity already recorded in the memory is sent to the de-scanning unit. Hereinafter, the explanation will be given below concerning a first embodiment. FIG. 3 illustrates the basic configuration diagram of the STEM used in the present embodiment. The direction parallel to an optical axis 20 of the housing is defined as the Z direction, and the plane perpendicular to the optical axis is defined as the XY plane. The present STEM includes the following configuration components: An electron gun 1 for emitting a primary electron beam, a first condenser lens 2 and a second condenser lens 3 for forming in shape the primary electron beam emitted from the electron gun 1, a condenser aperture 4 for restricting an aperture angle of the primary electron beam, a scanning coil 5 for allowing the primary electron beam to be scanned on a specimen, an objective lens 7 for allowing focal point of the primary electron beam to be achieved on the specimen, a specimen stage 8 for holding the specimen 21, a secondary-electron detector 6 for detecting secondary electrons emitted from surface of the specimen 21, a projection lens 9 for projecting a transmitted electron beam, which has passed through the specimen 21, onto an electron-beam detection system, a detected-electron alignment coil 10 for aligning the transmitted electrons onto the electron-beam detection system, a de-scanning coil 11 for canceling a displacement of the transmitted electron beam on an angle selection aperture 13 caused by an incident-electron-beam scanning on the specimen 21, an electron annular detector 12 for detecting high-angle scattered electrons out of the transmitted electrons, the angle selection aperture 13 for extracting from the transmitted electrons an electron beam having a desired detection angle, an electron detector 14 for detecting the transmitted electrons which have passed through the angle selection aperture 13, an electron detection camera 15 for photographing a diffraction image formed by the transmitted electrons, control circuits for controlling the electron gun, the respective electron lenses, the respective coils, the respective apertures, the specimen stage, and the respective detectors, and a computer 16 for controlling the respective control circuits. The angle selection aperture 13 and the electron detector 14 are formed into a movable-type mechanism which will deviate from the optical axis when the diffraction image is observed by the electron detection camera 15. Next, the explanation will be given below concerning processing steps for acquiring the STEM image by using the device illustrated in FIG. 3. The primary electron beam is extracted from the electron gun 1 by an extraction voltage V1. Then, an acceleration voltage V0 is applied to the primary electron beam extracted. The specimen 21 is mounted on the specimen stage 8, then being inserted into a specimen chamber. Current values for the first condenser lens 2, the second condenser lens 3, the objective lens 7, and the projection lens 9 are set at predetermined values. Although the objective lens 7 is, physically, a single lens, the specimen is set up in a gap of the objective lens. For this reason, an upper objective lens 7-1 and a lower objective lens 7-2 are assumed as the ray diagram. Here, the upper objective lens 7-1 allows focal point of the primary electron beam to be achieved on the specimen, and the lower objective lens 7-2 has a role of projecting the transmitted electron beam which has passed through the specimen 21. The primary electron beam is formed in shape by the first condenser lens 2 and the second condenser lens 3, then being focused on surface of the specimen 21 by the upper objective lens 7-1. The transmitted electron beam which has passed through the specimen 21, after being image-formed by the lower objective lens 7-2 and the projection lens 9 one after another, is projected onto the electron-beam detection system. Furthermore, the scanning coil 5 is controlled, thereby performing a raster scanning of the microscopically-converged electron beam within the XY plane on the surface of the specimen 21. Then, scanning position information by the scanning coil 5 and an output signal from the electron detector 14 are synchronized with each other, thereby forming a STEM image within the computer, and displaying the STEM image on a monitor. The characteristic of the STEM is that changing the electrons to be detected permits various specimen information to be imaged in an easy and convenient manner. For example, if the high-angle scattered electrons are detected using the electron annular detector 12, a high-angle annular dark field (: HAADF) image can be acquired. If low-angle scattered electrons in proximity to the optical axis 20 are selected using the angle selection aperture 13, and if the low-angle scattered electrons selected are detected using the electron detector 14, a bright field (: BF) image can be acquired. If a diffracted electron beam with a specific diffraction index is guided from the transmitted electron beam into the hole of the angle selection aperture 13 by using the detected-electron alignment coil 10, and if the diffracted electron beam guided is detected using the electron detector 14, a dark field image with the specified diffraction index can be acquired. Also, if the secondary electrons emitted from the surface of the specimen 21 are detected using the secondary-electron detector 6, a highly-accelerated secondary-electron image can also be acquired. The secondary-electron detector 6 is configured to take in the secondary electrons which have been emitted in a variety of directions. Consequently, even if incident position of the primary electron beam on the specimen 21 is changed, detection efficiency of the secondary-electron beam seldom changes. On the other hand, the transmitted-electron detector 14 detects the transmitted electrons in such a manner that, depending on a diffraction direction of the transmitted electron beam, the transmitted-electron detector 14 selects the transmitted electrons. For this selection to be implemented, there is provided a projection lens system including the units such as the projection lens 9, the detected-electron alignment coil 10, and the angle selection aperture 13. In this case, if the incident position of the primary electron beam on the specimen 21 is changed, orbits of the transmitted electrons change in the projection lens system. As a result, the incident position of the transmitted electron beam on the angle selection aperture 13 changes. FIG. 4 illustrates a change in the incident position of the transmitted electron beam on the angle selection aperture 13 in the case where the incident position of the primary electron beam on the specimen 21 is changed. If the incident position of the primary electron beam on the specimen 21 is displaced by a displacement quantity Rs from on the optical axis, the incident position of the transmitted electron beam entering the angle selection aperture 13 displaces by a displacement quantity Rd. It is possible to calculate a substantial value of the displacement quantity Rd from optical magnification of the lower objective lens 7-2 and that of the projection lens 9. Concretely, by calculating the optical magnification Mo of the lower objective lens 7-2, the optical magnification Mp of the projection lens 9, and a beam spread r on the angle selection aperture 13, the displacement quantity Rd can be approximately calculated based on the following expression 1: R d ≅ R s · M o · M p + r = R s · L o L s · L p L 3 - Lo + α · ( L 3 - L o ) · L 4 - L p L p ( 1 ) However, the settings of the objective lens 7 and the projection lens 9 change because of a variety of factors as well. For example, FIG. 5 illustrates changes in Rd which occur when z position of the specimen 21 within the objective lens is changed. For simplicity, the consideration will be given regarding the case where the projection lens is omitted. FIG. 5(a) illustrates the case where the specimen 21 is positioned at the center of the objective lens. If the specimen 21 is positioned in the downstream than the center of the objective lens, in order to achieve the focal point of the primary electron beam onto the specimen 21, the magnetizing excitation of the objective lens 7 is lowered thereby to lengthen the focal length thereof (FIG. 5(b)). Meanwhile, if the specimen 21 is positioned in the upstream than the center of the objective lens, the focal length of the objective lens 7 is shortened (FIG. 5(c)). As described above, if the height of the specimen 21 is changed, position of the image plane of the objective lens, i.e., Lo, changes. Moreover, it is obvious that position of the image plane of the projection lens, i.e., Lp, also changes. The expression 1 shows that, if Lo or Lp is changed, Rd will change. As factors for changing Rd, in addition to the z-position change in the specimen, z-position change in a virtual light-source can also be mentioned. Furthermore, there are some cases where, depending on an adjustment state of the electron-optics system, Rd will change asymmetrically with reference to the optical axis. For example, if an astigmatism exists in the projection lens, the change quantity of Rd also turns out to have an astigmatism. Also, if Rd becomes larger, it becomes impossible to neglect influence of an off-axis aberration of the projection lens. Implementation of the high-accuracy de-scanning requires a system which makes it possible to cancel these factors in an easy and convenient manner. Accordingly, in order to implement the high-accuracy de-scanning, we have devised the following system: Namely, this system actually measures Rd, then creating a de-scanning table on the basis of the actually-measured result. Moreover, based on this de-scanning table, this system controls the de-scanning coil 11. FIG. 1 illustrates a system configuration diagram thereof. In the present system, the control value for the scanning coil 5 is outputted from a DBC (: Digital Beam Control) circuit. While being synchronized with this digital control signal, values in the de-scanning table registered in a FM (: Frame Memory) (2) are outputted to the de-scanning coil 11. In addition to a system for writing the output from the electron detector 14 into a FM (: Frame Memory) (1) in synchronization with the scanning coil 5, the correction system in the present embodiment also includes a system for outputting the values recorded in the FM (2) in the manner of being synchronized with the scanning coil 5. The de-scanning table is caused to have the same pixel number as that of the STEM image. It is more efficient to make the pixel number of the de-scanning table and that of the STEM image equal to each other. This is because no calculation is needed when reading the control values from the de-scanning table. If, however, an algorithm is added which is used for performing a reading-through or overlapping of the control values, the pixel number of the de-scanning table and that of the STEM image may differ from each other. At the time of the factory shipping, it is preferable that the de-scanning table measured by electron-optics simulation values or the same kind of device be stored into the frame memory. By using this de-scanning table, a beginner who has not mastered a proofreading method for the de-scanning table executes the de-scanning. It is advisable that, after mastering the proofreading method, the beginner create the de-scanning table by using the most up-to-date measurement values thereby to enhance the de-scanning accuracy. Incidentally, if a device installer performs the proofreading of the de-scanning table at the time of the installment, it is allowable that the frame memory at the time of the factory shipping be left blank. The de-scanning table is created as follows: Diffraction images before and after the activations of the scanning coil and the de-scanning coil are photographed using a camera. Then, based on a result acquired by analyzing a resultant displacement quantity of the diffraction images by the image processing, the de-scanning table is created. Various pattern matching methods, such as phase-limited correlation method, normalization correlation method, and least-squares method, are usable as the analysis method for analyzing the displacement quantity between the diffraction images. Incidentally, the phase-limited correlation method has been used for the displacement-quantity analysis at this time. Hereinafter, referring to FIG. 18, the explanation will be given below concerning this phase-limited correlation method. Assume two pieces of discrete images S1(m, n) and S2(m, n) with a position shift D=(Dx, Dy) existing therebetween, and describe S1(m, n) as S1(m, n)=S2(m+Dx, n+Dy). Let two-dimensional discrete Fourier transformations of S1(m, n) and S2(m, n) be S1′(k, l) and S2′(k, l), respectively. Since a formula F{S(m+Dx, n+Dy)}=F{S(m, n)} exp (iDx·k+iDy·l) exists in the Fourier transformation, S1′(k, l) can be converted into S1′(k, l)=S2′(k, l) exp (iDx·k+iDy·l). Namely, the position shift between S1′(k, l) and S2′(k, l) can be expressed by the phase difference exp (iDx·k+iDy·l)=P′(k, l). P′(k, l) is also a wave whose period is equal to (Dx, Dy) As a result, a δ-function-like peak occurs at the position of (Dx, Dy) in an analysis image P(m, n) which results from applying an inverse Fourier transformation to the phase-difference image P′(k, l). Incidentally, instead of eliminating all the information on amplitude, a log or √{square root over ( )} processing is applied to the amplitude component of S1′(k, l)·S2′(k, l)*=\|S1′\| \|S2′\| exp(iDx·k+iDy·l), thereby calculating an image whose amplitude component is suppressed. Then, applying the inverse Fourier transformation to this image also causes a δ-function-like peak to occur at the position of (Dx, Dy) of the position-shift vector. Accordingly, the position-shift analysis may also be performed using this image. Moreover, applying the Fourier transformation to the phase-difference image P′(k, l) also causes a δ-function-like peak to occur at (−Dx, −Dy) Consequently, the position-shift analysis may also be executed using the Fourier transformation image of the phase-difference image P′(k, l). It can be assumed that only the δ-function-like peak exists in the analysis image P(m, n). This condition allows the position of the δ-function-like peak to be determined with an accuracy of digits to the right of the decimal point by using a center-of-gravity position calculation or function fitting. Also, the portion other than the δ-function-like peak can be regarded as a noise. Accordingly, the proportion of intensity of the δ-function-like peak relative to intensity of the entire analysis image P(m, n) can be regarded as the coincidence degree between the images. In the conventional position-shift analysis methods, it was difficult to evaluate reliability of the position-shift analysis result. Also, there was a shortage of the frequency component needed for the analysis. Accordingly, if a wrong position-shift quantity is outputted, it turns out that the analysis flowchart will proceed based on this wrong position-shift quantity. In contrast thereto, in the present position-shift analysis method, the coincidence degree is outputted. Consequently, it is possible to execute the following analysis flowchart: Namely, a lower-limit value of the coincidence degree is set. Then, if the coincidence degree is found to be smaller than the lower-limit value, the coincidence degree is eliminated as a value which could not be analyzed. Moreover, the coincidence degree is automatically interpolated by an analysis result in proximity thereto. FIG. 6 illustrates a flowchart for determining a control-value change quantity of the de-scanning coil 11 needed for displacing a diffraction image by a predetermined quantity. Since the de-scanning coil 11 includes an x deflection and a y deflection, the control value for the de-scanning coil 11 is specified as f=(fx, fy). First, a reference diffraction image is photographed at f0=(0, 0). After that, diffraction images at several control values fi=(fxi, fyi) are photographed, then analyzing displacement quantities Fi=(FXi, FYi) relative to the reference diffraction image by using the image processing. A measurement condition therefor is set. For example, if movable range of the de-scanning coil is set as being ±1024 digits in the fx deflection and ±1024 digits in the fy deflection, a 121-point measurement condition is set in a lattice-like manner such that 11 points from −1000 digits to +1000 digits with a 100-digit spacing are specified to fx, and 11 points from −1000 digits to +1000 digits with a 100-digit spacing are specified to fy. Next, a conversion expression is acquired which determines the control-value change quantities of the de-scanning coil 11 needed for displacing the diffraction image by the predetermined quantities from measurement values acquired. No electromagnetic-field lens is provided between the de-scanning coil and the camera. Accordingly, it has been assumed that the conversion expression between the control-value change quantities of the de-scanning coil 11 and the displacement quantities of the diffraction image is describable by a 2×2 matrix A. Concretely, assuming the direction tx and pitch a of the fx deflection by the de-scanning coil, and the direction ty and pitch b of the fy deflection thereby (refer to FIG. 7), the following expression 2 is assumed: [ FX i FY i ] = [ a · cos ⁡ ( t x ) b · cos ⁡ ( t y ) a · sin ⁡ ( t x ) b · sin ⁡ ( t y ) ] · [ fx i fy i ] ( 2 ) Based on the measurement result, the parameters of the matrix A are optimized so that \|\|Fi−A·fi\|\| becomes equal to its minimum. An inverse matrix A−1 of the optimized matrix A is calculated. The control-value change quantities fi of the de-scanning coil 11 needed for displacing the diffraction image by the displacement quantities Fi are calculated by multiplying the displacement quantities Fi of the diffraction image by A−1. Incidentally, if a shift is large which exists between the displacement quantities calculated from the control-value change quantities of the de-scanning coil 11 by using the expression 2 and the actually-measured displacement quantities, the case is conceivable where, e.g., the axis deviation is large, and where the activation of the electrons changes in a vortex-like manner. In that case, it is necessary to add a term of the off-axis aberration to the conversion expression. Next, the de-scanning table is created from a displacement quantity of the diffraction image caused by the control-value change quantity of the scanning coil 5. Since the objective lens 7 and the projection lens 9 exist between the scanning coil 5 and the camera 15, there exists a possibility that the displacement quantity of the diffraction image may complicatedly change depending on setting conditions therefor. Accordingly, instead of trying to fit, into a mathematical expression, the displacement quantity of the diffraction image caused by the control-value change quantity of the scanning coil 5, it has been decided that the measurement values be used just as they are. Incidentally, it is insufficient to measure the same pixel number as that of the STEM image. Also, it can be assumed that the electromagnetic field will never change so steeply. Consequently, it has been decided that measurement points be provided with a proper spacing set therebetween, and that interpolation be performed using the spline interpolation. FIG. 8 illustrates a flowchart therefor. Since the scanning coil 5 also includes an x deflection and a y deflection, the control value for the scanning coil 5 is specified as s=(sx, sy). Here, it is assumed that movable range of the scanning coil 5 has been set as being 1 digit to 640 digits in the sx direction and 1 digit to 480 digits in the sy direction. A reference diffraction image at the center of the field-of-view, i.e., at s=(320, 240) digits, is photographed. After that, diffraction images at several control values si=(sxi, syi) for the scanning coil 5 are photographed, then analyzing displacement quantities Si=(SXi, SYi) relative to the reference diffraction image by using the image processing. As the measurement points., for example, 192 points are specified in a lattice-like manner such that 16 points from 20 digits to 620 digits with a 40-digit spacing are specified in the sx direction, and 12 points from 20 digits to 460 digits with a 40-digit spacing are specified in the sy direction. This specification makes it possible to measure, at the respective points, the displacement quantities of the diffraction image caused by the control-value change quantities of the scanning coil 5. Consequently, the control-value change quantities of the de-scanning coil 11 needed for canceling the above-described displacement quantities of the diffraction image are determined from the expression 2. Furthermore, the control-value change quantities determined are allocated onto a table whose pixel number is equal to the pixel number (i.e., 640×480) of the STEM image in accordance with the control values of the scanning coil 5 at the respective points. The value at a position to which the actually-measured result has been not allocated is interpolated by a method such as the spline interpolation, thereby completing the de-scanning table. The de-scanning table simultaneously stores therein the STEM magnification with which the de-scanning table had been created. The setting of the scanning range of the primary electron beam, i.e., the STEM magnification, is modified many times during the specimen observation. It is quite inefficient to create the de-scanning table every time the setting is modified. Accordingly, the values in the de-scanning table are multiplied by a value which is inversely proportional to the magnification, then being outputted to the de-scanning coil. The magnification at the time when the de-scanning table had been created is recorded. When executing the de-scanning, the magnification at present is read. Moreover, the values in the de-scanning table are multiplied by the inverse of ratio of the read magnification relative to the recorded magnification, then being outputted to the de-scanning coil. If the oscillation of the diffraction image due to the primary-electron-beam scanning becomes conspicuous because the difference between the read magnification and the recorded magnification is large, re-creating the de-scanning table will be carried out. The above-described creation of the de-scanning table is automatically executed in accordance with the flowchart in FIG. 8. This makes it unnecessary for a user to perform the complicated and troublesome task. All that the user has to do is to issue the instruction by using a de-scanning table update icon. FIG. 9 illustrates a configuration example of the main control screen. In the system configuration in FIG. 3, the main control screen in FIG. 9 is displayed on the monitor of the PC 16. The main control screen illustrated in FIG. 9 displays thereon requirements such as the STEM image, the diffraction image, and the icon for issuing the instruction of creating the de-scanning table. The judgment on execution of the de-scanning table update is made mainly based on the diffraction image. As illustrated in FIG. 12A, if the diffraction image is being tremendously oscillated by the primary-electron-beam scanning, and if the transmitted electron beam to be defected deviates from the angle selection aperture 13, updating the de-scanning table can be said to be absolutely necessary. As illustrated in FIG. 12B, if the diffraction image is being slightly oscillated, a judgment which meets an observation purpose should be made. Incidentally, FIG. 12C illustrates the diffraction image and the STEM image at the time when the de-scanning has been executed with an excellent accuracy. For example, when measuring the substantial configuration of a specimen, there exists no necessity for updating the de-scanning table. In the crystalline structure analysis, however, updating the de-scanning table is more advisable even if a slight oscillation were to be observed in the diffraction image. Clicking on or double-clicking on the de-scanning table update icon allows the most up-to-date de-scanning table to be created in accordance with the flowchart in FIG. 8. After that, the de-scanning coil will be controlled in accordance with this de-scanning table. Even the present device allows the most up-to-date de-scanning table to be created in a few minutes. Decreasing the number of the measurement points allows the most up-to-date de-scanning table to be created in a few tens of seconds. Incidentally, all that the end user has to do is to issue the instruction by using the de-scanning table update icon. Depending on the user, however, it is necessary to confirm whether or not the de-scanning table has been created correctly. Several sub screens therefor are prepared. FIG. 10 illustrates an example of the sub control screen. In addition to the de-scanning table update icon for issuing the instruction of creating the de-scanning table in accordance with the flowchart in FIG. 8, the following two icons are added: A conversion-expression update and adjustment icon for issuing an instruction of updating the conversion expression in accordance with the flowchart in FIG. 6, and a de-scanning execution icon for instructing on/off of the de-scanning operation based on the de-scanning table. Also, there is provided a display unit for digital-displaying the control values for the respective coils and the displacement quantities analyzed by using the image processing. As the display function, it is also advisable that the control values for the scanning coil 5 be indicated by circle marks on the STEM image photographed immediately before the control. A position whose analysis has been already finished is indicated by a hollow circle mark, and a position whose analysis is under way at present is indicated by a circle mark with another display color. The analysis results of the displacement quantities of the diffraction image are indicated by cross-character marks on the diffraction image. The control values for the de-scanning coil 11 are also indicated by circle marks on a two-dimensional map. This makes it easier to intuitively understand progression state of the analysis and correspondence between the control values for the de-scanning coil and the analysis results of the displacement quantities. There is also provided a set unit for specifying the measurement points. Also, there is provided a display unit for displaying the created de-scanning table (FIG. 11A to FIG. 11C). The analysis results of the displacement quantities of the diffraction image are displayed as two-dimensional images. This display makes it easier to perform confirmation of the analysis results. The analysis results (FX, FY) of the diffraction-image displacement quantities at the coil control values (fx, fy) are respectively displayed as the two-dimensional images. The changes in the displacement quantities can be assumed to be smooth. As a result, if the displacement quantities are analyzed correctly, smooth analysis results are displayed as illustrated in FIG. 11A. If the displacement quantities could not be analyzed because the coil control-value change quantities are too large and thus the displacement quantities of the diffraction image become too large, as illustrated in FIG. 11B, the analysis results steeply change in proximity to the analysis-incapable locations. When performing the interpolation processing in order to create the de-scanning table from the displacement quantities, the interpolation processing needs to be performed in such a manner that the result at a point which could not be analyzed is excluded, and that the result at a point which could be analyzed is utilized. The judgment as to whether or not the displacement quantities could be analyzed can be made from the correlation values between the images. The confirmation as to whether or not the de-scanning control values on the peripheral portions are correctly evaluated by this interpolation processing is also performed on the above-described two-dimensional image display of the de-scanning table. If, as illustrated in FIG. 11C, the values on the peripheral portions are not correctly evaluated, it is required to modify the measurement condition and to perform the analysis once again. As having been explained so far, in the system where the de-scanning coil makes it possible to cancel the displacement of the diffraction image caused by the primary-electron-beam scanning, the most important device performance is the correction accuracy. In the conventional control system, however, implementation of an enhancement in the correction accuracy has required a complicated and troublesome device adjustment. Accordingly, we have devised the following system: Namely, the de-scanning table is created from the actually-measured results of the displacement quantities of the diffraction image. Then, the de-scanning coil is controlled in a manner of being synchronized with the primary-electron-beam scanning. This system has allowed simultaneous implementation of both the enhancement in the correction accuracy and the simplicity of the device adjustment. The enhancement in the correction accuracy makes it possible to enlarge the field-of-view of the STEM image. Also, it becomes possible to make smaller the hole diameter of the angle selection aperture. This also allows an enhancement in the selection accuracy of the transmitted electron beam. In the first embodiment, as the de-scanning coil 11, the one-stage deflection coil has been used which is set up between the projection lens 9 and the electron detector 14. It is also possible, however, to carry out the de-scanning by using some other coil. For example, as illustrated in FIG. 13, carrying out the de-scanning by using a two-stage deflection coil is also allowable. The two-stage deflection exhibits a disadvantage that the deflection quantity becomes smaller than that of the one-stage deflection. However, the two-stage deflection exhibits an advantage that it becomes possible to adjust both of the angle and the position of the electron beam. Also, in order to clarify the functions of the respective coils, the detected-electron alignment coil 10 and the de-scanning coil 11 have been described in the separate manner. It is also possible, however, to implement these functions by using a single coil. Namely, if it is wished to simplify the coil control mechanism, integrating the coils into the single coil is advisable. Also, it is also possible to set the position of the de-scanning coil between the objective lens and the projection lens. This setting, however, complicates the correspondence between the control-value change quantities of the de-scanning coil 11 and the displacement quantities of the diffraction image. FIG. 16 illustrates a basic configuration diagram of the STEM-EELS used in the present embodiment. The EELS is installed on the lower portion of the STEM. In order to allow the transmitted electron beam to be detected by the EELS, as is the case with the angle selection aperture 13 and the electron detector 14, the electron detection camera 15 is also formed into a movable-type mechanism. The EELS includes a quadrupole lens 31, an accelerator 32, an electron spectrometer 33, a quadrupole magnifying lens 34, an energy slit 35, a mapping-use electron detector 36, a spectrum-use electron detector 37, and control units therefor. Next, the explanation will be given below concerning processing steps for acquiring an EELS image by using the device illustrated in FIG. 16. First of all, the STEM image is acquired in accordance with the processing steps illustrated in the first embodiment. Moreover, a transmitted electron beam to which an energy spectroscopy is to be applied is selected using the angle selection aperture 13. Then, the electron detector 14 and the electron detection camera 15 are deviated from the optical axis, thereby allowing the transmitted electron beam to enter the EELS. The electrons which have entered the EELS are energy-dispersed by the electron spectrometer 33, then being magnified by the quadrupole magnifying lens 34, and being projected onto the detectors. Furthermore, at first, the mapping-use electron detector 36 and the energy slit 35 are deviated from the optical axis, thus acquiring the energy spectrum by using the spectrum-use electron detector 37. Then, the energy spectrum is displaced in parallel by using the accelerator 32, thereby making an adjustment so that the transmitted electron beam having an energy width to be detected will pass through the energy slit 35. After the adjustment, the energy slit 35 and the mapping-use electron detector 36 are inserted onto the optical axis. This insertion causes an output signal from the mapping-use electron detector 36 and the primary-electron-beam scanning signal to be synchronized with each other, thereby making it possible to acquire the EELS image. If the position of a diffraction image on the EELS incident surface is changed by the primary-electron scanning, orbits of the transmitted electrons which are passing through inside the EELS change. As a result, positions of the transmitted electrons projected onto the electron detectors change. On account of this, a transmitted electron beam whose energy width differs from a predetermined energy width will pass through the energy slit 35, eventually entering the mapping-use electron detector 36. In order to prevent this phenomenon from occurring, there is provided a de-scanning coil 11 for canceling the displacement of the diffraction image on the EELS incident surface, the displacement being caused by a change in the incident position of the primary electron beam on the specimen. Control over this de-scanning coil 11 is performed using the de-scanning coil control system illustrated in FIG. 1. A control value for the scanning coil 5 is outputted from the DBC circuit. While being synchronized with this digital control signal, values in a de-scanning table registered in the FM (2) are outputted to the de-scanning coil 11. Here, the de-scanning table is created as follows: Diffraction images before and after activating the scanning coil 5 and the de-scanning coil 11 are photographed using a camera. Then, based on a result acquired by analyzing a displacement quantity of the diffraction images, the de-scanning table is created. First, a displacement quantity of the diffraction images caused by the de-scanning coil 11 is analyzed. From this result, a conversion expression is acquired which is used for determining a control-value change quantity of the de-scanning coil 11 needed for displacing the diffraction image by a predetermined quantity. Next, a displacement quantity of the diffraction images caused by the scanning coil 5 is analyzed. From the above-described conversion expression, control-value change quantities of the de-scanning coil 11 needed for canceling displacement quantities of the diffraction image are determined, then being recorded into the FM (2) as the de-scanning table. The control values read from the de-scanning table are sent to the de-scanning coil 11 in response to the incident position Rs of the primary electron beam on the specimen. This makes it possible to execute the high-accuracy de-scanning. In the third embodiment, the explanation has been given concerning the technique where the use of the FM (2) and the de-scanning coil 11 makes it possible to correct the position change in the diffraction image on the EELS incident surface caused by the primary-electron-beam scanning. In the present embodiment, the explanation will be given below regarding a technique where the use of the accelerator 32 makes it possible to correct the position of the transmitted electron beam relative to the energy slit 35. FIG. 17 illustrates a configuration diagram of a control system for the de-scanning coil 11 and the accelerator 32. A de-scanning table to be used for the de-scanning coil 11 is created in accordance with the processing steps described in the first embodiment, then being recorded into the FM (2). Moreover, while being synchronized with the scanning coil 5, values in the de-scanning table registered in the FM (2) are sent to the de-scanning coil 11. In the EELS, the image on the incident surface reaches the emission surface in a manner of being magnified. As a result, the incident position of or the angle change in the transmitted electron beam on the incident surface is projected onto the energy slit 35 in a manner of being magnified inside the EELS. In order to correct this phenomenon so that the transmitted electron beam having a predetermine energy width will pass through the energy slit, the position of the transmitted electron beam relative to the energy slit 35 is corrected using the accelerator 32. In order to implement this correction by using the accelerator, a FM (2)′ is provided. Then, taking advantage of zero-loss spectrum, a displacement of the energy spectrum caused by the primary-electron-beam scanning is measured. Next, control values of the accelerator 32 needed for canceling this displacement are determined, then being stored into the FM (2)′. The data stored into the FM (2)′ are the data which are measured while correcting the position of the diffraction image on the EELS incident surface by using the FM (2) and the de-scanning coil 11. It is required to update the data in the FM (2)′ without fail whenever the data in the FM (2) are modified. Furthermore, while being synchronized with the control over the scanning coil 5, the data in the FM (2) are sent to the de-scanning coil 11. Simultaneously therewith, the data in the FM (2)′ are sent to the accelerator 32. This makes it possible to enhance the de-scanning accuracy even further. It should be further understood by those skilled in the art that although the foregoing description has been made on embodiments of the invention, the invention is not limited thereto and various changes and modifications may be made without departing from the spirit of the invention and the scope of the appended claims.
	abstract	The object of this invention is to provide a method for mitigating a stress corrosion cracking of reactor structural material which makes it possible to suppress the rise in the main steam line dose rate without secondary effects such as a rise in the concentration of radioactive cobalt-60, etc. in the reactor water. Hydrogen and a reductive nitrogen compound containing nitrogen having a negative oxidation number (for example, hydrazine) are injected into the core water of boiling water nuclear power plant. By injecting the reductive nitrogen compound containing nitrogen having a negative oxidation number into the core water, the stress corrosion cracking of structural material of reactor can be mitigated without side reactions such as a rise in the concentration of cobalt-60, etc.
050892142	summary	BACKGROUND OF THE INVENTION This invention generally relates to pressure monitors, and is specifically concerned with both an apparatus and a method for directly and reliably monitoring the pressure of helium gas within a cask containing radioactive materials. Devices for monitoring the pressure of the helium gas that is typically present within casks used to store or transport radioactive materials are known in the prior art. One of the primary purposes of such devices is to generate a warning signal when a leakage condition occurs which would allow the pressurized helium gas contained within such casks to escape in the ambient atmosphere. There are many reasons why the persons responsible for the maintenance and operation of such casks would want to be immediately informed of such a leakage condition. First, while the helium gas itself is not harmful, the leakage of such gas might carry out very fine radioactive particulates which may in turn pose a radiation hazard. Secondly, the loss of helium gas within such casks interferes with the ability of the cask to dissipate the heat generated by the decay of the radioisotopes disposed within the cask, as helium is a far better thermal conductor than air. Thirdly, the loss of helium in the cask might imply the introduction of ambient air into the cask interior, which is potentially corrosive due to the oxygen content of air. In one type of prior art pressure monitoring device, the space that is normally present between the inner and outer lids that cover the cask is pressurized to a higher pressure than the helium gas within the cask interior. This pressurized space is connected to a differential pressure switch that monitors the pressure difference between the gas in the cask and the gas in the space. A second switch monitors the pressure in the pressurized space between the inner and outer lids, and is used to determine whether or not a change in the differential pressure between the cask interior and the pressurized space is the result of a leak in the seals between the cask interior and the ambient atmosphere, or a leak in the seals between the pressurized space and the ambient atmosphere. While such prior art pressure monitoring devices are capable of fulfilling their intended purpose, the applicants have noted a number of areas in the design of these devices which could bear improvement. For example, in order to obtain access to these particular pressure monitoring devices, the lid of the cask itself must be completely removed. Such lid removal is not only troublesome in view of the size and weight of the lid, and the number of bolts used to fasten it to the cask, but further presents a radiation hazard since it results in the exposure of radioactive materials to the ambient atmosphere. Hence, whenever it becomes necessary to perform a maintenance operation or to replace a component in one of these prior art pressure monitoring devices, the cask must be moved into an area of containment, the heavy lid removed, and either the radioactive materials disposed inside must be removed, or the maintenance or replacement operation must be done remotely through the use of robotic tools so that maintenance personnel are not exposed to potentially harmful radiation. Still another shortcoming in the design of such prior art pressure monitoring devices is the fact that there is no practical way to test the operability of the pressure sensors once the cask is sealed, or to confirm the reliability of the pressure readings generated by the two pressure sensors when these sensors indicate that a leakage condition has occurred. Hence, if one or more of the pressure switches generates a spurious leakage signal as a result of drift in its set point or some other malfunction, the entire cask might be put through some unnecessary and expensive repair operation. Still a third shortcoming in the design of such prior art pressure monitoring devices is the fact that neither of the pressure switches makes a direct measurement of the actual pressure of the pressurized helium gas inside the cask. The lack of any such direct measurement adversely effects the reliability of the pressure readings generated by the switches. Clearly, what is needed is an improved pressure monitoring device which is easily accessible in the event that a repair or a maintenance operation is necessary, but yet does not adversely effect the shielding efficacy of the cask as a whole. Ideally, the readings of such a pressure monitoring device should be readily testable at any time during the operation of the device, and should be further verifiable in the event that a signal indicative of a leakage condition is generated. Finally, such a device should directly measure the pressure of the gas disposed within the cask without compromising the gas seals in the cask so that the output of the device may be as accurate and as reliable as possible. SUMMARY OF THE INVENTION The invention is both an apparatus and method for monitoring the pressure within a cask containing a hazardous gas that eliminates or at least ameliorates the aforementioned shortcomings of the prior art. Specifically, the invention is an apparatus that comprises a differential pressure sensor sealingly connected to an outer end of a bore that penetrates through a wall of the cask for both directly measuring the pressure of the hazardous gas and providing a first barrier between the gas and the ambient atmosphere, and an evacuated sensor chamber that contains the outer end of the through-wall bore as well as the differential pressure sensor for providing a second barrier between the gas and the ambient atmosphere. Preferably, the apparatus further comprises an absolute pressure sensor that is connected in parallel to the through-wall bore so that, in the event that the differential pressure sensor indicates a leak condition has occurred, the system operator may determine whether or not the leakage condition sensed is a result of a leakage condition in the cask, or in the evacuated sensor chamber. Both sensors may be switches. A wall of the evacuated sensor chamber may further include a test port coupling, and the apparatus may further comprise an auxiliary pressure sensor that is detachably and sealingly connectable to the test port coupling for measuring the pressure within the sensor chamber such a measurement may be made through the test port coupling in the event that the absolute pressure sensor indicates that the loss of differential pressure sensed by the differential pressure sensor is a result of leakage in the evacuated sensor chamber. Such a direct pressure measurement would either confirm that the leakage condition arose as a result of a leak in the evacuated pressure sensor, or would indicate that the leakage condition was a false alarm caused by a faulty differential pressure sensor. To facilitate the replacement of either the differential or the absolute pressure sensor, the evacuated sensor chamber is defined in part by a removable outer cover. Additionally, both the differential and the absolute pressure sensors may be connected to the aforementioned through-wall bore in the walls of the cask by means of a gas-conducting conduit which in turn includes at least one isolation valve for isolating the through-wall bore from the pressure sensors during a pressure sensor replacement operation. As an added safety measure in the event that a pressure sensor replacement operation become necessary, the apparatus may further include a venting assembly in the gas conducting conduit between the previously mentioned isolation valve, and the inputs of both the differential pressure sensor and the absolute pressure sensor. This venting assembly may include a vent port, and a vent valve fluidly connected within the segment of the gas-conducting conduit that connects the isolation valve to the inputs of the differential and absolute pressure sensors. Both the differential pressure sensor and the absolute pressure sensor generate electrical signals corresponding to a pressure output that are transmitted through wires which extend through the removable cover of the evacuated sensor chamber through a sealed electrical penetration. These wires are in turn connected to an electrical connector assembly or socket that is mounted in a hole in the housing of the pressure monitoring assembly. An output cable having a plug which is receivable within the socket transmits the electrical signals generated by the differential and absolute pressure sensors to an appropriate monitoring circuit, which converts these signals into pressure values and which is programmed to generate an alarm signal upon the receipt of a pressure signal indicative of a leakage condition. The apparatus provides a highly reliable system whose outputs are directly and reliably confirmable at all times during the operation of the device, and whose structure provides multiple safeguards against the release of any hazardous gases into the ambient atmosphere.

Subsets and Splits

No community queries yet

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